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Sample records for candu steam generator

  1. CANDU steam generator life management

    International Nuclear Information System (INIS)

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

  2. CANDU steam generator life management

    International Nuclear Information System (INIS)

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

  3. CANDU steam generator life management

    Energy Technology Data Exchange (ETDEWEB)

    Tapping, R.L. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Nickerson, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Spekkens, P.; Maruska, C. [Ontario Hydro, Toronto, Ontario (Canada)

    1998-01-01

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

  4. Design and performance of CANDU steam generators

    International Nuclear Information System (INIS)

    The recirculating U-tube steam generators at Ontario Hydro's Pickering and Bruce CANDU Nuclear Generating Stations have demonstrated excellent reliability over many years of operation. Tube failures have been rare, contributing to high plant capacity factors. Of the approximately 390,000 steam generator U-tubes at the Pickering and Bruce plants, one tube leak has occurred at Pickering to date and 12 tubes have leaked at Bruce in a total of 89 reactor years of operation. The success of these units is attributed to an age old respect for steam generation equipment, an ongoing pursuit of research and design advancements, and extensive cooperative efforts on the part of the utility, the system designer, and the supplier. The supplier's involvement began with the steam generator design and manufacture for the very first CANDU plant. The utility's involvement began with their direct participation in the earliest stages of nuclear plant design. The utilities' contribution to the success of these units relates to the rigorous approach used in definition of requirements, in understanding the supplier's design in detail and in making every effort to operate, monitor, and service the equipment with appropriate care. This paper presents the supplier's and utility's approach to achieving these remarkable results. In order to present these viewpoints separately and in some detail, this paper is divided into two parts: Part 1 - The Equipment Supplier Perspective, and Part 2 - The Utility Perspective

  5. Health monitoring requirements for CANDU steam generators

    International Nuclear Information System (INIS)

    AECL is developing an equipment health monitoring (EHM) module as part of SMART CANDU development to provide station maintenance personnel with the information required to assess the condition of critical station equipment and to predict when maintenance is required. SMART CANDU is a suite of software applications that is being developed by AECL to help station staff to efficiently and effectively implement their health monitoring and ageing management programs. The EHM application integrates information from all relevant station sources (e.g., on-line instruments, local 'smart' field components, walk-down data, and inspection and monitoring software) on the station local area network and presents the user with a snapshot of the current health of the component of interest. The EHM application also permits staff to be proactive by alerting them to the early warning signs of degraded equipment functionality and to potential equipment failure. User requirements for a steam generator EHM display described in the present work are being developed in consultation with station staff and subject matter experts. The required information is being collected from multiple sources and may include, for example, on-line and grab sample process and chemistry data, inspection results from eddy current and ultrasonic measurements and maintenance information (e.g., chemical cleaning, plugged tubes). These data can be combined with AECL predictive models for steam generator fouling, crevice chemistry, flow induced vibration, etc. to provide station staff with an assessment of current steam generator conditions and help mitigate key degradation mechanisms throughout the life of the CANDU station. (author)

  6. Advancing CANDU experience to the world steam generator market

    International Nuclear Information System (INIS)

    Tube degradation in certain recirculating nuclear steam generators has provided a market for steam generator replacement. Prior to this need, B and W supplied over 200 steam generators for CANDU nuclear plants. With this experience, and implementing extensive research and development improvements in material selection, design enhancements, and new manufacturing and analytical methods, B and W has supplied or secured orders for the replacement of 26 steam generators. Along with plans for new replacement orders, B and W will continue to supply steam generators for future CANDU plants. This paper will review the progression of B and W's CANDU experience to meet the replacement steam generator market, and examine the continuous improvements required for today's increasingly demanding nuclear specifications. (author). 1 tab., 4 figs

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

    International Nuclear Information System (INIS)

    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

  8. CANDU 6 steam generator thermalhydraulic modeling and simulation

    International Nuclear Information System (INIS)

    The main objective of this paper is to describe the process of accurately modeling the dynamic behavior of CANada Deuterium Uranium 6 (CANDU 6) steam generator and its control algorithm. The mathematical model of the steam generator was developed mainly through differential equations calculated from the physical properties of the components. In addition, empirical modeling techniques were utilized in order to incorporate more complex properties of steam generators. The controller design and resulting controller performance on the actual plant are both strongly dependent on the accuracy of the mathematical model used to describe the plant. During power manoeuvrings, the level control in a steam generator is complicated by the thermal reverse effects known as 'shrink and swell'. This non-linear non-minimum phase characteristic of the steam generators is most challenging to model. Based on various recent publications, the root cause of the deficiencies in model-based controller designs is the fact that the 'shrink and swell' phenomenon is not captured by the plant model. The model developed in this project captures this phenomenon and clearly presents the adverse effect of such characteristic in the performance of conventional controllers. The resulting mathematical model developed in this project is implemented in Simulink, a graphical block diagramming tool offered by MATLAB. In addition to the plant model, the complete control algorithm of the CANDU 6 steam generator is modeled in this project. This model duplicates the actual control strategy applied in the existing steam generators in order to create a realistic interaction between the plant and its control algorithm. The control model is developed in discrete-time in order to accurately simulate the output of the digital controller. The analog model of the plant is dynamically integrated with the digital control model in the Simulink environment creating a realistic presentation of the actual communication

  9. Steam generator cleanness management. A comprehensive concept for CANDU plants

    International Nuclear Information System (INIS)

    Steam generator tubes are by far the largest boundary between the primary and secondary side and the overall performance of the plant depends strongly on the steam generators. The steam generator efficiency can be negatively affected by Magnetite deposition on the tubes on either ID or OD. These deposits are a thread to the integrity of the SG due to either ID or OD corrosion. Deposits in the steam generators may promote corrosion phenomena, leading in worst case to a steam generator replacement. A comprehensive cleanness concept, based on optimized water chemistry, chemical and mechanical cleaning processes is mandatory and will be outlined in the following. (author)

  10. Research experience with the secondary side corrosion of CANDU steam generators

    International Nuclear Information System (INIS)

    All steam generator tube failures lead to transfer of the radioactive materials from the primary coolant circuit to the steam generator secondary circuit, and necessitate downtime to locate and plug the failed tubes. For the particular case of the CANDU plants, any steam generator tube failure results in an additional economic penalty through the loss of heavy water. Nearly all of the failures were attributed to secondary side water chemistry conditions and excursions, many of which resulted from condenser cooling water ingress. The investigation of the structural materials corrosion in correlation with the water chemistry, as well as the determination of impurities and corrosion products concentration and deposition and their removing from the CANDU steam generators is a very active field. Both the experimental works and the understanding of the mechanisms involved are submitted to some rapid changes and permanently open to research. To provide information about the corrosion behaviour of the structural materials from CANDU steam generators under normal and abnormal conditions of operation and to identify the failure types produced by corrosion there were performed a lot of corrosion experiments. These experiments consisted in chemical accelerated tests, static autoclaving and electrochemical investigations. The goal of this paper consists in the assessment of the corrosion kinetics for the Incoloy-800 at normal secondary circuit steam generator parameters. The gravimetric method, optical metallographic microscopy, XRD analysis as well as electrochemical measurements have been used to evaluate the corrosion behavior of Incoloy-800. (authors)

  11. CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

    International Nuclear Information System (INIS)

    CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

  12. Two-phase flow induced vibrations in CANDU steam generators

    International Nuclear Information System (INIS)

    The U-Bend region of nuclear steam generators tube bundles have suffered from two-phase cross flow induced vibrations. Tubes in this region have experienced high amplitude vibrations leading to catastrophic failures. Turbulent buffeting and fluid-elastic instability has been identified as the main causes. Previous investigations have focused on flow regime and two-phase flow damping ratio. However, tube bundles in steam generators have vapour generated on the surface of the tubes, which might affect the flow regime, void fraction distribution, turbulent intensity levels and tube-flow interaction, all of which have the potential to change the tube vibration response. A cantilevered tube bundle made of electric cartridges heaters was built and tested in a Freon-11 flow loop at McMaster University. Tubes were arranged in a parallel triangular configuration. The bundle was exposed to two-phase cross flows consisting of different combinations of void from two sources, void generated upstream of the bundle and void generated at the surface of the tubes. Tube tip vibration response was measured optically and void fraction was measured by gamma densitometry technique. It was found that tube vibration amplitude in the transverse direction was reduced by a factor of eight for void fraction generated at the tube surfaces only, when compared to the upstream only void generation case. The main explanation for this effect is a reduction in the correlation length of the turbulent buffeting forcing function. Theoretical calculations of the tube vibration response due to turbulent buffeting under the same experimental conditions predicted a similar reduction in tube amplitude. The void fraction for the fluid-elastic instability threshold in the presence of tube bundle void fraction generation was higher than that for the upstream void fraction generation case. The first explanation of this difference is the level of turbulent buffeting forces the tube bundle was exposed to

  13. Eddy Currents Inspection of CANDU Steam Generator Tubes using Zetec's ZR-1 Robot. Experience in Romania

    International Nuclear Information System (INIS)

    Full text of publication follows: The commercial operation of Unit 1 of Cernavoda NPP started on 2 December, 1996. The unit's reactor type is PHWR-CANDU 6 (electrical capacity 706 MWe), using natural uranium. The nuclear fuel is manufactured in Romania. The Cernavoda nuclear power plant has four CANDU - design steam generators that have been in service since 1996. The paper introduces the new ZR-1 Robot System for Inspection and Maintenance/Repair from Zetec that combines the newest state-of-the-art robotics technology with Zetec experience - based innovation to address the needs for inspection and repair of steam generators. The multipurpose ZR-1 can be easily installed to perform the necessary eddy current inspection and remain installed ready for follow-up maintenance and repair. It has superior technical performances and a modular three axis motion of arm that enables 100% coverage of tube sheet. Automated, repeatable, and precise positioning of tool heads ensures accurate delivery and reducing costly rework and reduces inspection time by 30%. The modular, light weight, and portable design permits easy assembly and disassembly through small openings and it reduces setup/tear down time by 30%. The first deployment of the new ZR-1 Robot was made in September 2004 at the Cernavoda NPP inspection outage. The unit's reactor type is PHWR-CANDU 6 (electrical capacity 706 MWe), using natural uranium; the nuclear fuel is manufactured in Romania. The Cernavoda nuclear power plant Unit 1 has four CANDU - design steam generators that have been in service since 1996. The paper presents also the Zetec's field experience and customer experience with this system. It describes the equipment setup in Cernavoda's steam generators mock-up, functional tests and calibration. Finally, provides details on the execution of the inspection, options for standardizing the inspection techniques and conclusions. (authors)

  14. 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.

  15. An improved ultrasonic downcomer flow-measurement system for CANDU steam generators

    International Nuclear Information System (INIS)

    Ultrasonic measurements of downcomer flow velocity have been successfully used in the past to determine re-circulation ratios and water inventory in CANDU steam generators. Knowledge of these process conditions allows operators to assess the effectiveness of maintenance programs, monitor the effects of tube fouling, observe flow conditions following component modifications, and provides designers with a means to validate or improve code predictions. Non-intrusive ultrasonic measurement systems were recently installed on four steam generators at the Bruce B Nuclear Generating Station as part of an investigation into the possible effects of long-term degradation due to internal Flow-Accelerated Corrosion (FAC). The most recent version of AECL's downcomer-flow measurement technology was used in which buffered ultrasonic transducers are magnetically attached and then welded to the steam-generator outer shell. This method of attachment eliminates the complications of precision surface preparation and high-temperature couplants. The paper outlines the new attachment method and summarizes flow velocities measured during start-up, shut-down and normal operation. It also briefly describes how the information may be used to assess thermalhydraulic conditions, verify design calculations, and support the case for reactor uprating. (author)

  16. The susceptibility of Candu steam generator tubing alloys to IGA/IGSCC in acid sulphate environments

    International Nuclear Information System (INIS)

    Constant extension rate tests (CERT) were carried out to assess the susceptibility of CANDU steam generator (SG) tube materials to intergranular stress corrosion cracking (IGSCC) in acidified sulphate solutions calculated to exist in SG crevices following sulphuric acid and Lake Huron water ingress. The results indicate significant susceptibility of Alloy 600 tubing (i.e. Bruce A) to IGSCC following sulphuric acid ingress and some susceptibility to intergranular attack (IGA) following Lake Huron water ingress. Alloy 800 was slightly susceptible to IGA in sulphuric acid ingress crevice chemistries but not at all to Lake Huron water ingress crevice chemistry. Alloy 690 was not susceptible to IGSCC or IGA in any of the chemistries tested. Preliminary results suggest that lead contamination of 1000 ppm does not increase the susceptibility of any of the Alloys tested to IGSCC following sulphuric acid ingress. (authors). 3 figs., 2 tabs., 6 refs

  17. Solubility and mobility of PB-bearing species under CANDU steam generator conditions

    International Nuclear Information System (INIS)

    As part of a coordinated program, AECL is developing a set of tools to aid with the prediction and management of steam generator performance. One important mode of degradation within steam generators is intergranular and transgranular stress corrosion cracking (SCC) of SG tubing. Numerous and extensive laboratory studies have demonstrated a definite link between lead and SCC. The prediction of the steam generator degradation is hindered by the lack of reliable thermodynamic data on the solubilities of lead compounds under operation steam generator conditions. For lead-SCC to occur, lead needs to be transported to the metal-oxide interface on the SG tube. AECL has been developing a model to predict such a 'migration' of lead towards (or away from) the metal interface. If the direction of the lead migration could be predicted as a function of water chemistry conditions, practical means of dealing with SG contamination with Pb could be devised and Pb-SCC could be mitigated by appropriate adjustment of the water chemistry. The solubilities for the relevant Pb-bearing species under SG operating conditions are necessary for the application of this model. However, the lack of reliable thermodynamic data makes it difficult to calculate the solubilities of lead compounds as a function of temperature, pH, and chloride or sulphate concentration in SG under-deposit environments. In this paper, the experimental method used for the determination of Pb solubility is presented and the initial experimental results for several lead compounds given as a function of temperature, pH and ionic strength under the range of conditions relevant to CANDU SGs. The model describing the possible migration of Pb species to the surface of alloys will also be presented. (author)

  18. Experimental modeling of flow-induced vibration of multi-span U-tubes in a CANDU steam generator

    International Nuclear Information System (INIS)

    Flow-induced vibration of the tubes in a nuclear steam generator is a concern for designers who are trying to increase the life span of these units. The dominant excitation mechanisms are fluidelastic instability and random turbulence excitation. The outermost U-bend region of the tubes is of greatest concern because the flow is almost perpendicular to the tube axis and the unsupported span is relatively long. The support system in this region must be well designed in order to minimize fretting wear of the tubes at the support locations. Much of the previous testing was conducted on straight single-span or cantilevered tubes in cross-flow. However, the dynamic response of steam generator multi-span U-tubes with clearance supports is expected to be different. Accurate modeling of the tube dynamics is important to properly simulate the dynamic interaction of the tube and supports. This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program was initiated at AECL to demonstrate that the tube support design for future CANDU steam generators will meet the stringent requirements associated with a 60 year design life. The main objective of the tests is to address the issue of in-plane and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with Anti-Vibration Bar (AVB) supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper. (author)

  19. Eddy currents inspection of CANDU steam generator' tubes using Zetec's ZR-1 Robot: experience in Romania

    International Nuclear Information System (INIS)

    'Full text:' The paper introduces the new ZR-1 Robot System for Inspection and Maintenance/Repair from Zetec that combines the newest state-of-the-art robotics technology with Zetec experience-based innovation to address the needs for inspection and repair of steam generators. The multipurpose ZR-1 can be easily installed to perform the necessary eddy current inspection and remain installed ready for follow-up maintenance and repair. It has superior technical performances and a modular three axis motion of arm that enables 100% coverage of tube sheet. Automated, repeatable, and precise positioning of toolheads, ensures accurate delivery and reducing costly rework and reduces inspection time by 30%. The modular, lightweight, and portable design permits easy assembly and disassembly through small openings and it reduces setup/tear down time by 30%. The first deployment of the new ZR-1 Robot was made in September 2004 at the Cernavoda NPP inspection outage. The Cernavoda plant has four Advanced 600 MW CANDU-design generators that have been in service since 1996. The paper presents also the Zetec's filed experience and customer experience with this system. It describes the equipment setup in Cernavoda's generator mock-up, functional testes and calibration. Finally, provides details on the execution of the inspection, options for standardizing the inspection techniques and conclusions. (author)

  20. Experimental research regarding the corrosion of incoloy-800 and SA 508 cl.2 in the CANDU steam generator

    International Nuclear Information System (INIS)

    Steam generators (SGs) are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Water Reactor (PWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators. Steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related a corrosion. The feedwater that enters into the steam generators under normal operating conditions is extremely pure, but nevertheless contains low levels (generally in the μg/l concentration range) of impurities such as iron, chloride, sulphate, silicate, etc. When water is converted to steam and exits the steam generator, the non-volatile impurities are left behind. As a result, their concentration in the bulk steam generator water is considerably higher than those in the feedwater. Nevertheless, the concentrations of corrosive impurities are still generally sufficiently low that the bulk water is not significantly aggressive towards steam generator materials. The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. The purpose of this paper consists in assessment of generalized corrosion behaviour of the tubes materials (Incoloy-800) and tubesheet material (carbon steel SA 508 cl.2) at the normal secondary circuit parameters (temperature-260 deg C, pressure-5.1MPa). The testing environment was the demineralized water without impurities, at pH=9.5 regulated with morpholine and ciclohexilamine (all volatile treatment - AVT). The results are presented like micrographies and graphics representing loss of metal

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

    International Nuclear Information System (INIS)

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

  2. Next generation CANDU plants

    International Nuclear Information System (INIS)

    Future CANDU designs will continue to meet the emerging design and performance requirements expected by the operating utilities. The next generation CANDU products will integrate new technologies into both the product features as well as into the engineering and construction work processes associated with delivering the products. The timely incorporation of advanced design features is the approach adopted for the development of the next generation of CANDU. AECL's current products consist of 700MW Class CANDU 6 and 900 MW Class CANDU 9. Evolutionary improvements are continuing with our CANDU products to enhance their adaptability to meet customers ever increasing need for higher output. Our key product drivers are for improved safety, environmental protection and improved cost effectiveness. Towards these goals we have made excellent progress in Research and Development and our investments are continuing in areas such as fuel channels and passive safety. Our long term focus is utilizing the fuel cycle flexibility of CANDU reactors as part of the long term energy mix

  3. Water chemistry in secondary side of cne candu steam generators and their related degradation processes

    International Nuclear Information System (INIS)

    This paper presents a brief overview of steam generator functional parameters of nuclear power plants how use PWR (Pressurized Water Reactor) and PHWR (Pressurized Heavy Water Reactor) reactors, followed by a description of fundamental aspects on steam generator degradation and water chemistry in the secondary side, and also, water chemistry improvement by controlling ph. During operating life, cooling radioactive contamination occurs, but water conditions must be maintained inside specific ranges. Feedwater must be maintained as free from impurities as possible. This requirement involves careful attention to the entire system through which the water flows, either in the form of steam or water, for even though water is used as feedwater be pure at the same time of its entry into the system, it may absorb impurities from the various parts of the installation. Specific attention should be directed to possible points of water leakage from the service water system, as in the main and auxiliary condensers. Feedwater must be treated to maintain the required water conditions. As the concentration of the impurities in deposits increase, the ph can shift locally in these areas to acidic or alkaline conditions, entering in a ph range where initiation of corrosion phenomena cannot be longer excluded. By maintaining of sufficiently reducing conditions, the occurrence of certain corrosion mechanisms will be excluded (like pitting), but certain forms of steam generator tube corrosion may still occur. (authors)

  4. INR experience concerning the assessment of the CANDU steam generator tubing material degradation

    International Nuclear Information System (INIS)

    Steam generator degradation has caused substantial losses of power generation, resulted in large repair and maintenance costs. Institute for Nuclear Research has carried out an extensive R and D program focused on the understanding of the degradation processes especially for the tubing material and on developing remedial actions in the purpose to prevent and diminish the ageing process of which evolution supposes some considerable economic costs. Because of the huge impact of corrosion, it is imperative to have a systematic approach to recognizing and mitigating corrosion problems as soon as possible after they become apparent. A proper failure analysis includes collection of pertinent background data and service history, followed by visual inspection, photographic documentation, material evaluation, data review and conclusion procurement. In analyzing corrosion failures, one must recognize the wide range of common corrosion mechanisms. The features of any corrosion failure give strong clues as to the most likely cause of the corrosion. The principal steps of analysis and diagnosis of the steam generator tubes degradations consist in: visual inspection, chemical analysis, cross section examination by optical and scanning electron microscopy and RDX, data review, conclusions and recommendations. This paper details a proven approach to properly determining the root cause of a failure, and includes metallographic illustrations of the most common corrosion mechanisms, including general corrosion, pitting, crevice corrosion, corrosion fatigue and intergranular corrosion. (author)

  5. Steam generator life management

    International Nuclear Information System (INIS)

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

  6. Steam generator life management

    Energy Technology Data Exchange (ETDEWEB)

    Tapping, R.L. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Nickerson, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Spekkens, P.; Maruska, C. [Ontario Hydro, Toronto, Ontario (Canada)

    1998-01-01

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

  7. Eddy Currents Inspection of CANDU Steam Generators' Tubes using Zetec's ZR-1 Robot. Experience in Romania

    International Nuclear Information System (INIS)

    This is a PowerPoint presentation on behalf of COMPCONTROL ING, a Romanian private company established in 1997 the main services of which are enlisted. It is stressed that the most suitable type of inspection in terms of safety and reliability for the steam generator tubes is eddy current (EC) method. The advantages of EC testing include the following: - Extremely fast; - Accurate in detection and sizing of discontinuities; - Very good method for baseline screening; - Very high detection sensitivity to physical-chemical variations of the test specimen; - Easy setup and application for automated inspection; - Portable equipment designing; - Use of multiple channels and multi-frequencies for a better screening of signals and efficiency; - High capability to store the data for future review and comparison (using data history to evaluate the rate of degradation and life assessment studies). Between 2003 and 2005 ECT was applied to Cernavoda NPP U1 SGs as follows: - in 2003, SG-4; - in 2004, SG-2; - in 2005, SG-1; - in 2005, SG-3; - in 2005, SG-4. The purpose of inspection with eddy currents of SGs tubes was: - Detection, sizing and evaluation of possible degradations of the tubes and at the interface tube/support structures (tubesheet, tube support plates and baffles); - Completion of the baseline data for future review and comparison. The software used for acquisition and analysis of eddy current data and for inspection management were: - ZETEC Eddynet-R Zetec Acquisition Control-ZAC; - ZETEC Eddynet-R Data Analysis (bobbin and MRPC); - ZETEC Eddynet-R Data Management. The equipment ZR-1 is described and its advantages as well. Advantages of the automated scanning system are highlighted as follows: - Repeatability; - High resolution mapping; - Accurate indexing; - Minimize changes in lift-off resulting from probe wobble, eccentricity of the tube and surface irregularities; - 3-part design makes each component lighter and more compact for easier, faster installation

  8. Experimental research on corrosion in constrained circulation zones in CANDU steam generator

    International Nuclear Information System (INIS)

    Corrosion is the major problem affecting safe operation of steam generator (SG). Most of the problems relating to corrosion are due to the local concentration effects of aggressive species and/or impurities in regions with constrained flow such as the cracks at tube-tubular plate joints. The concentration effects are of great importance and present interest in designing and SG operation. In this work we present the results of the corrosion tests performed in operation conditions specific to SG secondary circuit (temperature = 260 deg.C, pressure = 5.1 MPa) on devices simulating cracks in SA 508 cl.2 and Incoloy-800 carbon steel. Testing conditions were: demineralized water (pH = 9.5 controlled with volatile amines), raw Danube River water (pH = 9.5) and NaCl 100 g/l solution (pH 10.5). Investigation of corrosion behavior of the two materials was done metallographically and by X-ray diffraction. The results are presented as micrographs evidencing occurrence of pitting corrosion first on the tubular plate material (SA 508 cl.2) and in an extremely aggressive environment and than on the Incoloy-800 tubing material. A table is given with the geometrical, medium, electrochemical reaction, and alloy composition factors influencing the corrosion resistance in the crack. The following three conclusions are presented: 1. The main source of impurities (which can penetrate in SG) is the cooling water infiltration from condenser (raw Danube River water) as well as water of addition improperly treated. To these, one contribute high concentration factors (105 - 106) of the dissolved substances which can reach the crack and which can constitute extremely aggressive media generated from a medium seemingly non-aggressive. 2. The Incoloy-800 samples checked for 2,400 hours in demineralized and Danube River waters with pH = 9..5 presented an oxide protecting thin layer while in the case of the samples exposed in a solution containing NaCl 100 g/l with pH 10.5 the pitting attack is

  9. Next generation CANDU plants

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water Reactors systems featuring horizontal fuel channels and heavy water moderator will continue to evolve, supported by AECL's strong commitment to comprehensive R and D programs. There are three key CANDU development strategic thrusts: improved economics, fuel cycle flexibility, and enhanced safety operation based on design feedback. Therefore, CANDU reactor products will continue to evolve by incorporating further improvements and advanced features that will be arising from our CANDU Technology R and D programs in areas such as heavy water and tritium, control and instrumentation, fuel and fuel cycles, systems and equipment and safety and constructability. Progressive CANDU development will continue in AECL to enhance the medium size product - CANDU 6, and to evolve the larger size product - CANDU 9. The development of features for CANDU 6 and CANDU 9 is carried out in parallel. Developments completed for one reactor size can then be applied to the other design with minimum costs and risk. (author)

  10. Next Generation CANDU Performance Assurance

    International Nuclear Information System (INIS)

    AECL is developing a next generation CANDU design to meet market requirements for low cost, reliable energy supplies. The primary product development objective is to achieve a capital cost substantially lower than the current nuclear plant costs, such that the next generation plant will be competitive with alternative options for large-scale base-load electricity supply. However, other customer requirements, including safety, low-operating costs and reliable performance, are being addressed as equally important design requirements. The main focus of this paper is to address the development directions that will provide performance assurance. The next generation CANDU is an evolutionary extension of the proven CANDU 6 design. There are eight CANDU 6 units in operation in four countries around the world and further three units are under construction. These units provide a sound basis for projecting highly reliable performance for the next generation CANDU. In addition, the next generation CANDU program includes development and qualification activities that will address the new features and design extensions in the advanced plant. To limit product development risk and to enhance performance assurance, the next generation CANDU design features and performance parameters have been carefully reviewed during the concept development phase and have been deliberately selected so as to be well founded on the existing CANDU knowledge base. Planned research and development activities are required only to provide confirmation of the projected performance within a modest extension of the established database. Necessary qualification tests will be carried out within the time frame of the development program, to establish a proven design prior to the start of a construction project. This development support work coupled with ongoing AECL programs to support and enhance the performance and reliability of the existing CANDU plants will provide sound assurance that the next generation

  11. The next generation CANDU 6

    International Nuclear Information System (INIS)

    AECL's product line of CANDU 6 and CANDU 9 nuclear power plants are adapted to respond to changing market conditions, experience feedback and technological development by a continuous improvement process of design evolution. The CANDU 6 Nuclear Power Plant design is a successful family of nuclear units, with the first four units entering service in 1983, and the most recent entering service this year. A further four CANDU 6 units are under construction. Starting in 1996, a focused forward-looking development program is under way at AECL to incorporate a series of individual improvements and integrate them into the CANDU 6, leading to the evolutionary development of the next-generation enhanced CANDU 6. The CANDU 6 improvements program includes all aspects of an NPP project, including engineering tools improvements, design for improved constructability, scheduling for faster, more streamlined commissioning, and improved operating performance. This enhanced CANDU 6 product will combine the benefits of design provenness (drawing on the more than 70 reactor-years experience of the seven operating CANDU 6 units), with the advantages of an evolutionary next-generation design. Features of the enhanced CANDU 6 design include: Advanced Human Machine Interface - built around the Advanced CANDU Control Centre; Advanced fuel design - using the newly demonstrated CANFLEX fuel bundle; Improved Efficiency based on improved utilization of waste heat; Streamlined System Design - including simplifications to improve performance and safety system reliability; Advanced Engineering Tools, -- featuring linked electronic databases from 3D CADDS, equipment specification and material management; Advanced Construction Techniques - based on open top equipment installation and the use of small skid mounted modules; Options defined for Passive Heat Sink capability and low-enrichment core optimization. (author)

  12. CANDU 6 steam line break analysis with CATHENA

    International Nuclear Information System (INIS)

    Detailed thermalhydraulic simulations for CANDU 6 steam line break inside containment are performed to predict the response of the primary and secondary circuits. The analysis is performed using the thermalhydraulic computer code, CATHENA, with a coupled primary and secondary circuit model. A two-loop representation of the primary and secondary circuits is modelled. The secondary circuit model includes the feedwater line from the deaerator storage tank, multi-node steam generators and the steam line up to the turbine. Two cases were carried out using different assumptions for the efficiency of the steam separators. Case 1 assumes the efficiency of the steam separators becomes zero when the water level in the steam drum increases to the elevation of primary cyclones, or the outlet flow from the steam generator becomes higher than 150% of normal flow. Case 2 assumes the efficiency becomes zero only when the water level in the steam drum reaches the elevation of primary cyclones. The simulation results show that system responses are sensitive to the assumption for the efficiency of the steam separators and Case 1 gives higher discharge energy. Fuel cooling is assured, since primary circuit is cooled down sufficiently by the steam generators for both cases. 1 ref., 12 figs., 2 tabs

  13. CANDU energy for steam assisted gravity drainage

    International Nuclear Information System (INIS)

    Traditional open-pit mining has been used by industry for many years to remove oil sands from shallow deposits. To increase production capacity, the industry is looking for new technology to exploit bitumen from deep deposits. Among them, SAGD (Steam-Assisted Gravity Drainage) appears to be the most promising approach. It uses steam to remove bitumen from underground reservoirs. Recently, the SAGD recovery process has been put into commercial operation by major oil companies.Atomic Energy Canada Limited has assessed the use of the ACR-1000 as a source of heat and electricity for oil sand extraction and processing. The ACR-1000 design is an evolutionary development of the familiar CANDU technology, adding innovations to enhance economics, operations, and safety margins. The net electrical output from a standard ACR-1000 will be close to 1100 MWe, depending on local cooling water temperature

  14. CANDU-3: Features of next generation CANDU

    International Nuclear Information System (INIS)

    CANDU 3, with a net electrical output of 450 MW, is the latest and smallest version of the CANDU power system. Significant innovation built on proven CANDU reactor technology is the basis of the CANDU 3 design. This, coupled with the commitment to reduce plant cost, increase performance capacity factor, enhance safety features and incorporate technological improvements, makes CANDU 3 an advanced, world class product. This paper describes the following CANDU 3 features: A station layout to provide a flexible construction sequence, good system separation and ease of maintenance and operation. An up-front engineering and licensing process prior to beginning construction. Enhancement and simplification of safety features with extended time scales for response, thus limiting reliance on operator action for accident mitigation. Enhanced design capabilities through the use of the latest Computer Aided Design and Drafting (CADD) technology. A 38 month construction schedule achieved by using modularization and open-top construction and installation techniques. A more passive containment system incorporating a steel liner and eliminating the need for active spray. A grouping and separation philosophy for maximum protection of redundant safety systems. Ease of equipment qualification and maximum protection of critical components. Replacement of centralized control and monitoring computers with a redundant distributed control system and modern plant display system. A consistent, logical approach to control room design founded on human factors, automation and event management. (author). 6 refs, 3 figs

  15. Influence of aqueous environment pH on the corrosion behaviour of the CANDU steam generator tubing material

    International Nuclear Information System (INIS)

    The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism in order to evaluate the amounts of corrosion products which exist in the steam generator after a determined period of operation. The purpose of the experimental research consists in the assessment of corrosion behavior of the tube material (Incoloy-800) at normal secondary circuit parameters (temperature - 260 deg. C, pressure - 5.1 MPa). The testing environment was the demineralized water without impurities, at different pH values regulated with morpholine and cycloheyilamine (all volatile treatment). The results are presented as micrographs and graphics representing loss of metal by corrosion, corrosion rate, the total corrosion products, the adherent corrosion product, the released corrosion products and the release of the metal. (authors)

  16. Regulation of ageing steam generators

    International Nuclear Information System (INIS)

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

  17. 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.

  18. Steam generator tube failures

    International Nuclear Information System (INIS)

    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

  19. A comparison of the passive oxide films formed on CANDU steam generator tubing alloy 600 and alloy 800

    International Nuclear Information System (INIS)

    Alloy 600 (A600) steam generator (SG) tubing has been shown to be susceptible to stress corrosion cracking (SCC). Alloy 800 (A800) was developed as a replacement, though it has shown susceptibility to corrosion under certain conditions. The properties of the passive oxide films on both alloys were extensively analyzed to determine why the performance of A800 is superior to that of A600. Surface analysis to determine oxide composition was performed using X-ray photoelectron spectroscopy (XPS) and Auger Electron spectroscopy (AES). Electrochemical measurements were made using anodic polarization and electrochemical impedance spectroscopy (EIS). The oxide films on A600 and A800 were shown to have different electrochemical and compositional properties. (author)

  20. Replacement steam generators for pressurized water reactors

    International Nuclear Information System (INIS)

    Babcock and Wilcox Canada has developed an Advanced Series steam generator for PWR Systems. This design incorporates all of the features that have contributed to the successful CANDU steam generator performance. This paper presents an overview of the design features and how the overall design relates to the requirements of a PWR reactor system

  1. Thermal-hydraulics in recirculating steam generators

    International Nuclear Information System (INIS)

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

  2. The third generation CANDU control room

    International Nuclear Information System (INIS)

    In CANDU stations, as in most complex industrial plants, the man/machine interface design has progressed through three generations. First Generation control rooms consisted entirely on fixed, discrete components (handswitches, indicator lights, strip chart, recorder, annunciator windows, etc.). Human factors input was based on intuitive common sense factors which varied considerably from one designer to another. Second Generation control rooms incorporated video display units and keyboards in the control panels. Computer information processing and display are utilized. There is systematic application of human factors through ergonomic and anthropometric standards and cookbooks. The human factors are applied mainly to the physical layout of the control panels and the physical manipulation performed by the operators. Third Generation control rooms exploit the dramatic performance/cost improvements in computer, electronic display and communication technologies of the 1980's. Further applications of human factors address the cognitive aspects of operator performance. At AECL, second generation control rooms were installed on CANDU stations designed in the mid 70s and early 80s. Third generation features will be incorporated in the CANDU 3 station design and future CANDU stations. There have been significant improvements in the man/machine interface in CANDU stations over the past three decades. The continuing rapid technological developments in computers and electronics coupled with an increasing understanding and application of human factors principles is leading to further enhancements. This paper outlines progress achieved in earlier stations and highlights the features of the CANDU 3rd generation control room. (author). 13 refs, 5 figs

  3. The next generation of CANDU technologies: profiling the potential for hydrogen fuel

    International Nuclear Information System (INIS)

    This report discusses the Next-generation CANDU Power Reactor technologies currently under development at AECL. The innovations introduced into proven CANDU technologies include a compact reactor core design, which reduces the size by a factor of one third for the same power output; improved thermal efficiency through higher-pressure steam turbines; reduced use of heavy water (one quarter of the heavy water required for existing plants), thus reducing the cost and eliminating many material handling concerns; use of slightly enriched uranium to extend fuel life to three times that of existing natural uranium fuel and additions to CANDU's inherent passive safety. With these advanced features, the capital cost of constructing the plant can be reduced by up to 40 per cent compared to existing designs. The clean, affordable CANDU-generated electricity can be used to produce hydrogen for fuel cells for the transportation sector, thereby reducing emissions from the transportation sector

  4. Candu technology: the next generation now

    International Nuclear Information System (INIS)

    We describe the development philosophy, direction and concepts that are being utilized by AECL to refine the CANDU reactor to meet the needs of current and future competitive energy markets. The technology development path for CANDU reactors is based on the optimization of the pressure tube concept. Because of the inherent modularity and flexibility of this basis for the core design, it is possible to provide a seamless and continuous evolution of the reactor design and performance. There is no need for a drastic shift in concept, in technology or in fuel. By continual refinement of the flow and materials conditions in the channels, the basic reactor can be thermally and operationally efficient, highly competitive and economic, and highly flexible in application. Thus, the design can build on the successful construction and operating experience of the existing plants, and no step changes in development direction are needed. This approach minimizes investor, operator and development risk but still provides technological, safety and performance advances. In today's world energy markets, major drivers for the technology development are: (a) reduced capital cost; (b) improved operation; (c) enhanced safety; and (d) fuel cycle flexibility. The drivers provide specific numerical targets. Meeting these drivers ensures that the concept meets and exceeds the customer economic, performance, safety and resource use goals and requirements, including the suitable national and international standards. This logical development of the CANDU concept leads naturally to the 'Next Generation' of CANDU reactors. The major features under development include an optimized lattice for SEU (slightly enriched uranium) fuel, light water cooling coupled with heavy water moderation, advanced fuel channels and CANFLEX fuel, optimization of plant performance, enhanced thermal and BOP (balance of plant) efficiency, and the adoption of layout and construction technology adapted from successful on

  5. Nuclear steam generator

    International Nuclear Information System (INIS)

    A nuclear steam generator has a blowdown pump arranged to pump water from the blowdown line through a filter for return to the steam generator. The piping is arranged so that the pump may operate to reverse the direction of pumping through the blowdown line whereby reverse circulation may be established during wet lay up of the steam generator. A blower is arranged to withdraw nitrogen from an upper elevation in the steam generator and inject the nitrogen into the blowdown line in combination with the pumped reverse circulation during wet lay up. (author)

  6. Advancing the CANDU reactor: From generation to generation

    International Nuclear Information System (INIS)

    Emphasizing safety, reliability and economics, the CANDU reactor development strategy is one of continuous improvement, offering value and assured support to customers worldwide. The Advanced CANDU Reactor (ACR-1000) generation, designed by Atomic Energy of Canada Limited (AECL), meets the new economic expectation for low-cost power generation with high capacity factors. The ACR is designed to meet customer needs for reduced capital cost, shorter construction schedule, high plant capacity factor, low operating cost, increased operating life, simple component replacement, enhanced safety features, and low environmental impact. The ACR-1000 design evolved from the internationally successful medium-sized pressure tube reactor (PTR) CANDU 6 and incorporates operational feedback from eight utilities that operate 31 CANDU units. This technical paper provides a brief description of the main features of the ACR-1000, and its major role in the development path of the generations of the pressure tube reactor concept. The motivation, philosophy and design approach being taken for future generation of CANDU pressure tube reactors are described

  7. Replacement nuclear steam generators

    International Nuclear Information System (INIS)

    This paper reviews past and current practices in the replacement of nuclear steam generators. Plants where steam generator replacement has occurred are reviewed to see what changes have been made, and how the evolving technology has significantly reduced outage time and man-rem exposures. Current preferences in design and material are reviewed. 3 refs., 3 tabs., 2 figs

  8. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    International Nuclear Information System (INIS)

    The CANDU reactor is the most versatile commercial power reactor in the world. The acronym 'CANDU', a registered trademark of Atomic Energy of Canada Limited, stands for 'CANada Deuterium Uranium'. CANDU uses heavy water as moderator and uranium (originally, natural uranium) as fuel. All current power reactors in Canada are of the CANDU type. Canada exports CANDU type reactor in abroad. CANDU type is used as the nuclear power plants to produce electrical. Today, there are 41 CANDU reactors in use around the world, and the design has continuously evolved to maintain into unique technology and performance. The CANDU-6 power reactor offers a combination of proven, superior and state-of-the-art technology. CANDU-6 was designed specifically for electricity production, unlike other major reactor types. One of its characteristics is a very high operating and fuel efficiency. Canada Nuclear Power Generating Stations were succeeded in a commercial reactor of which the successful application of heavy water reactor, natural uranium method and that on-power fuelling could be achieved. It was achieved through the joint development of a major project by strong support of the federal government, public utilities and private enterprises. The potential for customization to any country's needs, with competitive development and within any level of domestic industrial infrastructure, gives CANDU technology strategic importance in the 21st century

  9. CANDU co-generation opportunities

    International Nuclear Information System (INIS)

    Modern technology makes use of natural energy 'wealth' (uranium) to produce useful energy 'currency' (electricity) that can be used to society's benefit. This energy currency can be further applied to help solve a difficult problem faced by mankind. Within the next few years we must reduce our use of the same fuels which have made many countries wealthy - fossil fuels. Fortunately, electricity can be called upon to produce another currency, namely hydrogen, which has some distinct advantages. Unlike electricity, hydrogen can be stored and can be recovered for later use as fuel. It also is extremely useful in chemical processes and refining. To achieve the objective of reducing greenhouse gas emissions hydrogen must, of course, be produced using a method which does not emit such gases. This paper summarizes four larger studies carried out in Canada in the past few years. From these results we conclude that there are several significant opportunities to use nuclear fission for various co-generation technologies that can lead to more appropriate use of energy resources and to reduced emissions. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  12. French steam generator

    International Nuclear Information System (INIS)

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

  13. Liquid metal steam generator

    International Nuclear Information System (INIS)

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

  14. Consolidated nuclear steam generator

    International Nuclear Information System (INIS)

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

  15. Consolidated nuclear steam generator

    International Nuclear Information System (INIS)

    The invention described relates to an improved system of providing power having a unique generating means of the nuclear reactor variety adapted with a plurality of steam generators in the form of replaceable modular units of the expendable type for the attainment of the optimum in effective and efficient vaporization of fluid during the process of generating power

  16. Trends in the capital costs of CANDU generating stations

    International Nuclear Information System (INIS)

    This paper consolidates the actual cost experience gained by Atomic Energy of Canada Limited, Ontario Hydro, and other Canadian electric utlities in the planning, design and construction of CANDU-PHWR (CANada Deuterium Uranium-Pressurized Heavy Water Reactor) generating stations over the past 30 years. For each of the major CANDU-PHWR generating stations in operation and under construction in Canada, an analysis is made to trace the evolution of the capital cost estimates. Major technical, economic and other parameters that affect the cost trends of CANDU-PHWR generating stations are identified and their impacts assessed. An analysis of the real cost of CANDU generating stations is made by eliminating interest during construction and escalation, and the effects of planned deferment of in-service dates. An historical trend in the increase in the real cost of CANDU power plants is established. Based on the cost experience gained in the design and construction of CANDU-PHWR units in Canada, as well as on the assessment of parameters that influence the costs of such projects, the future costs of CANDU-PHWRs are presented

  17. Key thrusts in next generation CANDU. Annex 10

    International Nuclear Information System (INIS)

    Current electricity markets and the competitiveness of other generation options such as CCGT have influenced the directions of future nuclear generation. The next generation CANDU has used its key characteristics as the basis to leap frog into a new design featuring improved economics, enhanced passive safety, enhanced operability and demonstrated fuel cycle flexibility. Many enabling technologies spinning of current CANDU design features are used in the next generation design. Some of these technologies have been developed in support of existing plants and near term designs while others will need to be developed and tested. This paper will discuss the key principles driving the next generation CANDU design and the fuel cycle flexibility of the CANDU system which provide synergism with the PWR fuel cycle. (author)

  18. The next generation of CANDU reactor: evolutionary economics

    International Nuclear Information System (INIS)

    AECL has developed the design for a next generation of CANDUR plants by applying a set of enabling technologies to well-established successful CANDU features from the CANDU 6 Reactors in service and the design of the CANDU 9. Advances made in the construction of the Wolsong reactors have been built upon in the current project in China. The basis for the new design is to evolve from the current CANDU units by replicating or adapting existing components for a new core design. Using slightly enriched uranium fuel, a core with light water coolant, and heavy water moderator and reflector has been defined, based on the existing CANDU fuel channel module. This paper summarizes the main features and characteristics of the reference next-generation CANDU design. The progress of the next generation of CANDU design program in meeting challenging cost, schedule and performance targets is described. AECL's cost reduction methodology is summarized as an integral part of the design optimization process. Examples of cost reduction features are given, together with enhancement of design margins

  19. Patricia steam generator tests

    International Nuclear Information System (INIS)

    The Patricia GV program is a joint program of CEA, French utilities, and FRAMATOME. It aims at a better knowledge of the pressurized water reactor steam generator behavior under small break loss of coolant accident conditions. It has two parts: Patricia GVl deals with the primary side of the steam generator and is presently completed. Patricia GV2 deals with the secondary side dry-out phenomenon and is still in progress. Objectives, test facilities, test procedures, and results of the program are discussed

  20. Steam generator sludge removal apparatus

    International Nuclear Information System (INIS)

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

  1. Steam generators - problems and prognosis

    Energy Technology Data Exchange (ETDEWEB)

    Tapping, R.L

    1997-05-01

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

  2. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

    AECL has developed a suite of technologies for CanduR reactors that enable the next step in the evolution of the Candu family of heavy-water-moderated fuel-channel reactors. These technologies have been combined in the design for the Advanced Candu Reactor TM1 (ACRTM), AECL's next generation Candu power plant. The ACR design builds extensively on the existing Candu experience base, but includes innovations, in design and in delivery technology, that provide very substantial reductions in capital cost and in project schedules. In this paper, main features of next generation design and delivery are summarized, to provide the background basis for the cost and schedule reductions that have been achieved. In particular the paper outlines the impact of the innovative design steps for ACR: - Selection of slightly enriched fuel bundle design; - Use of light water coolant in place of traditional Candu heavy water coolant; - Compact core design with unique reactor physics benefits; - Optimized coolant and turbine system conditions. In addition to the direct cost benefits arising from efficiency improvement, and from the reduction in heavy water, the next generation Candu configuration results in numerous additional indirect cost benefits, including: - Reduction in number and complexity of reactivity mechanisms; - Reduction in number of heavy water auxiliary systems; - Simplification in heat transport and its support systems; - Simplified human-machine interface. The paper also describes the ACR approach to design for constructability. The application of module assembly and open-top construction techniques, based on Candu and other worldwide experience, has been proven to generate savings in both schedule durations and overall project cost, by reducing premium on-site activities, and by improving efficiency of system and subsystem assembly. AECL's up-to-date experience in the use of 3-D CADDS and related engineering tools has also been proven to reduce both engineering and

  3. Steam generator life management

    International Nuclear Information System (INIS)

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

  4. Steam generator tube performance

    International Nuclear Information System (INIS)

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

  5. CANDU 9 - Overview

    International Nuclear Information System (INIS)

    The CANDU 9 plants are single unit versions of the very successful four unit Bruce B design, incorporating relevant technical advances made in the CANDU 6 and the newer Dalington and CANDU 3 designs. The CANDU 9 plant described in this paper is the CANDU 9 480/SEU with a net electrical output in the range of 1050 MW. In this designation 480 refers to the number of fuel channels, and SEU refers to slightly enriched uranium. Emphasis is placed on evolutionary design and the use of well-proven design features to ensure minimum financial risk to utilities choosing a CANDU 9 plant by assuring regulatory licensability and reliable operation. In addition, the CANDU 9 power plants reflect the important lessons learned by utilities in the construction and operation of CANDU units and, indeed, relevant experience gained by the world nuclear community in its operation of over 400 reactors of a variety of types. As a results, the CANDU 9 plants offer a high level of investment security to the owner, together with relatively low energy costs. The latter results from reduced specific capital cost, reduced operation and maintenance cost, and reduced radiation exposure to plant staff. A high level of standardization has always been a feature of CANDU reactors. This theme is emphasized in the CANDU 9 plants; all key components (steam generators, heat transport pumps, pressure tubes, fuelling machines, etc.) are of the same design as those proven in-service on operating CANDU power stations. The CANDU 9 power plants are readily adaptable to the individual requirements of different utilities and are suitable for a range of site conditions. (author). 12 figs

  6. Future generations of CANDU: advantages and development with passive safety

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) advances water reactor and CANDLT technology using an evolutionary development strategy. This strategy ensures that innovations are based firmly on current experience and keeps our development programs focused on one reactor concept, reducing risks, development costs, and product development cycle times. It also assures our customers that our products will never become obsolete or unsupported, and the continuous line of water reactor development is secure and supported into the future. Using the channel reactor advantage of modularity, the subdivided core has the advantage of passive safety by heat removal to the low- pressure moderator. With continuous improvements, the Advanced CANDU Reactor TM (ACR-1000TM) concept will likely remain highly competitive for a number of years and leads naturally to the next phase of CANDU development, namely the Generation IV CANDU -SCWR concept. This is conventional water technology, since supercritical boilers and turbines have been operating for some time in coal-fired power plants. Significant cost, safety, and performance advantages would result from the CANDU-SCWR concept, plus the flexibility of a range of plant sizes suitable for both small and large electric grids, and the ability for co-generation of electric power, process heat, and hydrogen. In CANDU-SCWR, novel developments are included in the primary circuit layout and channel design. The R and D in Canada is integrated with the Generation IV international Forum (GIF) plans, and has started on examining replaceable insulating liners that would ensure channel life, and on providing completely passive reactor decay heat removal directly to the moderator heat sink without forced cooling. In the interests of sustainability, hydrogen production by a CANDU- SCWR is also be included as part of the system requirements, where the methods for hydrogen production will depend on the outlet temperature of the reactor

  7. The enhanced CANDU 6 reactor - Generation III CANDU medium size global reactor

    International Nuclear Information System (INIS)

    Full text: The Enhanced CANDU 6TM (EC6TM) is a Generation III 700 class, heavy water moderated pressure tube reactor, designed to provide safe, reliable, nuclear power. The EC6TM has evolved from the proven CANDU 6 plants licensed and operating in five countries (four continents) with over 150 reactor years of safe operation around the world. In recent years. this global CANDU 6 fleet, with over 92% average gross capacity factor has ranked in the world's top performing reactors. The EC6 reactor builds on this success of the CANDU 6 fleet by using the operation, experience and project feedback to upgrade the design and construction techniques. A key objective of the EC6 has been to review and incorporate design improvements in the CANDU 6 to meet current safety standards. The key characteristics of the highly successful CANDU 6 reactor design include: - Powered by natural Uranium; - Ease of installation with modular, horizontal fuel channel core; - Separate low-temperature, low-pressure moderator providing inherently passive heat sinks; Reactor vault filled with light water surrounding the core; - Two independent safety shutdown systems; - On-power fuelling; - The CANDU 6 plant has a highly automated control system, with plant control computers that adjust and maintain the reactor power for plant stability (which is particularly beneficial in less developed power grids-where fluctuations occur regularly and capacities are limited). The major improvements incorporated in the EC6 design include, - More robust containment and increased passive features e.g., thicker walls, steel liner; - Enhanced severe accident management with additional emergency heat removal systems; - Improved shutdown performance for improved Large LOCA margins; - Upgraded fire protection systems to meet current Canadian and International standards; - Additional design features to improve environmental protection for workers and public- ALARA principle; - Automated and unitized back-up standby

  8. Steam generator module

    International Nuclear Information System (INIS)

    The module of the steam generator is arranged such that the first working medium flows through the tubes of the heat exchange bundle and the second working medium flows through the intertube space. At least one side of the module is provided with a lid which is provided with a system of through-flow apertures. The apertures are expanded and provided with a thread in the direction of the outer side of the lid. They are coaxial with the tubes of the heat exchange bundle at the point of their anchorage in the tube plate. The apertures are closed with plugs with a male thread and the sealing surfaces are formed between the thread joint and the space of the first working medium. The plugs extend into the space of the heat exchange bundle and form a throttle which replaces the classical stop and allow for dismantling. This arrangement of the modular steam generator allows the control of the inner surfaces of heat exchange pipes and also the cleaning of these inner surfaces. (E.S.)

  9. Steam generator materials

    International Nuclear Information System (INIS)

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

  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. Steam generator blowdown system upgrades

    International Nuclear Information System (INIS)

    The steam generator blowdown (SGBD) system is used to remove impurities from the steam generators in order to maintain steam generator (SG) water chemistry within specifications. The original SGBD systems at Diablo Canyon power plant (DCPP) were designed in the early 1970s, and since that time the industry has changed its practices regarding water chemistry. DCPP has operated its SGBD system above its design flow rate. This resulted in a history of high maintenance and unreliable operation. Subsequently, DCPP implemented extensive modifications in order to accommodate the higher industry standard flow rates. These modifications resulted in a more reliable and rugged system. Additionally, significant savings were realized due to an increase in net plant output and a reduction in the required plant makeup water by recovering steam generator blowdown. (author)

  13. Research program plan: steam generators

    International Nuclear Information System (INIS)

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

  14. Options for Steam Generator Decommissioning

    International Nuclear Information System (INIS)

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

  15. Steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

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

  16. Corrosion of the tube-tubesheet joint in the Candu steam generator at operation resuming after the chemical cleaning of depositions occurred due to the contamination of condense by the cooling water

    International Nuclear Information System (INIS)

    The purpose of this paper is to establish the corrosion behaviour of the Iy-800 (material of the tube) and SA 508 cl.2 (material of the tube sheet) after the chemical cleaning of the steam generator on the secondary circuit specific components. On the surface of these materials there are initial deposits consisting of oxides and corrosion products. The chemical cleaning was made at temperatures between 60-80oC. The substances included in the chemical cleaning solution are citric acid, Na2EDTA, HEDTA, ammonium citrate, hydrazine hydrate and thiourea. After the chemical cleaning the samples were passivated and corrosion-tested at the typical parameters of the steam generator secondary circuit (demineralized water, pH=9.5, all volatile treatment, temperature 265oC and pressure 5.1 MPa). The results of metallographic and X-ray diffraction examinations are presented as well as the reasons for selecting the best solution for the chemical cleaning. (author)

  17. Steamdrive direct contact steam generation for SAGD

    Energy Technology Data Exchange (ETDEWEB)

    Betzer-Zilevitch, Maoz [Society of Petroleum Engineers (Canada)

    2011-07-01

    Recently, focus has been shifted from the traditional once-through steam generator (OTSG) to the use of an evaporator and industrial high-pressure (HP) steam boilers in steam assisted gravity drainage (SAGD) steam generation technologies. As a result, the temperatures and pressures attainable are significantly higher than required for injection. This excess steam enthalpy drives the novel steam generation method presented in this paper. The process generates additional steam from highly contaminated oily water with zero liquid waste discharge using industrial boiler superheated steam. The process generates tailor-made pressure and temperature steam required for injection into the underground oil bearing formation. An additional 8-24% steam can be generated from highly contaminated oily water as is seen from the simulation results. The amount of additional steam generated is directly proportional to the temperature of the driving steam and the decrease in pressure of the formation.

  18. Embalse steam generators - status in 2009

    International Nuclear Information System (INIS)

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

  19. Thermal stability of chloroform in the steam condensate cycle of CANDU-PHW nuclear power plant

    International Nuclear Information System (INIS)

    Analysis of samples taken at the Gentilly 2 (Quebec) CANDU-PHW (CANadian Deuterium Uranium - Pressurized Heavy Water) plant after chlorination and demineralization revealed the presence of all four trihalomethanes (THMs) (CHCl3, CHBrCl2, CHBr2Cl and CHBr3) and other unidentified halogenated volatile compounds. Among the THMs, chloroform was the major contaminant. A study of its thermal stability in water at different temperatures confirmed the degradation of the CHCl3 molecule according to the equation CHCl3 + H2O → CO + 3 HCl. The reaction follows first order kinetics and has an activation energy of 100 kJ/mol. The estimated half-life is six seconds at 260 deg C, the maximum temperature of the steam condensate cycle

  20. Steam generator hand hole shielding.

    Science.gov (United States)

    Cox, W E

    2000-05-01

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

  1. Environmental codes of practice for steam electric power generation

    International Nuclear Information System (INIS)

    The Design Phase Code is one of a series of documents being developed for the steam electric power generation industry. This industry includes fossil-fuelled stations (gas, oil and coal-fired boilers), and nuclear-powered stations (CANDU heavy water reactors). In this document, environmental concerns associated with water-related and solid waste activities of steam electric plants are discussed. Design recommendations are presented that will minimize the detrimental environmental effects of once-through cooling water systems, of wastewaters discharged to surface waters and groundwaters, and of solid waste disposal sites. Recommendations are also presented for the design of water-related monitoring systems and programs. Cost estimates associated with the implementation of these recommendations are included. These technical guides for new or modified steam electric stations are the result to consultation with a federal-provincial-industry task force

  2. AI reference LMFBR steam-generator development

    International Nuclear Information System (INIS)

    The Design Data Sheets summarize the key parameters being used in the design and analysis of the AI Prototype LMFBR Steam Generator. These Data Sheets supplement SDD-097-330-002, Steam Generator System, 1450 psi Steam Conditions. This document will serve as the baseline design data control until a GE/RRD approved steam generator specification with ordering data is received

  3. Steam generators: learning from experience

    International Nuclear Information System (INIS)

    The North Anna power plant in Virginia, USA, comprises two PWR units. Each employs three Westinghouse series 51 steam generators. New lower steam generator assemblies, with 3592 tubes made of thermally treated Alloy 690, were installed in 1993, replacing the original assemblies which contained 3388 tubes of mild annealed Alloy 600. Since then, the operators have been engaged in planning a similar replacement for unit 2. This article examines how lessons learned in the unit 1 replacement are being taken into account. It is hoped to improve performance by 30% in the unit 2 replacement which is scheduled to start in September 1996. (UK)

  4. Primary separator replacement for Bruce Unit 8 steam generators

    International Nuclear Information System (INIS)

    During a scheduled maintenance outage of Bruce Unit 8 in 1998, it was discovered that the majority of the original primary steam separators were damaged in two steam generators. The Bruce B steam generators are equipped with GXP type primary cyclone separators of B and W supply. There were localized perforations in the upper part of the separators and large areas of generalized wall thinning. The degradation was indicative of a flow related erosion corrosion mechanism. Although the unit- restart was justified, it was obvious that corrective actions would be necessary because of the number of separators affected and the extent of the degradation. Repair was not considered to be a practical option and it was decided to replace the separators, as required, in Unit 8 steam generators during an advanced scheduled outage. GXP separators were selected for replacement to minimize the impact on steam generator operating characteristics and analysis. The material of construction was upgraded from the original carbon steel to stainless steel to maximize the assurance of full life. The replacement of the separators was a first of a kind operation not only for Ontario Power Generation and B and W but also for all CANDU plants. The paper describes the degradations observed and the assessments, the operating experience, manufacture and installation of the replacement separators. During routine inspection in 1998, many of the primary steam separators in two of steam generators at Bruce Nuclear Division B Unit 8 were observed to have through wall perforations. This paper describes assessment of this condition. It also discusses the manufacture and testing of replacement primary steam separators and the development and execution of the replacement separator installation program. (author)

  5. CANDU 300

    International Nuclear Information System (INIS)

    The CANDU nuclear power system is under continuous review by AECL in order to advance the CANDU concept in a manner that will assure competitiveness in both current and future markets. Over the past three years development effort has featured the CANDU 300, a CANDU nuclear generating station with a net output in the range of 320 MW9e) to 380 MW(e). At the outset AECL recognized that coal-fired power plants would be the primary competition for the CANDU advantages such as the use of natural uranium fuel and on-power refuelling, while enhancing capacity factor, reducing man-rem exposure, reducing capital cost, and minimizing construction schedules. AECL believes that the resulting CANDU 300 nuclear generating station will have substantial appeal to many utilities, in both developed and developing countries. The key features of the CANDU 300 are presented here, with particular attention to the station layout, construction methods, and construction schedules

  6. Tracking of Steam Generator Thermal Performance Trends

    International Nuclear Information System (INIS)

    A significant number of pressurized water reactor (PWR) plants have reported decreases in their steam generator steam pressure during the last several years. Because a steam pressure decrease causes a reduction of the electrical power generating capacity directly, a steam generator's thermal performance is one of the important issues for steam generator maintenance. Therefore, the Korea Hydro and Nuclear Power Company has established an on-line acquisition system for plant operational parameters as a part of the Steam Generator Management Program (SGMP). Recently, plant-specific tools for calculating the overall heat transfer coefficient and the global fouling factor were also constructed and applied to some plants

  7. Steam generator waterlancing at DNGS

    International Nuclear Information System (INIS)

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

  8. Steam generator replacement in Bruce A Unit 1 and Unit 2

    International Nuclear Information System (INIS)

    The Bruce A Generating Station consists of four 900 MW class CANDU units. The reactor and Primary Heat Transport System for each Unit are housed within a reinforced concrete reactor vault. A large duct running below the reactor vaults accommodates the shared fuel handling system, and connects the four reactor vaults to the vacuum building. The reactor vaults, fuelling system duct and the vacuum building constitute the station vacuum containment system. Bruce A Unit 2 was shut down in 1995 and Bruce A Units 1, 3 and 4 were shutdown in 1997. Bruce A Units 3 and 4 were returned to service in late 2003 and are currently operating. Units 1 and 2 remain out of service. Bruce Power is currently undertaking a major rehabilitation of Bruce A Unit 1 and Units 2 that will extend the in-service life of these units by at least 25 years. Replacement of the Steam Generators (eight in each unit) is required; this work was awarded to SNC-Lavalin Nuclear (SLN). The existing steam drums (which house the steam separation and drying equipment) will be retained. Unit 2 is scheduled to be synchronized with the grid in 2009, followed by Unit 1 in 2009. Each Bruce A unit has two steam generating assemblies, one located above and to each end of the reactor. Each steam generating assembly consists of a horizontal cylindrical steam drum and four vertical Steam Generators. The vertical Steam Generators connect to individual nozzles that are located on the underside of the Steam Drum (SD). The steam drums are located in concrete shielding structures (steam drum enclosures). The lower sections of the Steam Generators penetrate the top of the reactor vaults: the containment pressure boundary is established by bellows assemblies that connect between the reactor vault roof slab and the Steam Generators. Each Steam Generators is supported from the bottom by a trapeze that is suspended from the reactor vault top structure. The Steam Generator Replacement (SGR) methodology developed by SLN for Unit 1

  9. Steam generator specification design transients

    Energy Technology Data Exchange (ETDEWEB)

    Stevenson, D.H.

    1978-05-12

    This LTR documents LOFT Plant operational, upset, and emergency transients, and pressure and temperature conditions as generated by the LOFT Plant Dynamic analysis model of the primary system for use in steam generator specification. The results of this LTR have been supplemented by succeeding efforts as follows: FSAR analysis - (a) loss of load during full power operation, (b) loss of primary pump electrical power, and (c) loss of site power; and LOFT maneuvering analysis - (a) +-10% step change in steam flow, and (b) manual reactor trip. The initial plant conditions considered may differ from this LTR plant conditions. Items are outlined to document in this LTR other analyses which cover the same plant transients. 53 figs., 1 tab.

  10. The CANDU-PHW generating system waste arisings

    International Nuclear Information System (INIS)

    In this report, the volume of material and level of contained radioactive nuclides are tabulated for wastes arising from four fuel cycles which might be operated in CANDU-PHW (CANada Deuterium Uranium - Pressurized Heavy Water) reactors. The data presented, based on Canadian experience and/or studies, cover the range of conditioned waste volumes which could be expected from steady-state (no growth), CANDU-PHW-powered electrical generating systems. The wastes arising from operation and decommissioning of facilities in each phase of each fuel cycle are estimated. Each fuel cycle is considered to operate in isolation with the data given in terms of quantities per gigawatt-year of electricity produced. Three of the fuel cycles for which data are presented, the natural uranium once-through cycle, the plutonium-enriched uranium cycle (plutonium recycle) and the low-burnup uranium-enriched thorium cycle (thorium and uranium recycle), were studied by INFCE WG.7 (the International Nuclear Fuel Cycle Evaluation, Working Group 7) as fuel cycles 4, 5 and 6. The high-burnup uranium-enriched thorium cycle is included for comparison. INFCE WG.7 selected many common reference parameters which are applied uniformly to all seven INFCE WG.7 reference fuel cycles in determining waste arisings. Where these parameters differ from the data of Canadian origin given in the body of this report, the INFCE WG.7 data are given in an appendix. The waste management costs associated with operation of each INFCE WG.7 reference fuel cycle were calculated and compared by the working group. An arbitrary set of costing parameters and disposal technologies was selected by the working group for application to each of the reference fuel cycles. The waste management and disposal costs for the PHW reactor fuel cycles based on these arbitrary cost parameters are given in an appendix. (author)

  11. Operating the Gentilly-2 steam generating system without hydrazine

    International Nuclear Information System (INIS)

    The Gentilly-2 plant is a CANDU 600 operating in the province of Quebec. For the past two years the feedwater and steam generators have been operating with All Volatile Treatment using morpholine and have operated without hydrazine addition. This paper reviews the operating experience of the Gentilly 2 feedwater system and the steam generators. The Gentilly-2 steam and feedwater system has an all-ferrous feedtrain which includes a deaerator but has admiralty brass condenser tubes. The operating staff became concerned about decomposition of the hydrazine to ammonia and attack of ammonia on the condenser tubes. As a result of this the operating staff decided to halt hydrazine addition and to determine if the system experienced any problem due to the lack of hydrazine. After two years of operation the system continues to operate within the specified chemistry limits. In addition, inspection of the system shows that there has been no increased corrosion product transport and that the deaerator storage tank and the steam generators have not had any significant crud buildup. This experience indicates that an all-ferrous system may operate without hydrazine addition without experiencing deleterious effects. However, this does not imply that hydrazine addition is without benefit. The system performance will continue to be monitored and attempts will be made to determine whether the corrosion products being produced at Gentilly-2 are significantly different from other plants operating with hydrazine

  12. Evaluation on operation of liquid relief valves for steam line break accidents by RELAP5/CANDU+ code

    International Nuclear Information System (INIS)

    A development of RELAP5/CANDU+ code for regulatory audits of accident analysis of CANDU nuclear power plants is on progress. This paper is undertaken in a procedure of a verification and validation for RELAP5/CANDU+ code by analyzing main steam line break accidents of WS 2/3/4. Following the ECC injection in sequence of the steam line breaks, the mismatch in heat transfer between the primary and the secondary systems makes pressure of the primary system instantly peaked to the open setpoint of liquid relief valves. The event sequence follows the result of WS 2/3/4 FSAR, but there is a difference in pressure transient after ECC injection. Sensitivity analysis for main factors dependent on the peak pressure such as control logics of liquid relief valves. ECC flow path and feedwater flow is performed. Because the pressure increase is continued for a long time and its peaking is high, open and close of the liquid relief valves are repeated several times, which is obviously different from those of WS 2/3/4 FSAR. As a result, it is evaluated that conservative modeling for the above variables is required in the analysis

  13. 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.

  14. International examples of steam generator replacement

    International Nuclear Information System (INIS)

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

  15. Operating experience with steam generators

    International Nuclear Information System (INIS)

    The Belgian utilities operate 7 PWR units, the steam generators of which suffer from different corrosion attacks. While PWSCC at the roll transition has long been the major difficulty, degradations of the external surface of the tubes were recently observed in different units, at the level of the top of tubesheet, at the tube support plates and in the sludge piles. Many of the observed cracks are through wall but do not reduce excessively the tube strength, what led to the development of specific plugging criteria, thus allowing most of the affected tubes to be kept in service. The Belgian utilities have thus learned to operate imperfectly tight steam generators. They have improved the procedures for in-service leak monitoring and for detection of leaking tubes during outages, as well as the accuracy and efficiency of NDE tools. Many repair interventions were carried out, several of which were tests essentially aimed at assessing new techniques. In spite of the corrosion defects affecting the Belgian steam generators, good operating records in safe conditions have been achieved thanks to extensive R and D in NDE, innovative plugging criteria and large in-situ tests of repair techniques. The major past concern was PWSCC at roll transition. In Doel 3 this has led to a replacement scheduled for 1993 because the expected repair costs, added to the production losses exceed the replacement costs. Recently, OD corrosion appeared to be another major threat. The future of the affected units will depend on the progression rate of these new defects, presently under assessment

  16. 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.

  17. Strategic management of steam generators

    International Nuclear Information System (INIS)

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

  18. The Enhanced CANDU 6TM Reactor - Generation III CANDU Medium Size Global Reactor

    International Nuclear Information System (INIS)

    The Enhanced CANDU 6TM (EC6TM) is a 740 MWe class heavy water moderated pressure tube reactor, designed to provide safe, reliable, nuclear power. The EC6TM has evolved from the proven eleven (11) CANDU 6 plants licensed and operating in five countries (four continents) with over 150 reactor years of safe operation around the world. In recent years, this global CANDU 6 fleet has ranked in the world's top performing reactors. The EC6 reactor builds on this success of the CANDU 6 fleet by using the operation, experience and project feedback to upgrade the design and incorporate design improvements to meet current safety standards.The key characteristics of the highly successful CANDU 6 reactor design include: Powered by natural Uranium; Ease of installation with modular, horizontal fuel channel core; Separate low-temperature, low-pressure moderator providing inherently passive heat sinks; Reactor vault filled with light water surrounding the core; Two independent safety shutdown systems; On-power fuelling; The CANDU 6 plant has a highly automated control system, with plant control computers that adjust and maintain the reactor power for plant stability (which is particularly beneficial in less developed power grids-where fluctuations occur regularly and capacities are limited). The major improvements incorporated in the EC6 design include: More robust containment and increased passive features e.g., thicker walls, steel liner; Enhanced severe accident management with additional emergency heat removal systems; Improved shutdown performance for improved Large LOCA margins; Upgraded fire protection systems to meet current Canadian and International standards; Additional design features to improve environmental protection for workers and public-ALARA principle; Automated and unitized back-up standby power and water systems; Other improvements to meet higher safety goals consistent with Canadian and International standards based on PSA studies; Additional reactor trip

  19. 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.

  20. Innovations relative to steam generators

    International Nuclear Information System (INIS)

    In the last decade the main object of attention in nuclear engineering has been that of safety; safety understood fundamentally as a study and examination of the possible consequences of accidents and of the devices for and phases of automatic protective intervention. Another problem of safety, that which concerns the criteria aimed at a less complex construction with advantages for transport, setting up, management, maintenance and decomissioning, seems, instead, to be ignored. The use of less specialised workshops for construction, easier control of the state of the structures and the possibility of substituing components during the life of the plant are factors with a direct influence on safety. These aspects, mainly of a creative engineering nature, are the concern of the MARS (Multipurpose Advanced Reactor inherently Safe) project. This memo concerns the innovations introduced by the project relative to the steam generator which is being realised by means of the assembly in situ of 5 sub-components of considerably reduced dimensions and weight with respect to traditional methods of uni-block construction. The economic-management benefits appear significant. Added to the proposal is a brief study for the removal and substitution of the tubing of the steam generator inside the reactor building

  1. Steam generator tubing NDE performance

    International Nuclear Information System (INIS)

    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

  2. Saturated steam process with direct steam generating parabolic troughs

    Energy Technology Data Exchange (ETDEWEB)

    Eck, M. [DLR, Institute of Technical Thermodynamics, Pfaffenwaldring 38-40, 70569 Stuttgart (Germany); Zarza, E. [CIEMAT, Plataforma Solar de Almeria (PSA), P.O. Box 22, 04200 Tabernas (Almeria) (Spain)

    2006-11-15

    The direct steam generation (DSG) in parabolic trough collectors is an attractive option regarding the economic improvement of parabolic trough technology for solar thermal electricity generation in the multi Megawatt range. The European DISS project has proven the feasibility of the direct steam generation under real solar conditions in more than 4000 operation hours. Within the European R and D project INDITEP the detailed engineering for a pre-commercial DSG solar thermal power plant with an electrical power of 5MW is being performed. This small capacity was chosen to minimise the risk for potential investors. In regards to DSG solar thermal power plants, only steam cycles using superheated steam have been investigated so far. The paper will investigate the advantages, disadvantages, and design considerations of a steam cycle operated with saturated steam for the first time. For near term applications, saturated steam operated DSG plants might be an interesting alternative for power generation in the small capacity range due to some specific advantages: Simple set up of the collector field; Proven safe collector field operation; Higher thermal efficiency in the collector field. (author)

  3. Passive safety features for next generation CANDU power plants

    International Nuclear Information System (INIS)

    CANDU offers an evolutionary approach to simpler and safer reactors. The CANDU 3, an advanced CANDU, currently in the detailed design stage, offers significant improvements in the areas of safety, design simplicity, constructibility, operability, maintainability, schedule and cost. These are being accomplished by retaining all of the well known CANDU benefits, and by relying on the use of proven components and technologies. A major safety benefit of CANDU is the moderator system which is separate from the coolant. The presence of a cold moderator reduces the consequences arising from a LOCA or loss of heat sink event. In existing CANDU plants even the severe accident - LOCA with failure of the emergency core cooling system - is a design basis event. Further advances toward a simpler and more passively safe reactor will be made using the same evolutionary approach. Building on the strength of the moderator system to mitigate against severe accidents, a passive moderator cooling system, depending only on the law of gravity to perform its function, will be the next step of development. AECL is currently investigating a number of other features that could be incorporated in future evolutionary CANDU designs to enhance protection against accidents, and to limit off-site consequences to an acceptable level, for even the worst event. The additional features being investigated include passive decay heat removal from the heat transport system, a simpler emergency core cooling system and a containment pressure suppression/venting capability for beyond design basis events. Central to these passive decay heat removal schemes is the availability of a short-term heat sink to provide a decay heat removal capability of at least three days, without any station services. Preliminary results from these investigations confirm the feasibility of these schemes. (author)

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

    International Nuclear Information System (INIS)

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

  5. Probabilistic analysis and interpretation of steam generator tube pitting data

    International Nuclear Information System (INIS)

    'Full text:' Pitting corrosion is a serious form of degradation in steam generator tubing of some CANDU reactors. The occurrence and growth of pits over time are essential inputs to a life-cycle management model. Periodic inspections are carried out to detect and measure corrosion pits in steam generator tubing. However, probabilistic analysis of pit data is not a straightforward task. The reasons are the random measurement error associated with pit sizing, detection uncertainty, and censoring of pit sizes which complicate the estimation of growth rates. Tracking of pits over time to estimate the growth rate is limited by the ability to determine the exact location and elevation of pits between consecutive outages. This paper presents a comprehensive statistical analysis of the pitting data that accounts for limitations of inspection data. The pitting process is modeled as a random process and calibrated with the available data. The model is applied to estimate the distribution of the number extreme pits and probability of tube leakage per SG per EFPY. The application of this model is illustrated for the life-cycle management of steam generators. (author)

  6. Design of PFBR steam generator

    International Nuclear Information System (INIS)

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

  7. The CANDU contribution to environmentally friendly energy production

    International Nuclear Information System (INIS)

    National prosperity is based on the availability of affordable, energy supply. However, this need is tempered by a complementary desire that the energy production and utilization will not have a major impact on the environment. The CANDU energy system, including a next generation of CANDU designs, is a major primary energy supply option that can be an important part of an energy mix to meet Canadian needs. CANDU nuclear power plants produce energy in the form of medium pressure steam. The advanced version of the CANDU design can be delivered in unit modules ranging from 400 to 1200 MWe. This Next Generation of CANDU designs features lower cost, coupled with robust safety margins. Normally this steam is used to drive a turbine and produce electricity. However, a fraction of this steam (large or small) may alternatively be used as process steam for industrial consumption. Options for such steam utilization include seawater desalination, oil sands extraction and heating. The electricity may be delivered to an electrical grid or alternatively used to produce quantities of hydrogen. Hydrogen is an ideal clean transportation fuel because its use only produces water. Thus, a combination of CANDU generated electricity and hydrogen distribution for vehicles is an available, cost-effective route to dramatically reduce emissions from the transportation sector. The CANDU energy system contributes to environmental protection and the prevention of climate change because of its very low emission. The CANDU energy system does not produce any NOx, SOx or greenhouse gas (notably CO2) emissions during operation. In addition, the CANDU system operates on a fully closed cycle with all wastes and emissions fully monitored, controlled and managed throughout the entire life cycle of the plant. The CANDU energy system is an environmentally friendly and flexible energy source. It can be an effective component of a total energy supply package, consistent with Canadian and global climate

  8. Saturated steam process with direct steam generating parabolic troughs

    Energy Technology Data Exchange (ETDEWEB)

    Eck, M. [Deutsches Zentrum fuer Luft- und Raumfahrt, Stuttgart (Germany). Inst. of Technical Thermodynamics; Zarza, E. [CIEMAT, Plataforma Solar de Almeria (PSA), Tabernas (Spain)

    2004-07-01

    The direct steam generation (DSG) in parabolic trough collectors is an attractive option regarding the economic improvement of parabolic trough technology for solar thermal electricity generation in the multi Megawatt range. The European DISS project has proven the feasibility of the direct steam generation under real solar conditions in more than 4000 operation hours. Within the European R and D project INDITEP the detailed engineering for a pre-commercial DSG solar thermal power plant with an electrical power of 5 MW is being performed. This small capacity is chosen to minimise the risk for potential investors. Regarding DSG solar thermal power plants only steam cycles using superheated steam have been investigated so far. In this paper a steam cycle operated with saturated steam is investigated for the first time. For near term applications this might be an interesting alternative in the chosen small capacity range. This choice would offer some specific advantages: (a) Lower complexity of power block and thus lower investment but also lower efficiency of the power block, (b) Simple set up of the collector field, (c) Proven safe operation and higher thermal efficiency of the collector field. (orig.)

  9. Repair technique for steam generator tubes using electroforming

    International Nuclear Information System (INIS)

    Pickering B CANDU Unit 5 had experienced leakage at sleeve/tube joint due to severe and local pitting in 1992δ1993. One year later, OHT developed electrosleeving techniques for steam generator tube repair which was applied at Pickering B CANDU Unit 5, Oconee Unit 1 and Callaway in 1994, 1995 and 1999 respectively. In the results of electrosleeved tube test, electrosleeve materials were stronger than mother tubes in mechanical properties and corrosion resistance under design criteria. Two analytical models were originally developed for estimating the failure temperature under severe accident transients. Electrosleeve, a structural layer of fine grained nickel is electroformed onto the strike by circulating an aqueous solution of Ni sulfate or sulfamate with NiCO3. The patents published by FTI said that the electrolyte for electroforming the structural layer contains a pinning agent to inhibit growth of metal grains in the electroformed layer. The pinning agent contains phosphoric, phosphorous acid, molybdenum. In localization of electrosleeving, there are some problems like as 1)low plating rate, 2)high residual stress, 3)alloy composition, 4)low material properties at high temperature. Ni-Fe plating exhibit anomalous codeposition; that is less noble metal, Fe, deposits preferentially to the more noble metal, Ni. Ductility decrease and residual stress increase with increase of Fe content in plate layer. Addition of particle size of 10δ400μm makes residual stress compressive in plate layer. Composite plating show excellent high temperature properties

  10. Third international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

    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

  11. 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.

  12. Strategic maintenance plan for Cernavoda steam generators

    International Nuclear Information System (INIS)

    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)

  13. Circumferential cracking of steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Karwoski, K.J.

    1997-04-01

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

  14. Operating experiences with 1 MW steam generator

    International Nuclear Information System (INIS)

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

  15. Hockey-stick steam generator for LMFBR

    International Nuclear Information System (INIS)

    This paper presents the criteria and evaluation leading to the selection of the Hockey Stick Steam Generator Concept and subsequent development of that concept for LMFBR application. The selection process and development of the Modular Steam Generator (MSG) is discussed, including the extensive test programs that culminated in the manufacture and test of a 35 MW(t) Steam Generator. The design of the CRBRP Steam Generator is described, emphasizing the current status and a review of the critical structural areas. CRBRP steam generator development tests are evaluated, with a discussion of test objectives and rating of the usefulness of test results to the CRBRP prototype design. Manufacturing experience and status of the CRBRP prototype and plant units is covered. The scaleup of the Hockey Stick concept to large commercial plant application is presented, with an evaluation of scaleup limitations, transient effects, and system design implications

  16. CAREM-25 Steam Generator Stability Analysis

    International Nuclear Information System (INIS)

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

  17. Cleanliness criteria to improve steam generator performance

    International Nuclear Information System (INIS)

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

  18. Proceedings of the fourth international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. Testing installation for a steam generator

    International Nuclear Information System (INIS)

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

  1. Steam Generator Inspection Planning Expert System

    International Nuclear Information System (INIS)

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

  2. 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.

  3. Steam generator snubber elimination program

    International Nuclear Information System (INIS)

    Hydraulic snubbers are used in Steam Generator (SG) upper support systems because of the snubbers ability to accommodate the large thermal movement of the SG during plant heatup and cooldown while providing restraint to the Reactor Coolant Loop (RCL) for dynamic events such as postulated pipe ruptures and earthquakes. Due to the complexity of the design of the SG upper support system, several factors can significantly affect the performance of the hydraulic snubbers and contribute to the continuing need to monitor and test the functionality of the snubbers. In some cases, snubber malfunctions have caused extended plant outages. Recent advancements in computer technology have made it economically feasible to perform the engineering analyses required to eliminate all of the SG snubbers. Similar to the SG snubber reduction programs started in the 1980s, this paper presents a process of implementing load reduction techniques and taking advantage of component reserve margins. Non-linear time history analyses are required for seismic and pipe rupture loadings to more accurately predict the behavior of the RCL and loads on the primary components. Even without considering the ALARA savings, the payback period to implement a full SG snubber removal program is typically less than two fuel cycles. (author)

  4. An expert system for steam generator maintenance

    International Nuclear Information System (INIS)

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

  5. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

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

  6. Centrifugal steam-water separator for steam generators

    International Nuclear Information System (INIS)

    This invention concerns a centrifugal steam-water separator for steam generators, using natural circulation. The turbulence chamber includes a perforated venturi composed of a decreasing cone-shaped convergent duct and a cone-shaped divergent diffuser section increasing from the narrowest part to the turbulence chamber outlet. In this way, the jected liquid phase and any particles of solids it may contain can be discharged through the perforations into the annular space formed between the perforated venturi and the vessel to accumulate at the bottom of this annular space for subsequent removal. The advantages of the invention are that the diffuser of the perforated venturi is used as an additional separation path and with the recovery of pressure in mind, and that the water droplets ejected, as well as any particles contained in these droplets discharged or ejected outside the action area of the rotational flow into the annular space, can flow in a practically free way towards the bottom of the interior edge of the containment wall. Because of this, the pressure drop is reduced and the degree of separation improved. The steam-water separator of the invention is therefore particularly suitable for the high power steam generators of nuclear reactor facilities. For a given steam output, it is possible with the lay-out specified in this invention to reduce the required number of separation units

  7. 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

  8. Failure analysis of retired steam generator tubings

    International Nuclear Information System (INIS)

    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

  9. Steam generator tube laser sleeving

    International Nuclear Information System (INIS)

    For many years, Framatome has been used with different techniques and means to perform the steam generator tube sleeving operation, such as the 'mechanical' either GTAW welded or kinetic welding process. As soon as first power laser units appeared on the market we felt right interested in applying this process to the sleeving operation. After comparison between all the processes and equipments existing at that time (that is to say CO2 and YAG laser units), we chose the YAG and bought a 1.2 kW NEC laser unit in 1988. As it was installed in our Welding Center of Le Creusot, this equipment enabled us carrying out a preliminary test programme which targets were: getting the mastery of the equipment and associated technologies, implementing this process for the sleeve welding operation by improvement of the welding-pen. The NEC laser unit became afterwards transferred to our workshop in Chalon-sur-Saone (June 1990), where we achieved the final tests of the process at the same time we were investigating the development of industrial operation means. The actual program is mainly focused on 7/8'' tube steam generator repair process at tubesheet outlet. Yet made sure that our methods and means apply as well to 3/4'' tubes up to second tube support-plate level. Sleeves are made of heat-treated Inconel 690. The sleeved unit has been designed to provide the same breaking strength and leak-tightness as the tube. The upper part of sleeve consists of an anti-pop out length which ensures some locking-up in case the tube breaks in upper transition expansion area. Preliminary tests dealt with the various parameters which may exert an influence on geometry and quality of the, weld bead, as: - laser beam power (for continuous and pulsed modes), - welding speed, - focal spot size and location from the surface to be welded, - protective gas. After performance of preliminary tests on many thousands of weld beads we decided to use the process according to following criteria: Weld quality

  10. Performance surveillance and failure analysis research for steam generators, turbines and condensers

    International Nuclear Information System (INIS)

    Failure prevention and analysis research program is on going for the steam generators, condensers and turbines of PWR's and PHWR CANDU in order to obtain analysis and evaluation technology for material failures which take place and might occur in the Korean nuclear power reactors in future. As a part of this program, failure experiences have been surveyed for steam generators, condensers and turbines of PWR's in foreign countries, and operating experiences have been surveyed for nuclear power unit 1 in Korea with PSI and ISI eddy current test data, water chemistry data and failed specimens. The Korean nuclear power unit 1 has minor denting indications and boric acid soaking has been performed in order to control the denting. The effects of this boric acid soaking on the material degradation of the PWR secondary system are being studied. Stress corrosion cracking research work is also under way for steam generator U-tubes. (Author)

  11. Optimized CANDU-6 cell and reactivity device supercell models for advanced fuels reactor database generation

    International Nuclear Information System (INIS)

    Highlights: • Propose an optimize 2-D model for CANDU lattice cell. • Propose a new 3-D simulation model for CANDU reactivity devices. • Implement other acceleration techniques for reactivity device simulations. • Reactivity device incremental cross sections for advanced CANDU fuels with thorium. - Abstract: Several 2D cell and 3D supercell models for reactivity device simulation have been proposed along the years for CANDU-6 reactors to generate 2-group cross section databases for finite core calculations in diffusion. Although these models are appropriate for natural uranium fuel, they are either too approximate or too expensive in terms of computer time to be used for optimization studies of advanced fuel cycles. Here we present a method to optimize the 2D spatial mesh to be used for a collision probability solution of the transport equation for CANDU cells. We also propose a technique to improve the modeling and accelerate the evaluation, in deterministic transport theory, of the incremental cross sections and diffusion coefficients associated with reactivity devices required for reactor calculations

  12. Steam generator replacement in Bruce A Unit 1 and Unit 2

    International Nuclear Information System (INIS)

    The Bruce A Generating Station consists of four 900 MW class CANDU units. The reactor and Primary Heat Transport System for each Unit are housed within a reinforced concrete reactor vault. A large duct running below the reactor vaults accommodates the shared fuel handling system, and connects the four reactor vaults to the vacuum building. The reactor vaults, fuelling system duct and the vacuum building constitute the station vacuum containment system. Bruce A Unit 2 was shut down in 1995 and Bruce A Units 1, 3 and 4 were shutdown in 1997. Bruce A Units 3 and 4 were returned to service in late 2003 and are currently operating. Units 1 and 2 remain out of service. Bruce Power is currently undertaking a major rehabilitation of Bruce A Unit 1 and Unit 2 that will extend the in-service tile of these units by at least 25 years. Replacement of the Steam Generators (eight in each unit) is required; this work was awarded to SNC-Lavalin Nuclear (SLN). The existing steam drums (which house the steam separation and drying equipment) will be retained. Unit 2 is scheduled to be synchronized with the grid in 2009, followed by Unit 1 in 2009. Each Bruce A unit has two steam generating assemblies, one located above and to each end of the reactor. Each steam generating assembly consists of a horizontal cylindrical steam drum and four vertical Steam Generators. The vertical Steam Generators connect to individual nozzles that are located on the underside of the Steam Drum (SD). The steam drums are located in concrete shielding structures (steam drum enclosures). The lower sections of the Steam Generators penetrate the top of the reactor vaults: the containment pressure boundary is established by bellows assemblies that connect between the reactor vault roof slab and the Steam Generators. Each Steam Generators is supported from he bottom by a trapeze that is suspended from the reactor vault top structure. The Steam Generator Replacement (SGR) methodology developed by SLN for Unit 1

  13. Passive heat removal in CANDU

    International Nuclear Information System (INIS)

    CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D2O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

  14. Monte Carlo Few-Group Constant Generation for CANDU 6 Core Analysis

    OpenAIRE

    Seung Yeol Yoo; Hyung Jin Shim; Chang Hyo Kim

    2015-01-01

    The current neutronics design methodology of CANDU-PHWRs based on the two-step calculations requires determining not only homogenized two-group constants for ordinary fuel bundle lattice cells by the WIMS-AECL lattice cell code but also incremental two-group constants arising from the penetration of control devices into the fuel bundle cells by a supercell analysis code like MULTICELL or DRAGON. As an alternative way to generate the two-group constants necessary for the CANDU-PHWR core analys...

  15. Proceedings of the third international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    The third international conference on Candu maintenance included sessions on the following topics: predictive maintenance, reliability improvements, steam generator monitoring, tools and instrumentation, valve performance, fuel channel inspection and maintenance, steam generator maintenance, environmental qualification, predictive maintenance, instrumentation and control, steam generator cleaning, decontamination and radiation protection, inspection techniques, maintenance program strategies and valve packing experience, remote tooling/ robotics and fuel handling. The individual papers have been abstracted separately

  16. 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.

  17. Accoustic background of BN-600 steam generator

    International Nuclear Information System (INIS)

    In this paper the results of accoustic background for BN-600 steam generator in nominal operating conditions are presented. The 1-200 kHz accoustic background of evaporator and reheater modules are given

  18. Ultrasonic testing of steam generator tubes

    International Nuclear Information System (INIS)

    A system is developed for inspection of steam generator tube, especially near the tube plate. Imaging, thickness measurement, radial profilometry, longitudinal and circonferential crack detection and welded joints testing are reviewed

  19. US PWR steam generator management: An overview

    International Nuclear Information System (INIS)

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

  20. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  1. Natural circulation steam generator model for optimal steam generator water level control

    International Nuclear Information System (INIS)

    Several authors have cited the control of steam generator water level as an important problem in the operation of pressurized water reactor plants. In this paper problems associated with steam generator water level control are identified, and advantages of modern estimation and control theory in dealing with these problems are discussed. A new state variable steam generator model and preliminary verification results using data from the loss of fluid test (LOFT) plant are also presented

  2. Supporting CANDU operators-CANDU owners group

    International Nuclear Information System (INIS)

    The CANDU Owners Group (COG) was formed in 1984 by the Canadian CANDU owning utilities and Atomic Energy of Canada limited (AECL). Participation was subsequently extended to all CANDU owners world-wide. The mandate of the COG organization is to provide a framework for co-operation, mutual assistance and exchange of information for the successful support, development, operation, maintenance and economics of CANDU nuclear electric generating stations. To meet these objectives COG established co-operative programs in two areas: 1. Station Support. 2. Research and Development. In addition, joint projects are administered by COG on a case by case basis where CANDU owners can benefit from sharing of costs

  3. Electric-arc steam plasma generator

    Science.gov (United States)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. PWR steam generator chemical cleaning process testing in model steam generators

    International Nuclear Information System (INIS)

    Corrosion related problems in PWR power plant steam generators have caused high maintenance costs, increased radiation exposure to plant personnel, and reduced unit availability. Two cleaning methods were investigated for their ability to clean deposits from steam generators thereby increasing the integrity of the steam generators and reducing personnel radiation exposure, due to reduced maintenance. First, an on-line chemical cleaning process (Chelant Addition) was tested for its ability to prevent corrosion product buildup in a steam generator. Second, an off-line dilute chemical cleaning process was tested to evaluate its ability to remove corrosion product deposits and leave minimal waste for disposal. These two processes were tested in model steam generators which simulated the operating conditions of a typical full size steam generator. Six model steam generators (MSG) were fabricated and qualified for their ability to reproduce denting at tube support plates. The results of six chemical cleaning tests and the post-cleaning destructive metallurgical evaluation of two of the model steam generators are reported

  7. Steam generator tube inspection and maintenance technology

    International Nuclear Information System (INIS)

    I am very happy to introduce a report which I'm sure will shows an important progress to the inspection and repair of steam generator in nuclear power industry. It will describe a tube inspection which has been applied to detect PWSCC at the top of tubesheet and a tube repair KPS supplied during last outage (Dec. 1998 to Feb. 1999). The presentation describe what KPS performed for the PWSCC degradation mechanism found in Ulchin unit 1 and 2 and explain the effort for new techique application to improve conventional techniques. Ulchin Nuclear Power Plants consist of 4 units which are 2 units (no. 1,2) with 980 Mwg and 2 units (no. 3,4) with 1000 Mwg. Ulchin unit 1 did initial startup on Jan. 1986. It has three Westinghouse Series 51 B Steam Generator with closed crevice tubesheet and kiss roll expansion. 51 B Steam Generators in Ulchin unit 1 and 2 have: - Recirculating SG with 3,330 U tubes each (0.875 OD x 0.05 Thick) , - Inconel 600 MA, - Full hydraulic expansions, - Thot of 620 deg F. Primary Water Stress Corrosion Cracks (PWSCC) on steam generator tube have been found in Ulchin unit 1 and 2 since the forth outage (1993). So these units have been much affected by PWSCC for normal operation. Therefore, the utility has continued to look for the advanced inspection and maintenance technology to extend the life of degraded steam generator tube. (author)

  8. Fourth international seminar on horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  9. Fourth international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

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

  10. Behaviour of the steam generator tubing in water with different pH values

    International Nuclear Information System (INIS)

    Steam generators are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Heavy Water Reactor (PHWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators. The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. However, the steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related to corrosion. The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism with the purpose of evaluating the quantities of corrosion products which exist in the steam generator after a determined period of operation. The purpose of the experimental research consists in the assessment of corrosion behaviour of the tubes material, Incoloy-800, at normal secondary circuit parameters (temperature-260 oC, pressure-5.1 MPa). The testing environment was the demineralised water without impurities, at different pH values regulated with morpholine and cyclohexylamine (all volatile treatment-AVT). The results are presented like micrographics and graphics representing weight loss of metal due to corrosion, corrosion rate, total corrosion products formed, the adherent corrosion products, released corrosion products, release rate of corrosion products and release rate of the metal.

  11. Behaviour of the steam generator tubing in water with different pH values

    Energy Technology Data Exchange (ETDEWEB)

    Lucan, Dumitra, E-mail: dumitra.lucan@nuclear.r [Department of Corrosion and Circuits Chemistry, Institute for Nuclear Research, POB 78, Pitesti (Romania)

    2011-04-15

    Steam generators are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Heavy Water Reactor (PHWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators. The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. However, the steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related to corrosion. The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism with the purpose of evaluating the quantities of corrosion products which exist in the steam generator after a determined period of operation. The purpose of the experimental research consists in the assessment of corrosion behaviour of the tubes material, Incoloy-800, at normal secondary circuit parameters (temperature-260 {sup o}C, pressure-5.1 MPa). The testing environment was the demineralised water without impurities, at different pH values regulated with morpholine and cyclohexylamine (all volatile treatment-AVT). The results are presented like micrographics and graphics representing weight loss of metal due to corrosion, corrosion rate, total corrosion products formed, the adherent corrosion products, released corrosion products, release rate of corrosion products and release rate of the metal.

  12. Development of an on-line process for steam generator chemical cleaning

    Energy Technology Data Exchange (ETDEWEB)

    Semmler, J.; Guzonas, D.A.; Rousseau, S.C.; Snaglewski, A.P.; Chenier, M.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1998-07-01

    An on-line, preventative chemical cleaning process for the removal of secondary side oxides from steam generators is being developed. An on-line chemical cleaning process uses a low concentration of a chelant ({approx}1-10 mg L{sup -1}) to partially dissolve and dislodge the secondary side oxides while the steam generator is in operation. The dissolved and dislodged oxides can then be removed by blowdown. Feasibility tests were carried out in which the operating conditions of a CANDU steam generator were simulated in an autoclave containing either loose powdered magnetite or sintered magnetite on Alloy 800 (I-800) steam generator tube surfaces. The extent of magnetite dissolution in on-line solvent formulations containing either ethylenediaminetetraacetic acid (EDTA) or N-(2-hydroxyethyl)ethylenedinitrilo-N,N',N'-triacetic acid (HEDTA) at temperatures of 256 and 263 degrees C were measured. Powdered magnetite dissolved faster than sintered magnetite using both types of chelant. Dissolution continued as fresh chelant was added. The half-life (t{sup 1/2}) of Fe-EDTA complexes at 256 degrees C was approximately 3 h, sufficient to allow removal by blowdown. Hydrazine and morpholine were equally effective as oxygen scavengers. Increased dissolved oxygen concentration was found to result in chelant decomposition, reduced solvent capacity and increased carbon steel corrosion. Total corrosion of several materials relevant to CANDU stations were measured in 96-h tests. To minimize corrosion, low concentration of chelant and a high concentration of an oxygen scavenger should be used. The results from these feasibility tests are currently being used to define the application conditions for large-scale tests of on-line chemical cleaning in a model steam generator. (author)

  13. Development of an on-line process for steam generator chemical cleaning

    International Nuclear Information System (INIS)

    An on-line, preventative chemical cleaning process for the removal of secondary side oxides from steam generators is being developed. An on-line chemical cleaning process uses a low concentration of a chelant (∼1-10 mg L-1) to partially dissolve and dislodge the secondary side oxides while the steam generator is in operation. The dissolved and dislodged oxides can then be removed by blowdown. Feasibility tests were carried out in which the operating conditions of a CANDU steam generator were simulated in an autoclave containing either loose powdered magnetite or sintered magnetite on Alloy 800 (I-800) steam generator tube surfaces. The extent of magnetite dissolution in on-line solvent formulations containing either ethylenediaminetetraacetic acid (EDTA) or N-(2-hydroxyethyl)ethylenedinitrilo-N,N',N'-triacetic acid (HEDTA) at temperatures of 256 and 263 degrees C were measured. Powdered magnetite dissolved faster than sintered magnetite using both types of chelant. Dissolution continued as fresh chelant was added. The half-life (t1/2) of Fe-EDTA complexes at 256 degrees C was approximately 3 h, sufficient to allow removal by blowdown. Hydrazine and morpholine were equally effective as oxygen scavengers. Increased dissolved oxygen concentration was found to result in chelant decomposition, reduced solvent capacity and increased carbon steel corrosion. Total corrosion of several materials relevant to CANDU stations were measured in 96-h tests. To minimize corrosion, low concentration of chelant and a high concentration of an oxygen scavenger should be used. The results from these feasibility tests are currently being used to define the application conditions for large-scale tests of on-line chemical cleaning in a model steam generator. (author)

  14. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations both to the workers and the public

  15. Quality Products - The CANDU Approach

    International Nuclear Information System (INIS)

    The prime focus of the CANDU concept (natural uranium fuelled-heavy water moderated reactor) from the beginning has economy, heavy water losses and radiation exposures also were strong incentives for ensuring good design and reliable equipment. It was necessary to depart from previously accepted commercial standards and to adopt those now accepted in industries providing quality products. Also, through feedback from operating experience and specific design and development programs to eliminate problems and improve performance, CANDU has evolved into today's successful product and one from which future products will readily evolve. Many lessons have been learned along the way. On the one hand, short cuts of failures to understand basic requirements have been costly. On the other hand, sound engineering and quality equipment have yielded impressive economic advantages through superior performance and the avoidance of failures and their consequential costs. The achievement of lifetime economical performance demands quality products, good operation and good maintenance. This paper describes some of the basic approaches leading to high CANDU station reliability and overall excellent performance, particularly where difficulties have had to be overcome. Specific improvements in CANDU design and in such CANDU equipment as heat transport pumps, steam generators, valves, the reactor, fuelling machines and station computers, are described. The need for close collaboration among designers, nuclear laboratories, constructors, operators and industry is discussed. This paper has reviewed some of the key components in the CANDU system as a means of indicating the overall effort that is required to provide good designs and highly reliable equipment. This has required a significant investment in people and funding which has handsomely paid off in the excellent performance of CANDU stations. The close collaboration between Atomic Energy of Canada Limited, Canadian industry and the

  16. 49 CFR 229.105 - Steam generator number.

    Science.gov (United States)

    2010-10-01

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

  17. Steam turbine and generator system for ABWR

    International Nuclear Information System (INIS)

    The advanced boiling water reactor (ABWR), with a number of superior characteristics including high reliability and large capacity, has been developed. The first ABWR units have been realized as Units No. 6 and 7 of The Tokyo Electric Power Co., Inc.'s Kashiwazaki-Kariwa Nuclear Power Station. Based on its considerable experience in the construction and maintenance of nuclear steam turbine and generator systems, Toshiba has developed the world's largest class steam turbine as well as generator systems with high efficiency, high reliability, small turbine buildings and a short construction period, and has been constructing Kashiwazaki-Kariwa Unit No. 7 jointly with General Electric Co. of the United States. We have been developing additional techniques to further improve the efficiency and maintainability of the steam turbine and generator system of the next ABWR plant, based on techniques that have been verified with fossil-fuel power plants. (author)

  18. Recent advances in ultrasonic downcomer flow-measurement techniques for recirculating steam generators

    International Nuclear Information System (INIS)

    Non-intrusive ultrasonic measurements of downcomer flow velocity have been successfully used in the past to determine recirculation ratios and water inventory in CANDU steam generators. Knowledge of these process conditions allows operators to assess the effectiveness of maintenance programs, monitor the effects of tube fouling, and observe flow conditions following component modifications. It also provides designers with a means to verify or improve code predictions. Ultrasonic measurement systems have recently been installed on sixteen steam generators at the Bruce B Nuclear Generating Station, as part of an investigation into the possible effects of long-term boiler degradation. The most recent version of AECL's downcomer-flow technology was used, which features high-temperature transducers that are attached magnetically and then welded to the steam-generator outer shell. This method eliminates the complications of precision surface preparation, high-temperature couplants and awkward mechanical attachments. The paper will outline the method and summarize flow velocities measured during normal operation, over extended periods of time. It will also describe how the information might be used, e.g., to assess thermalhydraulic conditions, verify design calculations and support the case for reactor uprating. Further improvements that may allow the reliable measurement of flow in steam generators with steam carry-under are suggested, and preliminary results are presented from a dual-purpose single- and two-phase flow-measurement system. (author)

  19. 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)

  20. Acoustic leak detector in Monju steam generator

    International Nuclear Information System (INIS)

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

  1. Stability study in one step steam generators

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  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. Flow-induced vibration analysis of heat exchanger and steam generator designs

    International Nuclear Information System (INIS)

    Tube and shell heat exchange components such as steam generators, heat exchangers and condensers are essential parts of CANDU nuclear power stations. Excessive flow-induced vibration may cause tube failures by fatigue or more likely by fretting-wear. Such failures may lead to station shutdowns that are very undesirable in terms of lost production. Hence good performance and reliability dictate a thorough flow-induced vibration analysis at the design stage. This paper presents our approach and techniques in this respect. (author)

  5. Revised evaluation of steam generator testing alternatives

    International Nuclear Information System (INIS)

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

  6. Demonstration test of steam generator reliability

    International Nuclear Information System (INIS)

    The first breaking of steam generator tubes in Japan occurred in June, 1972, in Mihama No. 1 plant after only one and a half year operation. In 1973, the operation was started again by plugging the broken tubes. Since then, the breaking of steam generator tubes has been often reported though the treatment of secondary cooling water was changed from sodium phosphate treatment to all volatile treatment. In Japan, the breaking was caused by stress corrosion and the loss of wall thickness, and denting was never experienced. The Ministry of International Trade and Industry planned the demonstration test on the reliability of steam generators, and the large scale experiment was carried out in the Takasago Research Institute, Mitsubishi Heavy Industries, Ltd. The objectives were to examine the factors affecting corrosion in the case of all volatile water treatment, and the reason why the concentration of sodium phosphate corrodes heating tubes. The test was related to the thermal flow behavior of the secondary side, the corrosion of heating tubes and baffle plates, and the rupture behavior of heating tubes. This test was carried out for seven years since 1975, and the reliability and safety of steam generators were confirmed. (Kako, I.)

  7. Evaluation of surrogate boilers for steam generators

    International Nuclear Information System (INIS)

    Steam generator damage in pressurized water reactors is a continuing problem which results from a combination of factors including mechanical design, thermal hydraulics, materials selection, fabrication techniques, water chemistry, and system design and operation. A wide variety of steam generator damage mechanisms has been identified in operating PWRs including intergranular attack, thinning, stress corrosion cracking, erosion, denting, fatigue cracking, pitting, and fretting. Model boilers operated in parallel to the steam generators, i.e., surrogate boilers, may provide a useful tool in the study of these damage mechanisms, their causative factors, and the effects of corrosion actions. To evaluate the applicability of surrogate boilers to such studies, Steam Generator Owners Group I project S111-2 was established. Evaluation of numerous surrogate boiler design alternates led to identification of several possible acceptable approaches. The appropriate surrogate feedwater was identified as plant feedwater. Capability to operate with a tube-side temperature similar to the hot-leg temperature was considered necessary as was the ability to provide mechanical, thermal, and chemical corrosion acceleration. Practical and economically feasible surrogate boiler designs were developed in response to these design requirements

  8. Failure Analysis of Retired Steam Generator Tubings

    International Nuclear Information System (INIS)

    Since the first commercial operation of Kori-1 in 1978, 20 units of nuclear power plants are operated, and the it covers 40 % of total electricity in Korea as of 2008. A steam generator tube rupture incident occurred in the Ulchin unit 4 in 2002, which made the public sensitive to nuclear power plant. In order to keep the nuclear energy as a main energy source, the integrity of steam generator should be demonstrated. It is important to improve a flaw detection capability of the eddy current testing(ECT) in steam generator(SG) tubings in order to maintain the tube integrity. A quantified evaluation on the flaws on SG tubings, which is crucial for the tube integrity evaluation is not satisfactory. It is necessary to utilize the retired SG having various types of corrosion damages. In addition, an examination of pulled tube from Kori 1 retired steam generator will give us information about effectiveness of a remedial action(TiO2 addition) which was applied to mitigate ODSCC. A crack growth model is also needed to ensure a tube repair criteria for a next fuel cycle based on the ASME safety evaluation code, which has to meet a requirement that the flaws have to sustain under three times of normal operation pressure difference and 1.4 times of severe accident condition. In this project, hardware such as semi hot lab for pulled tube examination and modification of 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. The non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in the semi hot lab. An effect of remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. An electrochemical decontamination technology for pulled tube was developed to reduce a radiation exposure and enhance

  9. Challenges in manufacture of PFBR steam generators

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe pool type sodium cooled reactor. The main function of Steam Generator (SG) is to extract the reactor heat through secondary sodium system and convert the feed water into superheated steam in the tubes of SG. The Steam Generator is a vertical, once through, shell and tube type heat exchanger with liquid sodium in the shell side and water/steam in the tube side. The highly reactive nature of sodium with water/steam requires that the sodium to water/steam boundaries of the steam generators must possess a high degree of reliability against failure. This is achieved in design and manufacture by maximizing the tube and tubesheet integrity and more importantly by proper selection of tube to tubesheet joint configuration. Modified 9Cr-1Mo material is selected as major material of construction for steam generator as this material has excellent high temperature mechanical properties and has high resistance to stress corrosion cracking in caustic and chloride environment. Steam Generator has inlet and outlet nozzles which are made from the pullout shells of SG. Pullouts are made by hot forming process by heating the shell inside the furnace to the temperature of 950-1100 deg C followed by die and punch pressing. After forming, the pullouts are subjected to normalizing at 1040-1050 deg C followed by tempering at 780 deg C to restore the original material properties. The forming of nozzle pullouts is really difficult and challenging task as the dimensions are too large and dimensional tolerances are very tight. Steam Generator shell assembly is fabricated in horizontal condition after completion of tube bundle which requires insitu welding of shells around the tube bundle. Dimension control during shell welding is extremely difficult as internal fixtures/spiders for ovality control is ruled out due to existence of tube bundle. Utmost care is required during shell welding to avoid arc strike/fusion on the tube (i.e sodium

  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. Recent experience related to neutronic transients in Ontario Hydro CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Ontario Hydro presently operates 18 CANDU reactors in the province of Ontario, Canada. All of these reactors are of the CANDU Pressurized Heavy Water design, although their design features differ somewhat reflecting the evolution that has taken place from 1971 when the first Pickering unit started operation to the present as the Darlington units are being placed in service. Over the last three years, two significant neutronic transients took place at the Pickering Nuclear Generating Station 'A' (NGS A) one of which resulted in a number of fuel failures. Both events provided valuable lessons in the areas of operational safety, fuel performance And accident analysis. The events and the lessons learned are discussed in this paper

  12. Dynamic simulation of steam generator failures

    International Nuclear Information System (INIS)

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

  13. Corrosion problems of PWR steam generators

    International Nuclear Information System (INIS)

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

  14. Leakage experiences with 1 MW steam generator

    International Nuclear Information System (INIS)

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

  15. Regulatory perspective on steam generator tube operating experience

    International Nuclear Information System (INIS)

    The United States Nuclear Regulatory Commission (NRC), as part of its mission has to protect public health and safety and the environment. In fulfilling this mission the NRC: (1) monitors and assesses steam generator tube operating experience and (2) reviews industry proposals to revise steam generator tube inspection and repair requirements in the technical specifications. This paper reviews significant occurrences, trends, and issues in the recent past relating to steam generator tube integrity from a regulatory perspective. This paper focuses on (1) the status of replacing steam generators, (2) recent experience at plants with thermally-treated Alloy 600 tubes, (3) recently issued or planned generic communications, and (4) the status of revising the steam generator portion of the technical specifications. Each of these focus topics are discussed in the following sections: Steam generator replacement status; Thermally-treated alloy 600 steam generator tube experience; Steam generator generic communications; Status of new steam generator technical specifications. In conclusion, as a result of tube degradation associated with original steam generator designs, many plants have replaced their steam generators. The operating experience associated with the newer steam generator designs has been favorable with only a few instances of cracking being reported. The NRC and the U.S. nuclear industry continue to improve their programs for managing steam generator tube integrity. In addition, the NRC and the industry are making advances in improving the regulatory requirements pertaining to tube inspections. (authors)

  16. Super critical water reactor for use in steam generation for recovery of bitumen resources

    International Nuclear Information System (INIS)

    The process of recovering the bitumen (oil sand) resources in Alberta requires steam at high pressures. To help reduce the carbon footprint of exploiting these fuel resources, an innovative new design of a CANDU super critical water reactor (CANDU-SCWR) is being considered to provide the high pressure steam required for the steam assisted gravity drainage (SAGD) process. The high temperature and pressure associated with the CANDU-SCWR allow for the high pressure, temperature steam to be produced without supplementary energy. The Petroleum Technology Alliance of Canada (PTAC) has specified the SAGD process requires steam at 11 MPa and near 100% steam quality, and net electrical power of 106 MWe. This paper examines steam cycle and design options to meet the steam and power requirements defined by PTAC. Steam cycle options are examined focusing on the optimization of steam and power conversion. Additionally passive safety and cooling for both the heat transport and moderation systems are considered and their impact on performance are examined. As the CANDU-SCWR is at a preliminary stage of design, basic design parameters have been defined based on preliminary assessments. This paper is focused on a reactor with the following basic design assumptions: Vertical fuel channel; Re-entrant fuel channels; Pu-Th fuel; and Batch refuelling. (author)

  17. Steam generator auto-depressurization - An efficient way to ensure PHT heat sink

    International Nuclear Information System (INIS)

    This paper is a brief report of a sensitivity study of the role of auto-depressurization automatic action of steam generators in some special accident conditions in a CANDU 6 NPP. This action is controlled by the Boiler Level Control program in order to ensure a sink for the primary heavy water when the feedwater to steam generators is totally lost and there is not available another alternative sink. This action consists in 8 Main Steam Safety Valves opening when the boiler level and feedwater line pressure decrease under some established values and should determine the fast decreasing of boiler pressure at a low value that can allow the Boiler Make-up Water intervention and the make-up of the steam generator water inventory. One of the particular objectives of this study is to determine the minimum number of Main Steam Safety Valves that must be available and open when the auto-depressurization action is demanded, to ensure the required low boiler pressure to allow Boiler Make-up Water (BMW) injection before steam generators to be empty. The initiating event, that can lead to a total loss of feedwater flow to steam generators, analyzed in accident conditions described above, is the total loss of Class IV power supply coincident with the auxiliary feedwater pump unavailability. This event is supposed to occur when the reactor operates at nominal power in equilibrium fuel conditions. Results of the analysis and conclusions regard mainly the efficiency of the auto-depressurization action performed with a different number, less than 8, of MSSV's available and the safety state of the plant during the event. (author)

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

    International Nuclear Information System (INIS)

    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

  19. 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.

  20. Future aspects for liquid metal heated steam generators

    International Nuclear Information System (INIS)

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

  1. CANDU-BLW-250

    International Nuclear Information System (INIS)

    The plant 'La Centrale nucleaire de Gentilly' is located between Montreal and Quebec City on the south shore of the St. Lawrence River and start-up is scheduled for 1971. A CANDU-BLW reactor is the nuclear steam generator. his reactor utilizes a heavy water moderator, natural uranium oxide fuel, and a boiling light water coolant. To be economic, this type of plant must have a minimum light water inventory in the reactor core. A minimum inventory is obtained (a) by reducing the cross-sectional area for coolant flow to a minimum, and (b) by operating at a low-coolant density. In CANDU-BLW-250, this is accomplished by operating a closed spaced fuel rod bundle at high steam quality. These features and others in the BLW concept lead to a number of areas of concern and they are summarized below: (1) Heat Transfer: It is intended that under normal operating conditions the fuel sheaths will always be wetted with coolant. (ii) Hydrodynamic Stability: Experiments and analysis indicate that the plant has a considerable over-power capacity before instability is predicted. (iii) Control: This plant does have a positive power coefficient and the transient performance with various disturbances are detailed. (iv) Safety: The positive power coefficient leads to concern over the loss of coolant accident. The results of some accident analysis are presented. (author)

  2. COMMAND AND CONTROL STRATEGIES APPLIED TO HIGHPOWER STEAM GENERATORS

    OpenAIRE

    DUINEA. A.M.; MIRCEA P.M.

    2015-01-01

    The paper presents the analysis of the actual operation scheme existing for steam generator drum. Following the trend valid for forced circulation steam generator, it is proposed to replace the classical adjustment loops with new regulation scheme highlighting its advantages in steam generation operation.

  3. COMMAND AND CONTROL STRATEGIES APPLIED TO HIGHPOWER STEAM GENERATORS

    Directory of Open Access Journals (Sweden)

    DUINEA. A.M.

    2015-06-01

    Full Text Available The paper presents the analysis of the actual operation scheme existing for steam generator drum. Following the trend valid for forced circulation steam generator, it is proposed to replace the classical adjustment loops with new regulation scheme highlighting its advantages in steam generation operation.

  4. Modelling the horizontal steam generator with APROS

    International Nuclear Information System (INIS)

    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.)

  5. Artificial Intelligence Techniques for Steam Generator Modelling

    CERN Document Server

    Wright, Sarah

    2008-01-01

    This paper investigates the use of different Artificial Intelligence methods to predict the values of several continuous variables from a Steam Generator. The objective was to determine how the different artificial intelligence methods performed in making predictions on the given dataset. The artificial intelligence methods evaluated were Neural Networks, Support Vector Machines, and Adaptive Neuro-Fuzzy Inference Systems. The types of neural networks investigated were Multi-Layer Perceptions, and Radial Basis Function. Bayesian and committee techniques were applied to these neural networks. Each of the AI methods considered was simulated in Matlab. The results of the simulations showed that all the AI methods were capable of predicting the Steam Generator data reasonably accurately. However, the Adaptive Neuro-Fuzzy Inference system out performed the other methods in terms of accuracy and ease of implementation, while still achieving a fast execution time as well as a reasonable training time.

  6. Mathematical models for steam generator accident simulation

    International Nuclear Information System (INIS)

    In this contribution, the numerical methods used in the DeBeNe-LMFBR development for the analysis of the hydrodynamic and mechanical consequences of steam generator accidents are presented. At first the definition of the source term, i.e. the water leak rate which has to be assumed in the design basis accident as well as the thermochemistry of the sodium/water-reaction is discussed. Then the computer-codes presently used to describe the hydrodynamic and mechanical consequences of steam generator accidents on the basis of the above mentioned source term are presented. These comprise the code-system SAPHYR and the code PTANER and PISCES. Furthermore, developments which are planned or already under way for future use, such as the BEREPOT-code, are presented. (author)

  7. 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.

  8. Three Steam Generator Replacement Projects in 1995

    International Nuclear Information System (INIS)

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

  9. AGR operational experience - steam generator materials constraints

    International Nuclear Information System (INIS)

    Steam generator material problems which have arisen in Hinkley Point B and Hunterston B are discussed. Four examples are described in detail. These are: gas-side oxidation of the 9Cr-1Mo superheater, stress-corrosion of the austenitic superheater, creep of the transition joint between the 9Cr-1Mo and austenitic superheaters, erosion-corrosion of the economizer inlet orifice carriers. (U.K.)

  10. EP 1000 steam generator tube rupture analyses

    International Nuclear Information System (INIS)

    European electrical utility organizations together with Westinghouse and Ansaldo are participating in a program to utilize the Westinghouse passive nuclear plant technology to develop a plant which meets the European Utility Requirements (EUR) and is expected to be licensable in Europe. The program was initiated in 1994 and the plant is designated EP1000. The EP1000 design is notable for simplicity that comes from a reliance on passive safety systems to enhance plant safety. The use of passive safety systems has provided significant and measurable improvements in plant simplification, safety, reliability, investment protection and plant costs. These systems use only natural forces such as gravity, natural circulation, and compressed gas to provide the driving forces for the systems to adequately cool the reactor core following an initiating event. The EP1000 builds up on the Westinghouse passive nuclear plant technology to enhance plant safety and meet European Utility Requirements and specific European National Safety Criteria. This paper summarizes the main results of the Steam Generator Tube Rupture (SGTR) analysis activity, performed in Phase 2B of the European Passive Plant Program. The purpose of the study is to provide evidence that the passive safety system performance provides a significant improvement in terms of safety, providing significant margins to steam generator overfilling and reducing the need for operator actions. The behavior of the EP1000 plant following SGTR accidents has been analyzed by means of the RELAP5/Mod3.2 code. Sensitivity cases were performed, to address the impact of varying the number of steam generator tubes that rupture, and the potential adverse interactions that could result from operation of control systems (i.e., Chemical and Volume Control System, Startup Feedwater). Analyses have also been performed to define and verify improved protection system logic to avoid possible steam generator safety valve challenges both in the

  11. Perspectives of conventional and nuclear steam generation

    International Nuclear Information System (INIS)

    In the years to come, steam generation will be influenced by the following trends: 1) substitution of coal for petroleum, 2) a steady rise in energy costs, 3) environmental protection. The German boiler industry should try to maintain and further develop its high standard in order to be competitive on the world market in spite of high wages. Exports are absolutely necessary in view of the strongly fluctuating demand in Germany. (orig.)

  12. Fracture toughness determination in steam generator tubes

    International Nuclear Information System (INIS)

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

  13. Future steam generator designs. Single wall designs

    International Nuclear Information System (INIS)

    The easily removable 'U' tube design style adopted in the UK for the existing PFR Steam Generators, the Replacement Units now in production and for the future CDFR, gives the operator an extremely valuable option in the event of a water/steam leak occurring inside the Steam Generator. He can choose to shut-down, attempt to find the leak, assess damage, repair, revalidate and return to service in situ, or he can elect to remove the defect unit and replace with a 'proven' spare before returning the circuit to power. With the latter approach the resultant outage time is a known entity of about two weeks. If a repair is attempted in situ, predictions of outage time can become a matter of guesswork since one has no 'guaranteed' method of leak location and the assessment of secondary damage may be very time consuming, depending on the size and type of the original leak together with the particular design style of the Steam Generator. A further significant advantage of the removable 'U' tube design concept is that periodic interchanging of bundles with a spare enables routine chemical cleaning and thorough scheduled tube inspections, with specimen tube sample removal if required for monitoring purposes. If necessary, bundle decontamination can be undertaken to assess engineering deterioration to various degrees of thoroughness ranging from 100% equivalent factory final assembly inspection, to partial decontamination operating via a glovebox type of maintenance bag arrangement, examining local points of both shell and tube areas of the bundle. Many lessons from the last five years' experience of PFR will be incorporated into the design of the CDFR and PFR Steam Generators have two very good examples of how the designer can ease or severely handicap the operator in coping with sodium/water leakages. Good, quick access to tube ends is achieved in the existing PFR Evaporator by simply unbolting the steam/water closure head, but on the superheater and reheater hand-caps have

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

    OpenAIRE

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

    2016-01-01

    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...

  15. Evaluation of on-line chelant addition to PWR steam generators. Steam generator cleaning project

    International Nuclear Information System (INIS)

    The investigation of chelating agents for continuous water treatment of secondary loops of PWR steam generators were conducted in two general areas: the study of the chemistry of chelating agents and the study of materials compatability with chelating agents. The thermostability of both EDTA and HEDTA metal chelates in All Volatile Treatment (AVT) water chemistry were shown to be greater than or equal to the thermostability of EDTA metal chelates in phosphate-sulfite water chemistry. HEDTA metal chelates were shown to have a much greater stability than EDTA metal chelates. Using samples taken from the EDTA metal chelate thermostability study and from the Commonwealth Research Corporation (CRC) model steam generators (MSG), EDTA decomposition products were determined. Active metal surfaces were shown to become passivated when exposed to EDTA and HEDTA concentrations as high as 0.1% w/w in AVT. Trace amounts of iron in the water were found to increase the rate of passivation. Material balance and visual inspection data from CRC model steam generators showed that metal was transported through and cleaned from the MSG's. The Inconel 600 tubes of the salt water fouled model steam generators experienced pitting corrosion. Results of this study demonstrates the feasibility of EDTA as an on-line water treatment additive to maintain nuclear steam generators in a clean condition

  16. Ferritic steels for French LMFBR steam generators

    International Nuclear Information System (INIS)

    Austenitic stainless steels have been widely used in many components of the French LMFBR. Up to now, ferritic steels have not been considered for these components, mainly due to their relatively low creep properties. Some ferritic steels are usable when the maximum temperatures in service do not exceed about 5300C. It is the case of the steam generators of the Phenix plant, where the exchange tubes of the evaporator are made of 2,25% Cr-1% Mo steel, stabilized or not by addition of niobium. These ferritic alloys have worked successfully since the first steam production in October 1973. For the SuperPhenix power plant, an ''all austenitic stainless alloy'' apparatus has been chosen. However, for the future, ferritic alloys offer potential for use as alternative materials in the evaporators: low alloys steels type 2,25% Cr-1% Mo (exchange tubes, tube-sheets, shells), or at higher chromium content type 9% Cr-2% Mo NbV (exchange tubes) or 12M Cr-1% Mo-V (tube-sheets). Most of these steels have already an industrial background, and are widely used in similar applications. The various potential applications of these steels are reviewed with regards to the French LMFBR steam generators, indicating that some points need an effort of clarification, for instance the properties of the heterogeneous ferritic/austenitic weldments

  17. Measuring water level in a steam generator

    International Nuclear Information System (INIS)

    A method is provided for determining and controlling steam water level in a steam generator of a nuclear plant, comprising calibrating the water level sensor in terms of velocity head and also adjusting the high level setpoint in terms of a velocity head bias. The water level differential pressure sensor is calibrated so that maximum water level is indicated as that level corresponding to the upper tap level less velocity head at maximum power plant power. The high level set point is calculated as corresponding to the riser level less a velocity head bias in flow path, the bias being calculated as maximum velocity head at maximum velocity minus rider head percentage of span times velocity head at maximum power. (author)

  18. Corrosion Evaluation and Corrosion Control of Steam Generators

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  20. Corrosion Evaluation and Corrosion Control of Steam Generators

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-06-15

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

  1. LMFBR steam generators in the United Kingdom

    International Nuclear Information System (INIS)

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

  2. Regulations and proactive management of ageing steam generators in Canada

    International Nuclear Information System (INIS)

    Effective ageing management programs of key safety-related structures, systems, and components (SSCs) are an efficient means for ensuring the long-term safe and reliable operation of nuclear power plants. In the early days of design of nuclear power plants, it was assumed that the operational life cycle of steam generators would be the same or similar to that of other key components in the reactor primary heat transport system. However, widespread degradation of the steam generator tubing that has occurred at a number of plants has shown that this original assumption was incorrect or at least too optimistic. Observed degradations can be attributed to a number of factors ranging from shortcomings in the design codes manufacturing processes, or water chemistry, and unanticipated mechanisms of material and component degradation resulting from high temperature, high fluid flow, cycling loads and presence of corrosive species. The licensees have responded to this challenge with extensive inspection and maintenance programs, supported by research and development in the areas of corrosion and mechanical degradation of tubes and internals, chemistry, thermal hydraulics, fouling, inspection and cleaning, and specialized inspection tools. This paper discusses the Canadian Nuclear Safety Commission approach towards ensuring that licensees operate and maintain their plants in a safe operational condition. It briefly describes elements of the CNSC requirements and the overall regulatory oversight process to achieve these goals. The paper also discusses the known degradation mechanisms in Canadian steam generators and describes the requirements in place to ensure licensees sufficiently monitor the condition of their SSCs and appropriately disposition the results of inspections. Particularly, the tube degradation is a major driving force for development of CANDU specific fitness-for-service guidelines, and for specialized inspection and monitoring technology to control steam

  3. Progress of steam generator ageing management of Chinese NPPs

    International Nuclear Information System (INIS)

    In China, the first two NPPs have operated over 15 years (QNPC) and 13 years (Dayabay). Now, most of Chinese NPPs have started their ageing management project. Steam generator (SG) is one of key safety important components for PWR and CANDU, which is selected as the first group of components to develop component ageing management system. QNPC and TQNPC have made some progress in this area. The basic SG ageing management system has been set up, and some application has been under way. To set up SG basic ageing management system, we start with (1) SG ageing mechanism analysis (2) Developing SG ageing management programme (3) Developing SG ageing management information system, SGAMDB (4) Ageing degradation assessment and remained life evaluation SG ageing mechanisms are classified as function degradation and material/structure degradation. The specific tables with unified format are used to analyze the ageing mechanism. The numbering requirements for ageing mechanism are developed to manage them in ageing information system. All ageing mechanisms are graded I and II and III based on its feasibility, impact on safety and operation experience, so that the key mechanisms are managed more strictly. SG ageing management programme (SGAMP), which is fundamental document for ageing management, provides the requirements for working process, what to do and when to do. SGAMP reviewing and improving requirements are as well included. SG ageing management database (SGAMDB) is the first component ageing management database for Chinese NPP, which have been installed and are in operation in QNPC and TQNPC. SG ageing degradation assessment and remained life evaluation are based on ageing mechanism, which includes integrity of tube bundle, integrity and remained fatigue life of key assemblies, and heat transfer capability degradation. Most NPPs at Chinese mainland have started their plant ageing and life management projects. Steam generator (SG), which is one of safety important

  4. Steam generator replacement project in 2000

    International Nuclear Information System (INIS)

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

  5. Hideout in steam generator tube deposits

    International Nuclear Information System (INIS)

    Hideout in deposits on steam generator tubes was studied using tubes coated with magnetite. Hideout from sodium chloride solutions at 279 degrees C was followed using an on-line high-temperature conductivity probe, as well as by chemical analysis of solution samples from the autoclave in which the studies were done. Significant hideout was observed only at a heat flux greater than 200 kW/m2, corresponding to a temperature drop greater than 2 degrees C across the deposits. The concentration factor resulting from the hideout increased highly non-linearly with the heat flux (varying as high as the fourth power of the heat flux). The decrease in the apparent concentration factor with increasing deposit thickness suggested that the pores in the deposit were occupied by a mixture of steam and water, which is consistent with the conclusion from the thermal conductivity measurements on deposits in a separate study. Analyses of the deposits after the hideout tests showed no evidence of any hidden-out solute species, probably due to the concentrations being very near the detection limits and to their escape from the deposit as the tests were being ended. This study showed that hideout in deposits may concentrate solutes in the steam generator bulk water by a factor as high as 2 x 103. Corrosion was evident under the deposit in some tests, with some chromium enrichment on the surface of the tube. Chromium enrichment usually indicates an acidic environment, but the mobility required of chromium to become incorporated into the thick magnetite deposit may indicate corrosion under an alkaline environment. An alkaline environment could result from preferential accumulation of sodium in the solution in the deposit during the hideout process. (author)

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

    Directory of Open Access Journals (Sweden)

    Podlasek Szymon

    2016-01-01

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

  7. 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.

  8. 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.

  9. Repair technology for steam generator tubes

    International Nuclear Information System (INIS)

    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

  10. Ultrasonic examination techniques for steam generator tubing

    International Nuclear Information System (INIS)

    Ultrasonic examination techniques for FBR steam generator tubing have been developed which provide high accuracy and high inspection rates. Typical dimensions of the steam generator tubing are 24.2 mm inner diameter and approximately 80 m length, and all tubes are helically wound. In order to perform flaw detection at high speed, three types of electronic scanning multi-array transducer units for axially and circumferentially oriented flaws in the tube and for tube-wall thinning were incorporated into one probe. With this probe, notched flaws of 0.17 mm depth (5% of the wall thickness) and 3 mm length, and tube-wall thinning of 0.2 mm could be detected in the experiments. A probe transportation system using water flow has also been developed. This system is capable of carrying the probe through the entire length of a helically coiled tube at a rate of 4-16 m/min. Inspection tests using these techniques show that flaws can be successfully detected at an inspection rate of 4 m/min. (author)

  11. Advanced Eddy current NDE steam generator tubing

    International Nuclear Information System (INIS)

    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

  12. Heysham II/Torness AGR steam generator

    International Nuclear Information System (INIS)

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

  13. Long-Term Trends in Radionuclide Distribution in the Vicinity of a CANDU Nuclear Generating Station

    International Nuclear Information System (INIS)

    The Point Lepreau monitoring programme was established in 1978 to assess the environmental impact of radioactive, thermal and chemical releases from the Point Lepreau Nuclear Generating Station (NGS), a 600 MW CANDU reactor, located on the Bay of Fundy in eastern Canada. The programme was designed on a mass-balance approach whereby measurements of radionuclides on samples from the major environmental reservoirs (sea water, fresh water, sediments and marine, terrestrial and aquatic flora and fauna and atmospheric media) would be used to determine contaminant transport rates through different environmental phases. Environmental radioactivity levels measured in the 14 years since the reactor became operational have been compared with pre-operational levels to assess the implications of operating a CANDU nuclear reactor in a coastal region and to determine the critical parameters governing the long-term transport of radionuclides through the environment. Tritium is routinely measured in the marine, terrestrial and atmospheric components of the programme and has become a useful tool in assessing local meteorological influences on atmospheric radionuclide distributions. The environmental monitoring programme has provided an important and timely perspective on environmental radionuclide transport through eastern Canada from globally significant phenomena such as nuclear weapons fallout and the 1986 Chernobyl accident, thereby illustrating the potential advantages inherent in cost-effective, long-term environmental surveillance programmes. (author)

  14. Behaviour of steam generator tubing in the presence of silicon compounds

    International Nuclear Information System (INIS)

    The chemical reactions that take place between the components of concentrated solutions generate an aggressive environment. The presence of this environment and of the tubesheet crevices lead to localized corrosion and the affected tubes cannot ensure effective heat transfer between the fluids of the primary and secondary circuits. Thus, it becomes necessary to understand the corrosion process that occurs on the CANDU steam generator secondary side. The purpose of this paper is the assessment of the corrosion behaviour of the tube material Incoloy 800 at the normal secondary circuit parameters (temperature = 260 C, pressure = 5.1 MPa). The testing environment was demineralized water containing silicon compounds, at a pH = 9.5 regulated with morpholine and cyclohexylamine. The paper presents the results of metallographic, electronic microscopy and X-ray diffraction examinations, as well as the results of electrochemical measurements. (orig.)

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

    International Nuclear Information System (INIS)

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

  16. Evaluation of surrogate boilers for steam generators. Final report

    International Nuclear Information System (INIS)

    Corrosion in PWR systems is a continuing problem that can result in significant expenditures for inspection, repair, and replacement of steam generators as well as for power replacement during outages. Model boilers operated in parallel to the steam generator may provide a useful tool for monitoring and studying steam generator corrosion and corrosion prevention processes. The potential benefits of such boilers as well as several conceptual boiler design alternatives are described, and approximate costs for fabrication and operation of such systems are presented

  17. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations for the workers and the public

  18. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on worker and public safety, operating performance and costs, and reliability of system components

  19. CANDU 9 safety improvements

    International Nuclear Information System (INIS)

    The CANDU 9 is a family of single-unit Nuclear Power Plant designs based on proven CANDU concepts and equipment from operating CANDU plants capable of generating 900 MWe to 1300 MWe depending on the number of fuel channel used and the type of fuel, either natural uranium fuel or slightly enriched uranium fuel. The basic design, the CANDU 9 480/NU, uses the 480 fuel channel Darlington reactor and employs Natural Uranium (NU) fuel Darlington, the latest of the 900 MWe Class CANDU plants, consists of four integrated units with a total output of approximately 3740 MWe located in Ontario, Canada. AECL has completed the concept definition engineering for this design, and will be completing the design integration engineering by the end of 1996. AECL's design philosophy is to build-in product improvements in evolutionary from the initial prototype plants, NPD and Douglas Point, to today's operating CANDU's construction projects and advanced designs. CANDU 9 safety design follows the evolutionary path, including simple improvements based on existing well-proven CANDU safety concepts. The CANDU 9 builds on the experience base for the Darlington reference plant, and on AECL's extensive safety design experience with single unit CANDU 6 power plants. The latest CANDU 6 plants are being built in Korea by KEPCO at Wolsong 2,3 and 4. The Safety improvements for the CANDU 9 power plant are intended to provide the owner-operator with increased assurance of reliable, trouble-free operation, with greater safety margin, with improved public acceptance, and with ease of licensibility

  20. Corrosion aspects in steam generators of nuclear power plants

    International Nuclear Information System (INIS)

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

  1. 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.

  2. Steam generator operating experience update, 1982-1983

    International Nuclear Information System (INIS)

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

  3. Dismantling of the 50 MW steam generator test facility

    International Nuclear Information System (INIS)

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

  4. 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

  5. ROSA III, a third generation steam generator service robot targeted at reducing steam generator maintenance exposure

    International Nuclear Information System (INIS)

    The Westinghouse Nuclear Service Division has employed two delivery robots for the past eight years. The simplest is a two degree of freedom robot (WL-2) that has a design goal of delivering Eddy Current Acquisition and Mechanical Plugging services. The delivery capability of this robot is 111 N at a reach of 2.36 M. The robot is somewhat limited because two degrees of freedom cannot provide general end point approach or orientation alignments for maintenance tools which require cam-locks. But for delivery of the above two services the design goal is very much satisfied. The second robot is ROSA I, its design goal is to provide the heavy duty maintenance operations on steam generators and reactor vessels. ROSA I has six degrees of freedom, has a reach of 2.36 M, and a load capacity of 222 N. The actuators of ROSA I are electric motor driven through a 200/1 harmonic drive. There are 677 N-M actuators at axes 1, 2 and 3 and 338 N-M actuators at axes 4, 5 and 6. These are arranged in a elbow configuration with axes 2, 3 and 4 providing the elbow shape. The services provided by ROSA I include Eddy Current, Mechanical Plugging, Sleeving, U-bend and Support Plate Heat Treating, Plug Removal and Tube Removal. ROSA I, having six degrees of freedom, is capable of generalized tool placement and orientation to any point in space within its reach envelope. ROSA II is a extension of ROSA I. A mast, carriage and rotating base were added to provide inspection and maintenance services on reactor vessel shells and nozzles. ROSA III is the third generation of maintenance and inspection robots designed, manufactured and operated by Westinghouse. An integrated system approach built around a network architecture has led to many areas of improvement. The single 16 mm digital network cable replaces the bulky analog cables, reducing setup time and containment penetration requirements. The robot arm was configured specifically for steam generator service and has the capability of remote

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

    OpenAIRE

    Gustavo Boroni; Alejandro Clausse

    2011-01-01

    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...

  7. Maintenance and repair of LMFBR steam generators. II. Design philosophy of maintenance and repair for Super Phenix 1 steam generators

    International Nuclear Information System (INIS)

    The Creys Malville steam generators have been the subject of a number of papers presented at this congress. However, the general design of the components are outlined, and the in-service monitoring systems and protective devices they are equipped with are briefly described. The methods used in the event of leakage, are described for: leak location, steam generator inspection, steam generator repair, and putting the affected loop back into service

  8. Description of a program for steam generators

    International Nuclear Information System (INIS)

    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)

  9. Water chemistry of breeder reactor steam generators

    International Nuclear Information System (INIS)

    The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed

  10. Internal ultrasonic testing of steam generator tubes

    International Nuclear Information System (INIS)

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

  11. 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.

  12. Crevice chemistry control in PWR steam generators

    International Nuclear Information System (INIS)

    To establish a basis for predicting and eventually controlling crevice solution chemistry in PWR steam generators, hideout tests were performed at several units. Results indicated that impurity hideout rates varied with the species and with bulk water concentration. Field evaluations of crevice impurity inventory models based on the hideout rate data indicated that further model refinements were necessary, e.g., more frequent quantification of the relation of hideout rates and bulk water concentration. An alternate crevice inventory model based on a real-time mass balance approach also began to be pursued. Modeling results currently are being used at several PWRs to establish a chloride injection rate consistent with development of a near neutral crevice solution to minimize IGA/SCC. Hideout return data are being used to independently establish predictions of crevice chemistry and to substantiate the hideout rate and mass balance model predictions

  13. Combustion engineering: steam generator tube bending practices

    International Nuclear Information System (INIS)

    The tube bending practices and procedures employed by Combustion Engineering (CE), when bending inconel tubing is discussed. CE has two different type tube geometries in the steam generator. The innermost tubes are 1800 U-bends while the majority of the tubes have two (2) 900 bends with a straight leg between these 900 bends. The first 18 rows have U-bends (2 1/2'' to 11''R), while the remaining tubes have the double 900 geometry. All double 900 bends are bent to a 10'' radius. This presentation will address the following important parameters necessary to achieve a high quality bent tube: fabrication requirements at the tube mill; tube bending equipment; tube bending operation; inspection and final preparation; and packaging

  14. Steam generator issues in the United States

    International Nuclear Information System (INIS)

    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

  15. Comments on US LMFBR steam generator base technology

    International Nuclear Information System (INIS)

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

  16. Steam generators life extension experience at KANUPP

    International Nuclear Information System (INIS)

    Karachi Nuclear Power Plant (KANUPP) commissioned in 1972 has been operating under PLEX since Jan 2004, after completion of 30 years of its design life. It is planned to extend its life at least by another 15 years after necessary upgrades and re-licensing outages (RLO) by local regulators. KANUPP has six steam generators (SGs), with half-inch diameter Monel-400 tubes. In-service inspection is being carried out regularly in compliance with plant and regulatory requirements (CSA N285.4 for tubes and ASME codes for shell and internals). Degradation is prominent in tubes under sludge from pitting/wastage and denting at first tube support plate (Corten steel) in all the six SG units. Up til now there has been only one instance of a tube leakage and so far 99 tubes (i.e. 1.2% of the total tubes) have been plugged based on wall thinning and severely dented at first tube support plate. A regular monitoring program that is in place includes inspection of tubes, primary and secondary internals, shell, supports and connection welds. Plugging criteria for tubes is ≥ 40% for wall thinning and ≤ 0.250 inch opening for denting using stabilizer bars. An extensive monitoring program for condition assessment is in hand to keep a watch on the rate and morphology of degradation mechanisms and surveillance on susceptible areas unless remedial and control measures are effectively in place. KANUPP steam generators have so far undergone partial water lancing in 2000, hydraulic analysis study, mechanical integrity and comprehensive inspection of tube, overall condition assessment, internals, shell welds and supports inspection. (author)

  17. Steam generator tube inspection in Japan

    International Nuclear Information System (INIS)

    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

  18. Chooz-A Steam Generators Characterization

    International Nuclear Information System (INIS)

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

  19. 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.

  20. Comparison of steam generator methods in PISC

    International Nuclear Information System (INIS)

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

  1. Design improvement and test verification of steam flow limiter of steam generator

    International Nuclear Information System (INIS)

    Background: Steam flow limiter is an important device of steam generator in nuclear power plant. It limits the steam flow during the event of steam line break. However, it is required that the steam flow limiter has low pressure loss during normal operation of steam generator. Purpose: The aim is to design a steam flow limiter with lower pressure loss. Methods: An improved design of steam flow limiter is developed by increasing the number of Venturies from 7 to 19. Two test models of steam flow limiters of traditional design and improved design are tested. Results: The pressure loss factor of the traditional design test model is 6.9. The pressure loss factor of the improved design test model is 4.4. Conclusion: Based on the same total throat flow area, it is verified by tests that the pressure loss of steam flow limiter containing 19 Venturis is significantly lower than that containing 7 Venturis. The pressure loss calculation method is verified simultaneously. (authors)

  2. Reactor physics innovations in ACR-700 design for next CANDU generation

    International Nuclear Information System (INIS)

    ACR-700 is the 'Next Generation' CANDU reactor, aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs. A key element of cost reduction is the use of H20 as coolant and Slightly Enriched Uranium fuel in a tight D20- moderated lattice. The innovations in the ACR core physics result in substantial improvements in economics, as well as significant enhancements in reactor controllability and waste reduction. Fuel design is chosen to balance fuel performance, cost, and reactor-physics characteristics. Full-core coolant void reactivity in ACR-700 is about 3 mk. Power coefficient is substantially negative. Discharge fuel burnup is about three times the current natural-uranium discharge burn-up. The result is a core design which provides a high degree of inherent safety with attractive power-production efficiency and stability. (author)

  3. 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.

  4. 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

  5. Generator of steam plasma for gasification of solid fuels

    Science.gov (United States)

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

    2013-12-01

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

  6. 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 are...

  7. Flow-induced vibration study in the LOFT steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Carmichael, C.F.

    1977-10-18

    The consequences of flow induced vibration in the LOFT steam generator were studied. Tube-baffle contact and fretting wear are expected to occur while tube-tube contact is not predicted. The LOFT steam generator is, in all probability adequate from a fluid induced viewpoint for the scheduled service of the LOFT facility at power. 37 refs., 5 figs., 3 tabs.

  8. Examination of a steam generator tube removed from Maine Yankee

    International Nuclear Information System (INIS)

    Non-destructive and destructive examinations performed on an Alloy 600 steam generator tube with a circumferential indication confirmed that primary water stress corrosion cracking (PWSCC) had occurred in high temperature final mill annealled material. The tube material generally has low susceptibility to PWSCC. Additional PWSCC in this material is expected but would not by itself lead to steam generator replacement

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

    International Nuclear Information System (INIS)

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

  10. Polymer dispersant addition qualification program for steam generator use

    International Nuclear Information System (INIS)

    One of the major problems encountered in the steam generators of nuclear power plants is the accumulation of sludge. Sludge deposition affects the thermal efficiency of a steam generator and has often been implicated in corrosion of steam generator tubing at the tubesheet and tube support plate locations. The majority of the sludge in steam generators is composed of iron oxide (Fe3O4) with other metal oxides, depending upon the materials used in the construction of the pre-boiler cycle. Over time, even with careful system cycle water chemistry control, corrosion products may accumulate in the steam generator. Several utilities have undertaken programs to reduce the amount of sludge in their steam generators. In particular, Commonwealth Edison Company (ComEd) in conjunction with the Electric Power Research Institute (EPRI) and Babcock and Wilcox have defined a 'generic' chemical addition qualification program. The basic details of this program is reported here. This program will determine the effect of chemical additives (for instance polymer addition), to the feedwater, on the materials of construction of steam generators, with emphasis on the tubing material. It will also determine if the steam generator pressure boundary and internal components will be adversely affected during their design life by the use of the candidate chemical additives. (O.M.)

  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. Status of steam generators for NPPs with VVE-1000

    International Nuclear Information System (INIS)

    The status has been set forth on PGV-1000 (1000M) steam generators for nuclear power stations with VVER-1000. It has been shown that as a result of comprehensive investigations they mainly understood the causes and mechanisms of failures of the '' cold'' collectors of these steam generators and developed and implemented the measures providing an increase in SG operational reliability and safety. (Author)

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

    International Nuclear Information System (INIS)

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

  14. Fatigue analysis of steam generator cassette parts based on CAE

    International Nuclear Information System (INIS)

    Fatigue analysis has been performed for steam generator nozzle header and tube based on CAE. Three dimensional model was produced using the commercial CAD program, IDEAS and the geometry and boundary condition information have been transformed into input format of ABAQUS for thermal analysis, stress analysis, and fatigue analysis. Cassette nozzle, which has a complex geometry, has been analysed by using the three dimensional model. But steam generator tube has been analysed according to ASME procedure since it can be modelled as a two dimensional finite element model. S-N curve for the titanium alloy of the steam generator tube material was obtained from the material tests. From the analysis, it has been confirmed that these parts of the steam generator cassette satisfy the lifetime of the steam generator cassette. Three dimensional modelling strategy from the thermal analysis to fatigue analysis should be implemented into the design of reactor major components to enhance the efficiency of design procedure

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

    International Nuclear Information System (INIS)

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

  16. Equipment transporter for nuclear steam generator

    International Nuclear Information System (INIS)

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

  17. LPGC, Levelized Steam Electric Power Generator Cost

    International Nuclear Information System (INIS)

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

  18. Improvement factors for steam generator tubing alloys

    International Nuclear Information System (INIS)

    Predictions of reliability gains associated with the use of advanced alloys have been made in the past through the use of improvement factors. Improvement factors for thermally treated Alloy 600 (Alloy 600TT) and thermally treated Alloy 690 (Alloy 690TT) steam generator tubing were previously developed and have been used in the most recent revision of the EPRI Secondary Water Chemistry Guidelines. However, due to the long expected failure times relative to field experience, field-experience-based estimates of these improvement factors continue to be overly conservative (as shown by the absence of wide spread in-service cracking of these materials). A recent study updated the previously developed improvement factors associated with the use of advanced alloys. This paper will discuss the development of relative improvement factors for Alloy 600TT, Alloy 690TT, and Alloy 800 nuclear grade (Alloy 800NG) with respect to mill annealed Alloy 600 (Alloy 600MA) steam generator tubing. The various uses which are appropriate for these improvement factors will be discussed. This presentation focuses on primary side tube degradation (PWSCC), although this project also addressed secondary side tube degradation (ODSCC). The following four techniques were used to assess the performance of the Alloy 600TT, Alloy 690TT, and Alloy 800NG relative to that of Alloy 600MA: Field data on tube degradation were evaluated using statistical techniques, based on plant population Weibull/Weibayes analyses, similar to those employed in the past and reviewed by industry experts as part of the EPRI guidelines revision process. This paper presents updated improvement factors based on further accumulation of operating experience with Alloy 600TT, Alloy 800NG, and Alloy 690TT; Field data on tube degradation were evaluated using alternative statistical techniques which are not as overly conservative as those used in the past; Field data on tube plug cracking were evaluated to compare the performance

  19. Direct steam generation in line-focus solar collectors

    Science.gov (United States)

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

    1983-01-01

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

  20. Mechanical design of a sodium heated steam generator

    International Nuclear Information System (INIS)

    FBTR steam generator is a once through type unit consisting of four 12.5 MW thermal modules generating a total of 74 tons per hour of steam at 125 bar and 4800C. This paper outlines the mechanical design of such type of steam generator with emphasis on special design problems associated with this type of sodium to water steam heat exchanger, namely, thermal cycling of transition zone where nucleate boiling changes over to film boiling, application of pressure vessel design criteria for transient pressures, thermal stress evaluation resulting from differential expansion between shell and tube in this typical configuration, sodium headers support design, thermal sleeve, design, thermal shock analysis in thick tubes, thermal stress resulting from stratification and stability of expansion bends against vibration. Some of the possible design changes for the future large size steam generator are outlined. (author)

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

    International Nuclear Information System (INIS)

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

  2. 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.

  3. Welding for the CRBRP steam generators

    International Nuclear Information System (INIS)

    The rationale for selecting weld design, welding procedures and inspection methods was based upon the desire to obtain the highest reliability welds for the CRBRP steam generators. To assure the highest weld reliability, heavy emphasis was placed on the control of material cleanliness and composition substantially exceeding the requirements of the ASME Code for 2-1/4Cr-1Mo. The high tube/tubesheet weld quality was achieved through close material control, an extensive weld development program and the selection of high reliability welding equipment. A pre-production run involving 300 welds demonstrated the ability of the manufacturing team to work with the methods and tools provided during the development stage. Prior to the initiation of manufacturing, control of the process and equipment was demonstrated by a 52 weld qualification run. Shell and nozzle weld fabrication using TIG, MIG, and submerged arc procedures are also being controlled through precise specifications, including preheat and postheat programs, together with radiography and ultrasonic inspection to ascertain the weld quality desired. Details of the tube/tubesheet welding and shell welding are described and results from the weld testing program are discussed. (author)

  4. CANDU improvement

    International Nuclear Information System (INIS)

    The evolution of the CANDU family of nuclear power plants is based on a continuous product development approach. Proven equipment and system concepts from operating stations are standardized and used in new products. Due to the modular nature of the CANDU reactor concept, product features developed for CANDU 9 can easily be incorporated in other CANDU products such as CANDU 6. Design concepts are being developed for advanced CANDU 6 or larger advanced CANDU, depending on the number of fuel channels and the fuel cycle selected. This paper provides a description of the design improvements being incorporated in CANDU 9 and further design enhancements being studied for future incorporation in CANDU 6 or larger advanced CANDU meeting the requirements of future CANDU owners. The design enhancement objectives are: To improve operational simplicity by applying modern information technology; to improve safety in a cost effective way; to improve system and component reliability and to increase plant life; to improve economics and to reduce owners' risks during all phases of a project using up-front licensing, an improved engineering process and project tools during design, construction and operation; to continue to exploit the neutron economy of CANDU with the development of advanced fuels and fuel cycles. (author)

  5. V. C. Summer Nuclear Station steam generator replacement

    International Nuclear Information System (INIS)

    The Virgil C. Summer Steam Generator Replacement Project involved the first-ever replacement of an existing steam generator with a different and later vintage component from the original equipment manufacturer (OEM), in this case Westinghouse Model Delta-75 steam generators to replace Westinghouse Model D-3 steam generators, which had been plaguing South Carolina Electric and Gas Company's (SCE and G) Virgil C. Summer Nuclear Plant since shortly after initial operation in 1982. This project also involved the first use of laser metrology technology for steam generator-to-reactor coolant system severance cutting, machining, and component fitup and the first use of an impregnated sponge blast media for reactor coolant system pipe end decontamination. The sequence of events leading to the decision to replace steam generators and during the replacement process is described. Intensive planning and teamwork, combined with input from SCE and G and the use of mockups to train the work force in a simulated radiological environment, were instrumental in achieving world-record schedule performance and setting a new US record for the lowest accumulated radiation exposure during a steam generator replacement project while completing the project without a single lost workday case incident

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

    International Nuclear Information System (INIS)

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

  7. Solar installation for process steam generation for a refinery

    Science.gov (United States)

    Clark, L. D.; Hudson, S.; Pytlinski, J. T.; Lumsdaine, E.; Bridgers, F.

    A solar thermal system for steam generation in a refinery is presented. The system is installed in the Southern Union Refinery in Hobbs, New Mexico, U.S.A. The refinery processes 36,000 BPSD of crude oil (42 U.S. gallon barrels of product fuels per steam day). The solar system is a two loop system. A heat transfer oil (Therminol T-55) circulates through an array of parabolic collectors of 936 sq m area while saturated steam at 190 C/12 kg/sq m is generated in the steam generator loop. The steam flow is 658 kg/hr. A data acquisition system (ODAS) was designed and assembled to evaluate the solar system's thermal performance. It is expected that on an annual basis the solar system will provide a thermal process heat equivalent to 93,400 cu m of natural gas.

  8. Next Generation Steam Cracking Reactor Concept

    OpenAIRE

    Van Goethem, M.W.M.

    2010-01-01

    The steam cracking process is an important asset in the hydrocarbon processing industry. The main products are lower olefins and hydrogen, with ethylene being the world's largest volume organic chemical at a worldwide capacity of ~ 120 million tonnes per year. Feed stocks are hydrocarbons such as: ethane, LPG, naphtha's, gas condensates and gas oil. The research goal of this thesis is to search for the intrinsic optimal steam cracking reaction conditions, pushing the olefin yields to the maxi...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Methods of inspecting and repairing steam generator heating tubes

    International Nuclear Information System (INIS)

    In more than eighty interventions within eight years carried out worldwide, ABB Reactor inspected and upgraded steam generators made by various manufacturers. The tools and procedures employed were flexible enough, ensured a high level of positioning accuracy and, therefore, were able to cope with all irregularities of steam generator designs. The experience accumulated in these interventions was used to introduce significant product improvements, such as the microgearing of the removable plug with the steam generator tube, and the special cleaning procedures preceding all welding activities. Special attention was paid to the efficient use of only two types of welds. (orig.)

  12. Cooldown strategies for a steam generator tube rupture event with failure of main steam safety valve

    International Nuclear Information System (INIS)

    This paper provides an evaluation of the thermal-hydraulic response of a pressurized water reactor (PWR) during a steam generator tube rupture (SGTR) event with the failure of a main steam safety valve (MSSV). Operator actions to successfully mitigate the consequences of this SGTR event are proposed. The desired actions are those which provide for control of the affected steam generator water level and minimize radiological doses to the environment. Specifically, the purpose of this paper is to demonstrate the results of differences in operator actions to cooldown the power plant in terms of: (1) dose releases to the environment, (2) control of the affected steam generator level, (3) and optimal reactor coolant system cooldown and depressurization

  13. KANUPP operation with four out of six steam generators

    International Nuclear Information System (INIS)

    This paper highlights the experience at Karachi Nuclear Power Plant of operation with 4 out of 6 steam generators in service. Removal of two steam generators, one from each loop of the normal circuit, was necessitated due to the development of leak in one of the steam generators. The normal approach of leak search, plugging of tube and, subsequent ISI of steam generator could not be attempted mainly because of lack of tools, expertise, and experience in the relevant field. Suitable modifications were, however, carried out to isolate the faulty steam generator from the circuit and restart the plant with 4 steam generators instead of 6. The primary pumps in operation were also reduced from 3 pairs to 2 pairs. The modifications required a series of studies, analyses and changes to piping work in the steam boiler feed water circuit. Comprehensive stress analysis was carried out to make sure that steam and PHT headers can withstand the uneven expansion of hot and cold pipings steam generators. The plant rating was reduced to ensure that design criteria are not violated with the new configuration. Various trip limits and set points were adjusted accordingly. A number of special commissioning tests were done to validate the theoretical predictions. Finally the Plant was restarted and connected to the grid on Jan. 03, 1991 and loaded to 50 MWe (45% R.P). The present modified mode of operation is considered temporary and all efforts and resources, both indigenous and international, are being mobilized to carry out leak search, ISI, plugging the leaking tube and restoring the plant to normal configuration for high power operation. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  16. Siemens' steam turbine generator packages for advanced nuclear power plants

    International Nuclear Information System (INIS)

    Despite the current economic crisis and the increasing share of renewable energy, the long term perspective predicts an increasing global demand for nuclear power generation applications. [1] In response to the growing demand for new nuclear power plants (NPPs), Siemens is implementing and further developing a modular platform of half speed steam turbines and generators covering the most relevant power range from 1000 - 1900 MWe. The paper presents details of the Siemens' Steam Turbine Generator Packages (turboset - Fig. 1) consisting of: Modular Steam Turbine Platform: SST-9000 series 4 pole turbo generator fleet SGEN5-4000W The design of the turbosets for NPPs is based on excellent operational experience with Siemens KONVOI saturated steam turbosets together with service and retrofit experience as well as on experience gained during the project execution of the world largest turboset in Olkiluoto 3. (orig.)

  17. Characterization of corrosion products from a steam generator

    International Nuclear Information System (INIS)

    Corrosion products from the secondary side of the steam generator(SG) accumulate at the critical points resulting in the degradation of the performance of the steam generator. An understanding of the nature of these corrosion products is useful in selection of right kind of chemical formulation for cleaning the corrosion products. Samples of corrosion products from the secondary side of the steam generator of the Narora Atomic Power Station (NAPS), operational over a period of two years, was obtained during the lancing of the steam generators of NAPS-l. The samples were analyzed for its components by XRD and TGA techniques and chemical analysis. The results showed the presence of magnetite as a major phase, along with the oxides of copper, silicon and nickel in the sample. Silicon was identified in the quartz phase while the oxides of copper and nickel were found to be non crystalline in nature. (author)

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

    International Nuclear Information System (INIS)

    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.)

  19. Development of probabilistic integrity evaluation program for steam generator tube

    International Nuclear Information System (INIS)

    The primary water stress corrosion cracking of steam generator tube is the principal aging mechanism deteriorating the integrity of steam generator. To predict the period maintaining the integrity of steam genarator tube, the damage degree of tube is statistically predicted by the conservative method using the data for sizes and numbers of cracks collected during the in-service inspection. But, the probabilistic integrity evaluation method has been recently developed and applied to reduce the conservatism of the previsious methods. Therefore, in this paper, the prediction methodology for crack generation and growth is established. Finally, the probabilistic integrity evaluation program predicting the failure probability of steam generator tube is developed by Monte Carlo simulation

  20. Steam generator tube integrity program. Phase I report

    International Nuclear Information System (INIS)

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

  1. Corrosion of steams generator tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    The paper summarizes the various corrosion phenomena which have impaired the reliability of PWR steam generators, and reviews the remedies adopted against them: relaxation of residual stresses in the tubes, improved specifications for water chemistry, selection of new materials

  2. Improvements in steam cycle electric power generating plants

    International Nuclear Information System (INIS)

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

  3. 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.

  4. Development of explosive plugging for steam generator tubes

    International Nuclear Information System (INIS)

    This paper reports the development of explosive plugging techniques as a repair method for leaking steam generator tubes. The study has been in progress since 1977 and results confirm that the explosive plugging method is applicable to the steam generator. The following developmental tests were performed. 1. Preliminary test - Kind and amount of explosive, shape of plug, stand-off, initiation method, etc. 2. Evaluation test - Shearing test, measurement of bonded zone (U.T. and microscopic observation), He-leakage test, water pressure test, hardness of plug and tube after plugging, P.T. for plugged tube plate surface, etc. 3. Re-plugging after failed plugging 4. Demonstration plugging of a 50 MW steam generator 5. Thermal shock test 6. Sodium vapour atmosphere test The results of these tests indicate that the explosive plugging method is effective for the repair of leaking tubes in steam generators. (author)

  5. Overview of the United States steam generator development programs

    International Nuclear Information System (INIS)

    The LMFBR steam generator development program of the USA was initiated to support the development of reliable designs and meaningful performance data for these critical components. Since the steam generators include the structural boundary between heated sodium and water, the consequences of small flaws in the materials that form the boundary are significant. Successful development and demonstration of commercial LMFBR power plants requires the consideration of many factors in addition to the design, construction and operation of a particular plant. Additional factors which must be assessed include: economics, reliability, safety, environment, operability, maintainability and conservation of the resources. In terms of the steam generator these items led to the selection of a single wall tube design using a forced recirculating system for the present Clinch River Breeder Reactor. There are strong economic incentives to use a once-through steam generating system in future designs

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

    International Nuclear Information System (INIS)

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

  7. 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.

  8. Dynamic and control of a once through steam generator

    International Nuclear Information System (INIS)

    This paper presents a non linear distributed parameter model for the dynamics and feedback control of a large countercurrent heat exchanger used as a once through steam generator for a breeder reactor power plant. A convergent, implicit method has been developed to solve simultaneously the equations of conservation of mass, momentum and energy. The model, applicable to heat exchanger systems in general, has been used specifically to study the performance of a once-through steam generator with respect to its load following ability and stability of throttle steam temperature and pressure. (author)

  9. Steam generator channel head decontamination by remote grit blast methods

    International Nuclear Information System (INIS)

    A decontamination technique using a high pressure water spray containing an abrasive grit has been developed and employed in the decontamination of steam generator channel heads. The spray, which is remotely controlled, removes the corrosion product deposits that form on primary system surfaces and reduces the area dose rates. The remote grit blast technique has proven to be a viable method for decontamination of steam generator channel head surfaces

  10. Modelling of a Coil Steam Generator for CSP applications

    OpenAIRE

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

    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 are developed to optimize the behavior of the system in different start-up scenarios. The results improve the effective life time (ELT) of the CSG, the thermal exibility of the overall CSP plant to have...

  11. Nuclear steam generator inspection and repair using robotics

    International Nuclear Information System (INIS)

    Although the steam generators in pressurized water reactors were not originally envisioned as high maintenance components, quite the opposite has been experienced. Steam generator work, including periodic inspection, tube plugging, sleeving, machining and welding not only represents a sizeable portion of an outage schedule, but can be a major contributor to personnel radiation exposure. This last area, radiation exposure, is one of the primary reasons for development of the remotely operated equipment discussed in the paper

  12. Development of data management system for steam generator inspection

    International Nuclear Information System (INIS)

    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)

  13. Design and construction of a steam generator with feedback

    International Nuclear Information System (INIS)

    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

  14. Second international seminar of horizontal steam generator modelling

    International Nuclear Information System (INIS)

    The Second International Seminar on Horizontal Steam Generator Modelling was arranged to continue the international cooperation that was started during the first seminar in March 1991. The main topics of the seminar were: (1) further experimental results on horizontal steam generator behaviour, (2) resent developments in modelling, (3) results on analyses and studies on primary-to-secondary side leakages, and (4) results of the common exercise calculations

  15. 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.

  16. Automated Diagnosis and Classification of Steam Generator Tube Defects

    International Nuclear Information System (INIS)

    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

  17. Steam generator replacement: a story of continuous improvement

    International Nuclear Information System (INIS)

    This paper provides a review of the history of steam generator replacement in the US focusing on the last five years. From the early replacements in the 1980s, there have been major technology improvements resulting in dramatically shorter outages and reduced radiological exposure for workers. Even though the changes for the last five years have been less dramatic, the improvement trend continues. No two steam generator replacement (SGR) projects are the same and there are some major differences including; the access path for the components to containment (is a construction opening in containment required), type of containment, number of steam generators, one piece or two piece replacement, plant type (Westinghouse, CE or B and W) and plant layout. These differences along with other variables such as delays due to plant operations and other activities not related to the steam generator replacement make analysis of performance data difficult. However, trends in outage performance and owner expectations can be identified. How far this trend will go is also discussed. Along with the trend of improved performance, there is also a significant variation in performance. Some of the contributors to this variation are identified. This paper addresses what is required for a successful outage, meeting the increasing expectations and setting new records. The authors will discuss various factors that contribute to the success of a steam generator replacement. These factors include technical issues and, equally important, organizational interface and the role the customer plays. Recommendations are provided for planning a successful steam generator replacement outage. (author)

  18. Next Generation Steam Cracking Reactor Concept

    NARCIS (Netherlands)

    Van Goethem, M.W.M.

    2010-01-01

    The steam cracking process is an important asset in the hydrocarbon processing industry. The main products are lower olefins and hydrogen, with ethylene being the world's largest volume organic chemical at a worldwide capacity of ~ 120 million tonnes per year. Feed stocks are hydrocarbons such as: e

  19. Steam generator tube fretting - Darlington NGS experience

    International Nuclear Information System (INIS)

    Early signs of tube fretting in the U-bend region of Darlington NGS Steam Generators (SGs) were observed during the metallurgical examination of the removed peripheral tube U-bend sections from Unit 4 SG3 in 1995. During a forced outage in early 1998, Eddy Current (ECT) tube inspections in Unit 2 SG4 revealed more extensive fretting of the tubes at the U-bend AVB support locations. Subsequently in the period of 1999-2001, planned Eddy Current tube inspections have been carried out in all units covering all SGs. These inspections have revealed considerable U-bend tube fretting with a number of these fret depths in excess of 40% tw. Evaluation of the ECT and UT results, in conjunction with engineering assessment of the SG design and construction, have determined tube fretting in the U-bend region as an active and reportable degradation mechanism in these SGs. To date, all 16 Darlington SGs have undergone a major ECT inspection. In these inspections as a minimum, the identified fretting region of the U-bend has been adequately covered. Analyses of the inspection results have been carried out to provide trends and observations of the fretting in the U-bend. These showed the fretted U-bend tubes to be localized in the area bounded by Rows 70 and above, and Columns 39 to 83 which has been defined as the 'Area at Risk' of U-bend fretting for Darlington SGs. In the distribution of the frets at the U-bend support locations, they showed a strong biasing of the fretting towards the cold leg supports with the mean centered a third the way between CU4 and CU3. A general understanding of the 'Root Cause of Fretting' shows it to be associated with tube clearance, which invariably results and acts together with conditions of insufficient support preload. While the fretting by tube tends to exhibit a certain degree of randomness, the fretting remains localized to the 'Area at Risk'. This offers a unique opportunity of localized corrective measures that are both simpler in design

  20. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU (CANada Deuterium Uranium) Pressurized Heavy Water (PHW) type of nuclear-electric generating station was developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper summarizes Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components, and nuclear safety considerations to both the workers and the public

  1. Numerical simulation on steady operation characteristic of the steam generator

    International Nuclear Information System (INIS)

    Based on the four-equation drift flux model, this paper establishes a one-dimension distribution model for the vertical U tube steam generator. The model considers the area of the primary side, the secondary side, the U tube and the steam dome. Firstly, the discrete equations are obtained by using the first order difference method with the staggered grid, and solves with iteration by applying intersected calculation of thermal and hydraulic process. This work is compiled into a simulated program with the MATLAB software. Applying the program to simulate the thermal-hydraulic parameters under steady state operation of Qinshan NPP steam generator (SG), and the calculation is compared with the RELAP5 program. Finally, the operation characteristics of steam generator under 100%, 75%, 50%, 30%, 15% powers are computed and analyzed. (authors)

  2. Comparing dynamic responses of recirculating and once-through steam generators for next generation LWRs

    International Nuclear Information System (INIS)

    In this paper two types of steam generators are under consideration for next-generation (pressurized) light water reactors: a recirculating type and a once-through type. The steady-state and dynamic characteristics of these steam generators were compared to facilitate optimization of a particular reactor system design. To compare, the dynamic responses of the two types, as indicated by the feedwater flow, steam generator level, steam flow, steam pressure, steam enthalpy, primary-side pressure and cold-leg temperature, were assessed using Babcock and Wilcox's Modular Modeling System. The once-through steam generator showed a tremendous flexibility to produce superheated steam under diverse conditions (i.e., constant or variable steam throttle pressure and constant or variable average primary temperature) with excellent speed and accuracy in following the load demand. Since the primary and steam sides are closely coupled with the feedwater, the pressurizer should be sized liberally to lessen the sensitivity of the primary response to feedwater upsets and the reliability of the feedwater train should be enhanced. In contrast, the recirculating steam generator must be operated with variable steam throttle pressure and variable primary average temperature, and the speed and accuracy of following the load demand are not as good. While the recirculation provides an effective cushion for the primary and steam sides from feedwater upsets, it also amplifies the level response caused by upsets in steam pressure and feedwater temperature affecting the level controllability and moisture separation performance. The recirculating steam generator should be designed to incorporate features to improve level controllability by constant-inventory control strategy. Also to survive a reactor-coolant pump trip, the design with one reactor-coolant pump per loop should be considered

  3. Steam generator deposit control program assessment at Comanche Peak

    International Nuclear Information System (INIS)

    Comanche Peak has employed a variety of methods to assess the effectiveness of the deposit control program. These include typical methods such as an extensive visual inspection program and detailed corrosion product analysis and trending. In addition, a recently pioneered technique, low frequency eddy current profile analysis (LFEC) has been utilized. LFEC provides a visual mapping of the magnetite deposit profile of the steam generator. Analysis of the LFEC results not only provides general area deposition rates, but can also provide local deposition patterns, which is indicative of steam generator performance. Other techniques utilized include trending of steam pressure, steam generator hideout-return, and flow assisted corrosion (FAC) results. The sum of this information provides a comprehensive assessment of the deposit control program effectiveness and the condition of the steam generator. It also provides important diagnostic and predictive information relative to steam generator life management and mitigative strategies, such as special cleaning procedures. This paper discusses the techniques employed by Comanche Peak Chemistry to monitor the effectiveness of the deposit control program and describes how this information is used in strategic planning. (authors)

  4. Advancing the technologies of CANDU

    International Nuclear Information System (INIS)

    CANDU standard product design will continue to evolve, building upon the success of current operating units. Progressive improvements and enhancements will continue to be made to the CANDU system with heavy water moderated, pressure tube reactor technology of high neutron efficiency, based on the results of advanced technology R and D and operational experience from operating CANDU stations. The directions of development will respond to customer's requirements for economical, reliable and safe generating stations

  5. 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.

  6. 6. CNS international conference on CANDU maintenance. Proceedings

    International Nuclear Information System (INIS)

    The 6th CNS International Conference on CANDU Maintenance took place in Toronto, Ontario on November 16-18, 2003. The theme for the conference was 'Maintenance for Life'. About 270 delegates attended the conference held by the Canadian Nuclear Society. The conference consisted of four parallel sessions, a pattern that continued throughout the conference. Papers were grouped under the following headings: Fuel Channels and End Fittings - Assessments; Fuel Channels and End Fittings - Inspections; Fuel Channels and End Fittings - Maintenance; Fuel Channels and End Fittings - Universal Delivery Machine; Water Upgrading; Performance and Plant Life Improvement; Steam Generator Life Management; Steam Generator Modifications; Steam Generators - Inspections; Steam Generators - Assessments; Maintenance Programs; Feeder Inspections; Feeder Assessment and Mitigation; Valve Maintenance; Instrumentation and Control; Inspection Technology; and Fuel Handling

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

    International Nuclear Information System (INIS)

    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

  8. Heat removal capacity of steam generators under unsteady reflux condensation cooling mode in a pressurized water reactor primary heat transport system

    International Nuclear Information System (INIS)

    In a figure-of-eight configuration of the primary heat transport loop of a nuclear power plant, such as in the CANDU design, experimental observations have shown that reflux condensation in the steam generator tubes may become the major natural circulation mode for removing long-term decay heat under reduced water inventories following a LOCA. The resulting two-phase flow in the reactor core may exhibit oscillatory modes in which the steam flow to each steam generator is oscillatory in nature. The present study was aimed at investigating the transient characteristics of the reflux condensation phenomenon under conditions in which the steam flow was oscillatory, especially with respect to its heat removal capability under such conditions. Various frequencies of oscillations in the inlet steam flow were investigated over a range of average mass flow rates. Experimental results showed that oscillations in the steam flow destabilized the water column formation associated with flooding in the tube and further suggest that the total heat removal capability of reflux condensation would be much higher than the values obtained from steady steam flow experiments. (author)

  9. 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.

  10. 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.

  11. Hydrogen-based power generation from bioethanol steam reforming

    International Nuclear Information System (INIS)

    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. Decontamination of Steam Generator tube using Abrasive Blasting Technology

    International Nuclear Information System (INIS)

    As a part of a technology development of volume reduction and self disposal for large metal waste project, We at KAERI and our Sunkwang Atomic Energy Safety (KAES) subcontractor colleagues are demonstrating radioactively contaminated steam generator tube by abrasive blasting technology at Kori-1 NPP. A steam generator is a crucial component in a PWR (pressurized Water Reactor). It is the crossing between the primary, contaminated, circuit and the secondary waste-steam circuit. The heat from the primary reactor coolant loop is transferred to the secondary side in thousands of small tubes. Due to several problems in the material of those tube, like SCC (Stress Corrosion Cracking), insufficient control in water chemistry, which can be cause of tube leakage, more and more steam generators are replaced today. Only in Korea, already 2 of them are replaced and will be replaced in the near future. The retired 300 ton heavy Steam generator was stored at the storage waste building of Kori NPP site. The steam generator waste has a large volume, so that it is necessary to reduce its volume by decontamination. A waste reduction effect can be obtained through decontamination of the inner surface of a steam generator. Therefore, it is necessary to develop an optimum method for decontamination of the inner surface of bundle tubes. The dry abrasive blasting is a very interesting technology for the realization of three-dimensional microstructures in brittle materials like glass or silicon. Dry abrasive blasting is applicable to most surface materials except those that might be shattered by the abrasive. It is most effective on flat surface and because the abrasive is sprayed and can also applicable on 'hard to reach' areas such as inner tube ceilings or behind equipment. Abrasive decontamination techniques have been applied in several countries, including Belgium, the CIS, France, Germany, Japan, the UK and the USA

  13. Technique for hydraulic calculation of a vertical steam generatoring channel

    International Nuclear Information System (INIS)

    The application of technique for hydraulic calculation of two-phase flow to high pressure steam generatoring channels is described. The assumption, that the effect of disturbed liquid film on pressure losses in steam-water flow is similar to the effect of the wall roughness on pressure losses in one-phase flow, forms the basis of the calculational technique. But in comparison with all known models, field of application of the above technique ranges from underheated water to superheated steam. For this aim, all ranges of mass steam content (relative enthalpy) are divided in 5 typical zones. The calculation is conducted with the help of appropriate dependence according to the range of the initial steam content value (average value along the tube length). The program for calculation of water pressure losses, steam-water flow, overheated steam at raising flow in Vertical tubes is worked out on the basis of the above technique. The program demonstrated proper efficiency in the following interval of working parameters: 68.8 2s), 4x10-3 -3 m; 0 2, 150 deg C <= t <= 5000 deg C. The broad verification program established that the maximal error, while calculating relative pressure losses, does not exceed +-15% for the majority of data

  14. Monitoring of thermal stresses in steam generators

    International Nuclear Information System (INIS)

    An analysis of transient temperature and stress distribution in boiler components during start-up and shut-down operations is presented. Thermal stresses are determined indirectly on the basis of temperature measured at selected points on the outer surface of the construction element. The transient temperature distribution in the whole construction element is calculated and the thermal stresses are determined by using the finite element method. The observed pressure changes are used to calculate the internal-pressure caused stresses. The results of temperature and stress monitoring in selected pressure components of the OP-650 boiler with 650.103 kg/h steam output capacity are analysed

  15. Incentives for improvement of CANDU

    International Nuclear Information System (INIS)

    CANDU is a relatively young technology which has demonstrated many achievements as an electrical power generation system. These achievements include an unsurpassed safety record, high annual and lifetime capacity factors, low electricity cost and a broad range of other performance strengths which together indicate that the CANDU technology is fundamentally sound. Known capabilities not yet fully exploited, such as advanced fuel cycle options, indicate that CANDU technology will continue to pay strong dividends on research, development and design investment. This provides a strong incentive for the improvement of CANDU on a continuing basis

  16. Steam generator assessment for sustainable power plant operation

    International Nuclear Information System (INIS)

    Water and steam serve in the water-steam cycle as the energy transport and work media. These fluids shall not affect, through corrosion processes on the construction materials and their consequences, undisturbed plant operation. The main objectives of the steam water cycle chemistry consequently are: - The metal release rates of the structural materials shall be minimal - The probability of selective / localized forms of corrosion shall be minimal. - The deposition of corrosion products on heat transfer surfaces shall be minimized. - The formation of aggressive media, particularly local aggressive environments under deposits, shall be avoided. These objectives are especially important for the steam generators (SGs) because their condition is a key factor for plant performance, high plant availability, life time extension and is important to NPP safety. The major opponent to that is corrosion and fouling of the heating tubes. Effective ways of counteracting all degradation problems and thus of improving the SG performance are to keep SGs in clean conditions or if necessary to plan cleaning measures such as mechanical tube sheet lancing or chemical cleaning. Based on more than 40 years of experience in steam-water cycle water chemistry treatment AREVA developed an overall methodology assessing the steam generator cleanliness condition by evaluating all available operational and inspection data together. In order to gain a complete picture all relevant water chemistry data (e.g. corrosion product mass balances, impurity ingress), inspection data (e.g. visual inspections and tube sheet lancing results) and thermal performance data (e.g. heat transfer calculations) are evaluated, structured and indexed using the AREVA Fouling Index Tool Box. This Fouling Index Tool Box is more than a database or statistical approach for assessment of plant chemistry data. Furthermore the AREVA's approach combines manufacturer's experience with plant data and operates with an

  17. 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.).

  18. Replacement of steam generators for Embalse NGS - the steam generator cartridge design and manufacturing issues, localization and site assembly challenges

    International Nuclear Information System (INIS)

    Embalse Nuclear Generating Station (Central Nuclear Embalse) was placed in service in 1983 and the outage for refurbishment is foreseen for 2011/2012. Embalse is equipped with four vertical inverted 'U' tube-type Steam Generators (SG) with integral preheater, I-800 tubes and carbon steel internals. Between 2002-2006, the owner assessed the potential for SG life extension; Nucleoelectrica Argentina S.A. (NA-SA) and AECL and a number of actions were completed towards meeting this objective (i.e.: primary divider plate replacement, additional U Bend support and inspection port installation). However, degradation of the tube supports (carbon steel broached plate) and U-bend supports due to Flow-accelerated corrosion (FAC) compromised the possibility for life extension of these Steam Generators. This issue, coupled with the plan to increase the plant power output during the life extension of the station, resulted in the strategic decision by NA-SA, to replace the Steam Generators. Several options were considered for SG replacement: In-situ replacement of the SG tube bundle, the original steam drum to be re-used; Removal and replacement of the entire SG (including the steam drum); and, Replacement of the bottom portion of the SG, i.e. the shell, the tube bundle, the tube sheet, the primary head and its internals and the primary nozzles with a factory assembled cartridge (collectively called the 'SG cartridge'). In this option, the original steam drum would be retained for the extended life. The final decision, based on the recommendations from the Life Assessment Study performed during the Pre-project Condition Assessment Process, is to replace only the Steam Generator cartridges. NA-SA requested AECL's support for the preparation of the Technical Specification for the replacement cartridges, allowing for the higher plant output. This paper presents the design basis for the technical requirements covered in the Technical Specification. The specified requirements include

  19. Explaining the absence of Co-58 radiation fields around CANDU reactor primary circuit

    International Nuclear Information System (INIS)

    Radiation fields from Co-58 are rarely detected in CANDU plants. For example, Ge(Li) surveys of the Inconel 600 steam generators at some CANDU plants may show radiation attributed to Co-58 only early in plant life, and most artefacts removed from the primary circuit later in plant operation show no Co-58 present. However, Pressurized Water Reactor plants experience relatively large fields from Co-58 on their isothermal piping, e.g., steam generator channel head, and steam generators tube sampling programs do show deposits in the tubes with significant Co-58 compared to other radionuclides such as Co-60. CANDU reactors have high concentrations of dissolved iron due to the extensive use of carbon steel for the isothermal piping, e.g., feeders, headers, and steam generator channel heads. A dissolved iron transport diagram that was proposed recently for the primary circuit of CANDU plants has been validated by comparison of predicted deposit weights with plant deposit data from various components. One feature of the diagram is dissolved iron precipitation inside the steam generators tubes. An hypothesis is advanced here in which precipitating dissolved iron is proposed to occlude dissolved nickel. This removal mechanism may prevent the solubility of dissolved nickel from being exceeded anywhere around the primary circuit. In particular, this mechanism could avoid NiO precipitation in the core and the generation of large quantities of Co-58. Using this mechanism along with the known solubility behaviour of NiO with temperature, a dissolved nickel transport diagram has been proposed for CANDU plants. (authors)

  20. Evaluation of a dryer in a steam generator

    International Nuclear Information System (INIS)

    The hooked-vane-type dryer is used in vertical, natural circulation steam generators used in PWR-type nuclear power stations. it separates the fine droplets of water carried by steam so that the steam generator outlet steam moisture is below 0.25%. Such low moisture is demanded to ensure a safe and economic operation of the unit. The dryer is composed of hooked vanes and a draining structure. A series of tests to screen different designs were performed using air-water mixture. The paper presents the results of the investigation of the effect of the number of drainage hooks , the bending angle , distance between two adjacent vanes, and other geometrical parameters on the performance of a hooked-vane-type steam dryer. It indicates that the dryer still works effectively when the moisture of the steam at the dryer inlet changes in a wide range, and that the performance of the dryer is closely related to the geometry of the draining structure . On the basis of the results of this program, a draining structure with an original design was selected and it is presented in the paper. The performance of the selected draining structure is better than that of similar structures in China and abroad. (author)

  1. Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: Steam Generators. 2011 Update

    International Nuclear Information System (INIS)

    generator of the PWR, WWER and CANDU nuclear power plants. The objective of this report is to update and supersede IAEA-TECDOC-981 in order to provide current ageing management guidance for PWR, WWER and CANDU steam generators to all involved in the operation and regulation of nuclear power plants and thus to help ensure steam generator integrity in IAEA Member States throughout their entire service life.

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

    Science.gov (United States)

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

    2008-10-01

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

  3. Steam turbine generators for Sizewell 'B' nuclear power station

    International Nuclear Information System (INIS)

    The thermodynamic cycle of the modern 3000 rpm steam turbine as applied at Sizewell 'B' is presented. Review is made of the factors affecting thermal efficiency including the special nature of the wet steam cycle and the use of moisture separation and steam reheating. Consideration is given to the optimisation of the machine and cycle parameters, including particular attention to reheating and to the provision of feedheating, in order to achieve a high overall level of performance. A modular design approach has made available a family of machines suitable for the output range 600-1300 MW. The constructional features of the 630 MW Sizewell 'B' turbine generators from this range are described in detail. The importance of service experience with wet steam turbines and its influence on the design of modern turbines for pressurised water reactor applications is discussed. (author)

  4. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54... ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except...

  5. Questions raised in developing fast reactor steam generator designs

    International Nuclear Information System (INIS)

    The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

  6. Rolling test of turbine generator by non-nuclear steam

    International Nuclear Information System (INIS)

    The object and procedure of rolling turbine generator test by non-nuclear steam in Qinshan NPP are presented. The steam source of rolling test is compared and chosen. The steam quantity during the rolling test is simply calculated. The limits of parameters of the test and the preparing for rolling of turbine generator are introduced. Procedures of rolling test are divided into three stage of speed: 600 r/min, 1200 r/min, 3000 r/min. Parameters measured as a result of rolling test, such as absolute vibration of axle, metal temperature of bearings, temperature of oil from the exit of bearings and vacuity of condenser etc, show that the design, manufacture, installation and speed control of turbine generator are satisfactory

  7. Strain measurements of nuclear power plant steam generator antiseismic supports

    International Nuclear Information System (INIS)

    The nuclear power plants steam generators have different types of structural supports. One of these types are the antiseismic supports, which are intended to be under stress only if a seismic event takes place. Nevertheless, the antiseismic supports lugs, that are welded to the steam generator vessel, are subjected to thermal fatigue because of the temperature cycles related with the shut down and start up operations performed during the life of the nuclear power plant. In order to evaluate the stresses that the lugs are subjected to, several strain gages were welded on two supports lugs, positioned at two heights of one of the Embalse nuclear power plant steam generators. In this paper, the instrumentation used and the strain measurements obtained during two start up operations are presented. The influence of the plant start up operation parameters on the lugs strain evolution is also analyzed. (author)

  8. Study group meeting on steam generators for LMFBR's. Summary report

    International Nuclear Information System (INIS)

    The Meeting organised by IAEA international working group on fast reactors which considered that the subject of sodium heated steam generators was a topic which needed study by the experts of several disciplines. For example: people who design such steam generators, specialists in the field of sodium water reactions, experts in material and water chemistry and members of the utilities who would be the customers for such units. Besides the exchange of large amount of information, it was considered that further special studies were necessary for the following subjects: materials; maintenance and repair; operating procedures and control of steam generators. A separate study of sodium-water reactions was recommended considering the safety aspects related to large water leakage and economic advantage of possible detection and protection against small water leaks

  9. Steam generator inspection activities at the EPRI NDE center

    International Nuclear Information System (INIS)

    This report describes a multi-tasked project under the joint sponsorship of EPRI and EPRI Steam Generator Owners Group II. The overall project objectives include the evaluation and transfer of steam generator tube inspection NDE technology. Including the Center's first offering of the eddy current data analysis course on steam generator tubing in July-August 1985, the data analysis program has been offered six times to a total of 31 students. The students are from seven utilities, four vendors, and the NDE Center. To maximize the effectiveness of the course, a students-to-instructor ratio no greater than seven is typically maintained for each class. The program is fully compatible with the existing hardware/software being used in the industry and can be given in either analog or digital format, depending on the student's selection of eddy current analysis equipment

  10. Czechoslovak modular steam generators for LMFBR power plants

    International Nuclear Information System (INIS)

    Two modular steam generators manufactured in Czechoslovakia have been operating at the BN 350 power station in the U.S.S.R. without any complaint by the plant personnel since 1980 and 1982, respectively, supplying the actual thermal power higher by about 4 per cent than designed. The design and lay-out of the surge tanks flown trough by sodium and connected to modular steam generators are based on results obtained from the experiments performed in scale models with water and nitrogen. Whirling induction, level waviness and argon gas priming and transport into the sodium pump inlet are removed by suitable lay-out of sodium in- and outlet and the surge tank internal structure. The safety and protective system of each modular steam generator consists of one horizontal surge tank and one vertical relief tank. The connection of both tanks is carried out by two pipelines with two rupture discs

  11. Forced circulation type steam generator simulation code: HT4

    International Nuclear Information System (INIS)

    The purpose of this code is a understanding of dynamic characteristics of the steam generator, which is a component of High-temperature Heat Transfer Components Test Unit. This unit is a number 4th test section of Helium Engineering Demonstration Loop (HENDEL). Features of this report are as follows, modeling of the steam generator, a basic relationship for the continuity equation, numerical analysis techniques of a non-linear simultaneous equation and computer graphics output techniques. Forced circulation type steam generator with strait tubes and horizontal cut baffles, applied in this code, have be designed at the Over All System Design of the VHTRex. The code is for use with JAERI's digital computer FACOM M200. About 1.5 sec required for each time step reiteration, then about 40 sec cpu time required for a standard problem. (author)

  12. Proceedings: support-structure corrosion in steam generators

    International Nuclear Information System (INIS)

    Objectives of the workshop were to define limits of water chemistry impurities for various support materials exposed to secondary water for a 40-year lifetime; to review factors affecting corrosion, hideout and return of salts in steam generator crevices; and to discuss additional candidate alloys and tests that would further define and optimize support design to minimize corrosion. The subject areas included: crevice corrosion rates of alloy steels compared to carbon steel; results from examination of support plate segments removed from steam generators; and review of models for corrosion and salt hideout in crevices. The data are intended to provide a basis for utilities to determine their particular design and material needs for steam generator support structures

  13. Significance of chemical return in nuclear steam generators

    International Nuclear Information System (INIS)

    A reasonable understanding of PWR steam generator corrosion mechanisms such as denting and wastage has been developed, and adequate chemistry control programs defined to obviate the magnitude and effects of these modes of attack. However, relatively unique corrosion attack modes have been encountered at several plants notwithstanding the presence of a reasonable to very good chemistry control program when considered in light of the Steam Generator Owners Group chemistry guidelines. The uniqueness of attack also suggests that parameters not routinely measured or monitored may be playing a significant role. In the authors opinions, the only reasonable method of routinely identifying corrosion accelerating species present in crevices, sludge piles, and deposits in PWR steam generators is by performing detailed chemical return studies during power transients, shutdowns, and long term layups. Although it would be preferable to obtain samples from regions of attack, such samples generally are not available for obvious reasons

  14. Overview of steam generator tube degradation and integrity issues

    International Nuclear Information System (INIS)

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

  15. Design of jet manipulator for sludge lancing for steam generators

    International Nuclear Information System (INIS)

    The sludge accumulation in secondary side of mushroom type steam generators of Indian Pressurised Heavy Water Reactors (PHWRs) may lead to loss of thermal efficiency and corrosion. Sludge removal is required to minimise such effects for safe and enhanced operating life of the steam generators. A sludge lancing system has been developed for sludge removal from the secondary side of the steam generators. Jet Manipulator is one of the various modules of the sludge lancing system. The JM consists of three modules namely walker, elevator and nozzle heads. Each module is designed to pass through hand hole, having 180 mm diameter and 100 mm wide gap between steam generator shell and shroud. These three modules are connected to each other by quick connecting type joints and are having their specific functions. The walker crawls by step of single pitch of the tube along the central no-tube lane of the steam generator by taking lateral supports on the nearest tubes. The elevator is capable of lifting the nozzle head to a suitable height required for lancing operation of entire tube sheet of the steam generator. The nozzle head directs the multiple jets along the narrow inter tube lanes having 3 mm width, on both sides of the central no-tube lane. The nozzle can be set to move at different elevations such that the multiple jets will graze along the narrow tube lane to create the sludge lancing action. The provision exists for movement of JM in both directions, i.e. forward and reverse. This paper highlights the objective, design and development, selection of nozzles, qualification and performance evaluation of JM. The manipulator is remotely operable by compressed air in the forward and reverse direction in the central no-tube lane to position the nozzle head in the horizontal direction. (author)

  16. Health and safety impact of steam generator tube degradation

    International Nuclear Information System (INIS)

    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

  17. Chemical cleaning of steam generators: application to Nogent 1

    International Nuclear Information System (INIS)

    EDF has patented a chemical cleaning process for PWR steam generators, based on the use of a mixture or organic acids in order to dissolve iron oxides and copper with a single solution and clean dented crevices. Qualification tests have permitted to demonstrate effectiveness of the solution and its innocuousness related to steam generator materials. The process, the licence of which belongs to SOMAFER RA and Framatome has been implemented in France at Nogent. The goal was to dissolve iron oxides allowing metallic particles, aggregated on the tubesheet, to be released and mechanically removed. The effectiveness was satisfactory and this treatment is to be extended to other units. (author)

  18. 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.

  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. How safe is defect specific maintenance of steam generator tubes?

    International Nuclear Information System (INIS)

    Outside diameter stress corrosion cracking at the tube to tube support plate intersections is assessed in the paper. The impact of defect specific maintenance on steam generator operation safety and reliability was investigated. This was performed by comparing efficiencies of defect specific and traditional maintenance strategy. The efficiency was studied through expected primary-to-secondary leak rate and tube rupture probability in a case of postulated accidental operating conditions, and number of tubes which shall be plugged using both maintenance strategies. In general, the efficiency of specific maintenance is function of particular steam generator and operating cycle. (author)

  1. 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

  2. Raziskave cevi uparjalnika: Tube investigation of steam generator:

    OpenAIRE

    Gudek, K.; Korošec, Darko; Vojvodič-Tuma, Jelena

    1999-01-01

    The addy current testing of the steam generator of nuclear power plant Krško are discussed. After 121.000 hours of operation the surface of transport of heat from the primary to secondary part of the plant is diminished consequently by 17,4% and it is near the limit of the licenced operation at full power. The last investigations in 1998 confirmed that the operation time of the steam generators is running out. V prispevku so obravnavane raziskave cevi uparjalnika jedrske elektrarne Krško z...

  3. Steam generator chemical cleaning experience and prospects in Korea

    International Nuclear Information System (INIS)

    Several SGCC for the secondary side of steam generators have been successfully applied in Korea. Specific processes were developed by KEPCO research institute based on EPRI/SGOG processes. The primary purposes of SGCC in Korea are to mitigate the ODSCC of inconel-600HTMA tubes, CE-system80, and to restore the water level oscillation of inconel-600 TT tubes, Westinghouse, model F. This paper described background, field application, visual inspection, and a noted change of the application as a result of the sludge removal from the steam generators. And also the prospects of SGCC in Korea are briefly discussed. (author)

  4. The impact of maintenance strategy on steam generator reliability

    International Nuclear Information System (INIS)

    When employing a new or alternate maintenance strategy in a nuclear power plant, the question of safety and reliability is raised. Therefore, a comparative investigation of maintenance strategies is performed in this paper. A distribution of defects has been assumed to estimate the probability of having at least one tube in the steam generator failed at hypothetical accidental conditions. As a criterion of success, the tube failure probability has been employed. The methodology is illustrated by a numerical example. A typical steam generator has been analyzed during a hypothetical accident. The analysis was carried out by the means of Monte Carlo simulation and probabilistic fracture mechanics. (author)

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

    International Nuclear Information System (INIS)

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

  6. Aerosol retention in the flooded steam generator bundle during SGTR

    International Nuclear Information System (INIS)

    Research highlights: → High retention of aerosol particles in a steam generator bundle flooded with water. → Increasing particle inertia, i.e., particle size and velocity, increases retention. → Much higher retention of aerosol particles in the steam generator bundle flooded with water than in a dry bundle. → Much higher retention of aerosol particles in the steam generator bundle than in a bare pool. → Bare pool models have to be adapted to be applicable for flooded bundles. - Abstract: A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out. To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated. Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet-bundle interactions close to the tube break. Droplets were entrained from the water surface with

  7. Mitigating aging in CANDU plants

    International Nuclear Information System (INIS)

    Aging degradation is a phenomenon we all experience throughout life, both on a personal basis and in business. Many industries have been successful in postponing the inevitable impact on their related systems and components through programs to maintain long-term reliability, maintainability and safety. However, this has not always been the case for nuclear power. While all power plants are experiencing the world trend of increasing operating costs with age, few (if any) have been able to fully define the parameters that solve the aging equation, particularly in relation to major components. Inspection and preventive maintenance have not been effective in predicting life-limiting degradation and failure. In CANDU nuclear plants, utilities are taking a comprehensive approach in dealing with the aging problem. Programs have been established to identify the current condition and degradation mechanisms of critical components, the failure of which would impact negatively on station competitiveness and safety. These include subcomponents under the general headings of reactor components, civil structures, piping (nuclear and conventional), steam generators, turbines and cables. In support of these efforts, R and D projects have been defined under the CANDU Owners Group to deal with generic issues on aging common to its members (e.g., investigation of degradation mechanisms, development of tools and techniques to mitigate the effects of aging, etc.). This paper describes recent developments of this cost-shared program with specific reference to concrete aging and crack repairs, flow-assisted corrosion in piping, elastomer service life, cable aging, degradation mechanisms in steam generators and lubricant breakdown. (author)

  8. Evolution of design of steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    The first sodium cooled reactor was the experimental breeder reactor (EBR-I) in usa which was commissioned in 1951 and was incidentally the first nuclear reactor to generate electrical energy. This was followed by fast breeder reactors in USSR, UK, france, USA, japan, germany and India. The use of sodium as a coolant is due to its low moderation which helps in breeding fissile fuel from fertile materials and also its high heat transfer coefficient at comparatively low velocities. The good heat transfer properties introduce thermal stresses when there are rapid changes in the sodium temperatures. Also sodium has a chemical affinity with air and water. The steam generators for sodium cooled reactors have to allow for these novel conditions and in addition, unlike other components. Choices have to be made whether it is a recirculation type as in most fossil plants or an once through unit, the power rating, shape of the tube (straight, helical, U-tube), materials (Ferritic or austenitic), with free level of sodium or not, sodium on tube side or shell side and so on. With higher pressures and steam temperatures reheating steam after partial expansion in the turbine becomes essential as in conventional turbines. For this purpose the choice of reheating fluid viz sodium or live main steam has to be made. This paper traces the evolution of steam generator designs in the different sodium cooled reactors (chronologically) and the operation experience. 16 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

  10. Maintenance and repair of LMFBR steam generators. IV. Phenix steam generators Na H2O reaction incident

    International Nuclear Information System (INIS)

    In 1982 and 1983, four leaks occurred on the steam generators of Phenix reactor, causing each time minor sodium-water reactions. Besides the short description of the steam generators, this paper describes in detail these incidents and subsequent repair procedures. The first leak occurred after 9 years of of operation when it was no longer expected. It provided experience to the personnel and represented a trial for the procedure. Apart from the check valve the equipment behaved well. Excellent behaviour of the hydrogen detection system is emphasised, meaning sensitivity and high reliability. Its limits are mentioned. They are concerned with response time of several tens of seconds inherent in the sampling system which could be a detrimental factor in the event of a higher leakage flow rate. The advantage of the modular steam generator design in respect to easy location of the faulty module is stressed. This enables rapid action for removing the module from service

  11. New steam generation system for lead-cooled fast reactors, based on steam re-circulation through ejector

    International Nuclear Information System (INIS)

    Highlights: • Innovative steam generation system for lead-cooled fast reactors secondary loop. • Water evaporation outside of vessel heated by recirculation steam in a surface exchanger. • Steam recirculation occurs through steam jet ejector feeding bayonet heat exchangers. • Improvement of safety, availability and efficiency with respect to Loeffler system (EBBSG). - Abstract: The EBBSG (External Boiling Bayonet Steam Generator) system, proposed in previous publications, offers an alternative to the classical once-through high pressure steam generators. This system exploits the combination between the Loeffler external boiling scheme and the bayonet-tube steam generator and is expected to provide advantages in terms of safety while keeping good values of cycle performance and vessel size. The main disadvantages result in the increased size of the heat exchangers with respect to once-through steam boilers and in the need of steam blowers, as envisaged under the Loeffler scheme. In the present paper, a new and more efficient system is proposed, in which the steam circulation is assured by steam-jet ejectors instead of blowers. The innovative solution, named SJ-EBBSG (Steam-Jet External Boiling Bayonet Steam Generator), is expected to provide several advantages with respect to the original scheme. In particular, the advantages envisage an increased global efficiency (+0.49% with respect to EBBSG) due to the lower power consumption of the auxiliaries and smaller size of the bayonet heat exchangers (−6.1% diameter, −7.3% length), other than increased safety and plant availability. Throughout the article, the two steam generation solutions are compared and the advantages demonstrated by calculations

  12. Services focused on steam generation; Dienstleistungen rund um die Dampferzeugung

    Energy Technology Data Exchange (ETDEWEB)

    Rogatty, W. [Viessmann Werke GmbH, Allendorf (Germany); Schibel, T. [Viessmann Werke GmbH, Berlin (Germany)

    2008-01-15

    Cost-efficient steam generation is vital in industrial production, in foodstuffs processing, in clinics, breweries and laundries - in fact everywhere, where steam is used in large volumes. In addition, availability and operational dependability also play an important role in many applications. Problems with steam generation can disrupt operations and may thus cause high consequential costs. To meet these requirements, manufacturers such as Viessmann provide not only efficient and dependable technical solutions, but also more extensive support. The comprehensive range of services extends from highly competent advice and consulting, plus planning support, through provision of the entire system equipment from a single source, up to and including commissioning and long-term after-sales support in the form of maintenance and servicing. (orig.)

  13. Extending steam generator life at maximum electrical power output

    International Nuclear Information System (INIS)

    This paper reports the results of a study performed to support the steam generator (SG) strategic planning activities of the Northeast Utilities' Connecticut Yankee (CY) nuclear power plant. The goal of the strategic plan is to enable Cy to operate to the end of its licensed life with the current steam generators in a safe, reliable, and economical manner. one element of the strategic plan involved evaluating Tave reductions for the purpose of reducing tube corrosion rates. The evaluation showed that while reducing Tave would provide corrosion relief, it would also result in unacceptable electrical power output losses. Encotech's Steam Turbine Diagnostic Program provided the plant performance estimates required to perform this evaluation. It was used to quantify electrical power output as a function of Tave turbine control valve (TCV) configuration, SC plugging level, and core thermal power level. In addition, this work lead to the reconfiguration of the turbine control valves, resulting in increased power output

  14. Improve steam generator moisture carryover rate at Maanshan NPS by cleaning steam drum internal sludge

    International Nuclear Information System (INIS)

    2013 August Maanshan Nuclear Power Plant commissioned perform steam generator moisture carryover test (MCO) and get a high rate of both unit. The reported MCO values for the unit 2 SGs significantly higher and thus more urgent to address, as the average MCO value of 0.31% is substantially higher than the design limit and what is considered acceptable (0.25%) by most turbine vendors. With both unit MCO beyond the design limit, a plan needs to be developed to determine the cause of these high values (via an inspection of the steam drum region of the SG, and then develop actions necessary to improved the MCO rate). Westinghouse made steam drum inside inspection and report, there is no obvious regional water separator device degradation phenomena, resulting in MCO phenomenon is due to sludge accumulation in the dryer equipment bend, causing the dryer device function is reduced. Maanshan nuclear plant decided using chemical and join some manhand process to remove sludge in the dryer at NOV-2013 (unit 1EOC-21). Washing steps are as follows: 1. manually remove visible mud inside the steam drum. 2. Loose soaked with chemicals and solvents to wash portion of sludge. 3. Low pressure water flush with the bottom of the dryer to remove loose sludge Results: Step 1 manually remove visible mud (SG A: 14 kg, B: 16.5 kg, C: 19.8 kg). Step 2 Chemical dissolve and remove sludge from the steam generator (SG A: 149.8 kg, B: 213.9, C: 248.9 kg). Step 3 flush and collect insoluble sludge (SG A: 97.3 kg, B: 93 kg, C: 50.1 kg) After steam drum washing process, Maanshan NPS check the photo from micro cameras and find most of the sludge was remove from the dryer vane pocket, but we still need to perform MCO test to confirm the result. (author)

  15. A balanced strategy in managing steam generator thermal performance

    International Nuclear Information System (INIS)

    This paper presents a balanced strategy in managing thermal performance of steam generator designed to deliver rated megawatt thermal (MWt) and megawatt electric (MWe) power without loss with some amount of thermal margin. A steam generator (SG) is a boiling heat exchanger whose thermal performance may degrade because of steam pressure loss. In other words, steam pressure loss is an indicator of thermal performance degradation. Steam pressure loss is mainly a result of either 1) tube scale induced poor boiling or 2) tube plugging historically resulting from tubing corrosion, wear due to flow induced tube vibration or loose parts impact. Thermal performance degradation was historically due to tube plugging but more recently it is due to poor boiling caused by more bad than good constituents of feedwater impurities. The whole SG industry still concentrates solely on maintenance programs towards preventing causes for tube plugging and yet almost no programs on maintaining adequate boiling of fouled tubes. There can be an acceptable amount of tube scale that provides excellent boiling capacity without tubing corrosion, as operational experience has repeatedly demonstrated. Therefore, future maintenance has to come up balanced programs for allocating limited resources in both maintaining good boiling capacity and preventing tube plugging. This paper discusses also thermal performance degradation due to feedwater impurity induced blockage of tube support plate and thus subsequent water level oscillations, and how to mitigate them. This paper provides a predictive management of tube scale for maintaining adequate steam pressure and stable water level without loss in MWt/MWe or recovering from steam pressure loss or water level oscillations. This paper offers a balanced strategy in managing SG thermal performance to fulfill its mission. Such a strategy is even more important in view of the industry trend in pursuing extended power uprate as high as 20 percent

  16. CASTOR - Advanced System for VVER Steam Generator Inspection

    International Nuclear Information System (INIS)

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

  17. Modeling flow-accelerated corrosion in CANDU

    International Nuclear Information System (INIS)

    Flow-accelerated corrosion (FAC) of large areas of carbon steel in various circuits of CANDU plants generates significant quantities of corrosion products. As well, the relatively rapid corrosion rate can lead to operating difficulties with some components. Three areas in the plant are identified and a simple model of mass-transfer controlled corrosion of the carbon steel is derived and applied to these areas. The areas and the significant finding for each are given below: A number of lines in the feedwater system generate sludge by FAC, which causes steam generator fouling. Prediction of the steady-state iron concentration at the feedtrain outlet compares well with measured values. Carbon steel outlet feeders connect the reactor core with the steam generators. The feeder surface provides the dissolved iron through FAC, which fouls the primary side of the steam generator tubes, and can lead to derating of the plant and difficulty in tube inspection. Segmented carbon steel divider plates in the steam generator primary head leak at an increasing rate with time. The leakage rate is strongly dependent on the tightness of the overlapping joints. which undergo FAC at an increasing rate with time. (author) 7 refs., 5 tabs., 6 figs

  18. Fission product retention during faults involving steam generator tube rupture

    International Nuclear Information System (INIS)

    In some PWR fault conditions, such as stuck open safety relief valve in the secondary circuit or main steam line break, the release of fission products to the atmosphere may be increased by the leakage of primary coolant into the secondary circuit following steam generator tube rupture. The release may be reduced by retention either within the primary circuit or within the affected steam generator unit (SGU). The mechanisms leading to retention are reviewed and quantified where possible. The parameters on which any analysis will be most critically dependent are identified. Fission product iodine and caesium may be retained in the secondary side of a SGU either by partition to retained water or by droplet deposition on surfaces and subsequent evaporation to dryness. Two extreme simplifications are considered: SGU 'dry', i.e. the secondary side is steam filled, and SGU 'wet', i.e. the tube bundle is covered with water. Consideration is given to: the distribution of fission products between gaseous and aerosol forms; mechanisms for droplet formation, deposition and resuspension; fission product retention during droplet or film evaporation primary coolant mixing and droplet scrubbing in a wet SGU; and the performance of moisture separators and steam driers. (author)

  19. Corrosion allowances for sodium heated steam generators: evaluation of effects and extrapolation to component life time

    International Nuclear Information System (INIS)

    Steam generator tubes are subjected to two categories of corrosion; metal/sodium reactions and metal/water-steam interactions. Referring to these environmental conditions the relevant parameters are discussed. The influences of these parameters on the sodium corrosion and water/steam-reactions are evaluated. Extrapolations of corrosion values to steam generator design conditions are performed and discussed in detail. (author)

  20. The Design Concept of a Steam Generator Cassette Mock-Up for ISI of Helical Tubes in SMART Steam Generator

    International Nuclear Information System (INIS)

    The SMART reactor steam generator is composed of 8 Steam Generator Cassettes (SGC) and each the SGC has a once-through-type, helical-coil-tube bundle structure using INCONEL alloy 690 tubes. The SGC installed in reactor vessel is a kind of heat exchanger made of INCONEL alloy 690 tubes. This paper introduces the design concepts of an SGC mock-up for the test probe insertion ability of In- Service Inspection (ISI). The backgrounds of selected tube material, size and tube composition are described

  1. The Design Concept of a Steam Generator Cassette Mock-Up for ISI of Helical Tubes in SMART Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Choung, Yun Hang; Kim, Dong Ok; Park, Jin Seok; Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    The SMART reactor steam generator is composed of 8 Steam Generator Cassettes (SGC) and each the SGC has a once-through-type, helical-coil-tube bundle structure using INCONEL alloy 690 tubes. The SGC installed in reactor vessel is a kind of heat exchanger made of INCONEL alloy 690 tubes. This paper introduces the design concepts of an SGC mock-up for the test probe insertion ability of In- Service Inspection (ISI). The backgrounds of selected tube material, size and tube composition are described.

  2. Steam generators of Phenix: Measurement of the hydrogen concentration in sodium for detecting water leaks in the steam generator tubes

    International Nuclear Information System (INIS)

    The Phenix secondary circuits are provided with measurement systems of hydrogen concentration in sodium, that allow for the detection of possible water leaks in steam generators and the location of a faulty module. A measurement device consists of : a detector with nickel membranes of 0, 3 mm wall thickness, an ion pump with a 200 l/s flow rate, a quadrupole mass spectrometer and a calibrated hydrogen leak. The temperature correction is made automatically. The main tests carried out on the leak detection systems are reported. Since the first system operation (October 24, 1973), the measurements allowed us to obtain the hydrogen diffusion rates through the steam generator tube walls. (author)

  3. Steam generator of Vandellos nuclear power station: Operational experience

    International Nuclear Information System (INIS)

    The Central Nuclear de Vandellos power station at Hospitalet del Infante, Spain, is a 500 MWe gas graphite moderated natural uranium reactor. The plant has generated over 46,000 million KWh over the past thirteen years of service. Throughout this service, the plant has suffered from THO phase erosion-corrosion damage in the steam generator sections of the system. The Vandellos steam generators are once-through units constructed of 1386 mild steel tubing (panels) each fabricated into a serpentine containing 83 horizontal passes. Four independent steam generator circuits are combined to feed two, 250 MWe turbines. Erosion-corrosion damage has caused panel element leakage in the evaporation of some tubing elements. The rate of erosion-corrosion damage has been modified through different operational changes since damage was first detected in 1975. This paper describes the different operating behavior of the four steam generators and an evaluation of damage through the expertise of different technical resource groups. The changes in plant operating technique discussed include hydrodynamic conditions and chemical treatment parameters. One of the most important changes in plant operation has been in the use of amines as alkaline agents. Solutions of ammonia were initially used for pH control of feedwater. In an effort to reduce erosion-corrosion levels below rates experienced using ammonia, a change was made to the use of morpholine, and more recently, a change to the use of AMP(2 amino-2-methyl-1-propanol) has shown favorable results. The paper outlines the overall behavior of steam generator function under plant transition conditions, and contrasts that behavior with current chemical parameters experienced using AMP treatment. Water chemistry characteristics are used to present an evaluation of the development of erosion-corrosion damages from 1976 through present operating conditions. (author)

  4. In-service inspection of steam generator pipes

    International Nuclear Information System (INIS)

    Tasks achieved during in-service inspection by Eddy currents in steam generator pipes of Atucha Reactor are described. Argentinian technicians made an ''on-the-job training'' under the supervision of KWU's personnel. This inspection was realized between November and December 1987

  5. Applied research concerning the direct steam generation in parabolic troughs

    Energy Technology Data Exchange (ETDEWEB)

    Eck, M.; Eickhoff, M. [German Aerospace Center (DLR), Inst. of Technical Thermodynamics, Stuttgart (Germany); Zarza, E.; Valenzuela, L. [CIEMAT - Plataforma Solar de Almeria, Almeria (Spain); Rheinlaender, J. [Center for Solar Energy and Hydrogen Research, Baden-Wuerttemberg (ZSW), Stuttgart (Germany)

    2003-04-01

    With levelized electricity costs (LEC) of 10-12 USCts/kWh the well-known SEGS (Solar Electric Generating Systems) plants in California are presently the most successful solar technology for electricity generation [Price and Cable (2001) Proc. ASME Int. Solar Energy Conf. Forum 2001]. The SEGS plants apply a two-circuit system, consisting of the collector circuit and the Rankine cycle of the power block. These two-circuits are connected via a heat exchanger. In the case of the Direct Steam Generation (DSG) in the collector field [Zarza et al. (2001) Proc. Solar Forum 2001, Washington], the two-circuit system turns into a single-circuit system, where the collector field is directly coupled to the power block. This renders a lower investment and higher process temperatures resulting in a higher system efficiency. Due to the lower investment and the higher efficiency a reduction of the LEC of 10% is expected when the DSG process is combined with improved components of the solar collectors [Zarza (2002) DISS Phase II Final Report, EU Contract No. JOR3-CT98-0277]. Within the European DISS (Direct Solar Steam) project the feasibility of the direct steam generation has been proven in more than 3700 operation hours. Steam conditions of 100 bar and 400 deg C have been demonstrated. This paper presents the main scientific results of the DISS project that aims at the investigation and demonstration of the DSG process in parabolic troughs under real solar conditions. (Author)

  6. Denting in steam generators; Denting en los generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Mendoza Gomez, C.

    2014-07-01

    Steam generators installed in power plants of Asco and Almaraz in the 1990s presented a degradation called denting and is probably associated with the accumulation of muds and impurities from the secondary system. Why are applying strategies to maintain an adequate level of safety in plants.

  7. Status of the CRBRP steam-generator design

    International Nuclear Information System (INIS)

    Fabrication of the Prototype Unit is near completion and will be delivered to the test site in August, 1981. The Plant Unit design is presently at an advanced stage and will result in steam generator units fully capable of meeting all the requiments of the CRBRP Power Plant

  8. Corrosion and Rupture of Steam Generator Tubings in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-15

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

  9. Polyvalent remote manipulator robot for nuclear reactor steam generators

    International Nuclear Information System (INIS)

    Remote controlled manipulator for work inside the water box of a steam generator has a base plate bolted over the man hole, an orienting base that can rotate the robot about the vertical axis, three coupled articulating joints and three arms

  10. 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.

  11. Radiological protection for the ANGRA 1 steam generator replacement outage

    International Nuclear Information System (INIS)

    The Angra 1 Nuclear Power Plant (NPP) is a Westinghouse two-loop plant with net output before its 1P16 Outage of 632 MWe, with the Old Steam Generators (OSG) type model D3, which were replaced by two new Steam Generators with feed water-ring system. Localized in Angra dos Reis, Rio de Janeiro - Brazil, Angra 1 started in commercial operation in 1985 and, from the beginning problems related to corrosion have appeared in the Inconel 600 alloy of the tubes. The corrosion problems indicated the necessity for a strong control of the tubes thicknesses and, after a time, the ELETRONUCLEAR decided to replace the OSG. In 2009, ELETRONUCLEAR initiated in January 24, the actions for the Steam Generators Replacement - SGR. During the SGR process, several controls were applied in field, which made possible to have no radiological accidents, no dose limits exceeded, and permitted to achieve a very good result in terms of Collective Dose. This paper describes the radiological controls applied for the Angra 1 Steam Generator Replacement Outage, the radiological protection team sizing and distribution and the obtained results. (author)

  12. Effect of liquid waste discharges from steam generating facilities

    International Nuclear Information System (INIS)

    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

  13. Design, Construction and Testing of a Parabolic Solar Steam Generator

    OpenAIRE

    Folaranmi, Joshua

    2009-01-01

    This paper reports the design, construction and testing of a parabolic dish solar steam generator. Using concentrating collector, heat from the sun is concentrated on a black absorber located at the focus point of the reflector in which water is heated to a very high temperature to form steam. It also describes the sun tracking system unit by manual tilting of the lever at the base of the parabolic dish to capture solar energy. The whole arrangement is mounted on a hinged frame supported with...

  14. Direct steam generation using a water injection system

    International Nuclear Information System (INIS)

    One way to reduce plant price is by an increase in efficiency. About 3/4 of the energy in the Rankine cycle is used to evaporate water. By allowing the evaporation to take place directly in the collector tube, most of the tube temperature would be around the saturation temperature. Another advantage is the high amount of energy that can be store in the water-to-steam phase change. This reduces the required mass flow (tubing cost, (parasitics) in the solar field. At the moment there are two Direct Steam Generator systems which show promise. Luz developed the once-through boiler, BII developed the injection system

  15. 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.

  16. Multipurpose expert-robot system model for control, diagnosis, maintenance, and repairs at the steam generators of the NPP

    International Nuclear Information System (INIS)

    The paper presents the model concept for a multipurpose expert-robot system for control, diagnosis, forecast, maintenance, and repairs at the steam generators of CANDU type nuclear power plants. The system has two separate parts: the expert system and the robot (manipulator) system. These parts compose a hierarchic structure with the expert system on the upper level. The expert system has a blackboard architecture, to which tree interfaces with the robot system, with the control system of the NPP and with the methods and techniques of control, maintenance and repairs system of the steam generator are added. Due to complex nature of its activities the expert-robot system model combines the deterministic type reasons with probabilistic, fuzzy, and neural-networks type ones. The information that enter the expert system comes from the robot system, from process, from user, and human expert. The information that enter robot system comes from the expert system, from the human operator (when connected) and from process. Control maintenance and repair operations take place by means of the robot system that can be monitored either directly by the expert system or by the human operator who follows its activity. All these activities are performed in parallel with the adequate information of the expert system directly, by the human operator, about the status parameters and, possibly, operating parameters of the steam generator components. The expert-robot system can work independently, but it can be connected and integrated in the control system of NPP, to take over and develop some of its functions. The activities concerning diagnosis and characterization of the state of steam generator components subsequent to control, as well as the forecast of their future behavior, are performed by means of the expert system. Due to these characteristics the expert-robot system can be used successfully in personnel training activities. (Author)

  17. Luncheon address: Development of the CANDU reactor

    International Nuclear Information System (INIS)

    The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book Canada Enters the Nuclear Age. The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. 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.

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

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. Steam generators clogging diagnosis through physical and statistical modelling

    International Nuclear Information System (INIS)

    Steam generators are massive heat exchangers feeding the turbines of pressurised water nuclear power plants. Internal parts of steam generators foul up with iron oxides which gradually close some holes aimed for the passing of the fluid. This phenomenon called clogging causes safety issues and means to assess it are needed to optimise the maintenance strategy. The approach investigated in this thesis is the analysis of steam generators dynamic behaviour during power transients with a mono dimensional physical model. Two improvements to the model have been implemented. One was taking into account flows orthogonal to the modelling axis, the other was introducing a slip between phases accounting for velocity difference between liquid water and steam. These two elements increased the model's degrees of freedom and improved the adequacy of the simulation to plant data. A new calibration and validation methodology has been proposed to assess the robustness of the model. The initial inverse problem was ill posed: different clogging spatial configurations can produce identical responses. The relative importance of clogging, depending on its localisation, has been estimated by sensitivity analysis with the Sobol' method. The dimension of the model functional output had been previously reduced by principal components analysis. Finally, the input dimension has been reduced by a technique called sliced inverse regression. Based on this new framework, a new diagnosis methodology, more robust and better understood than the existing one, has been proposed. (author)

  4. Joint studies on large CANDU

    International Nuclear Information System (INIS)

    from economic, safety and strategic viewpoints. A large number of research and development programs are now in place at AECL and KAERI that will permit substantial improvements to be realized in the next generation of CANDU okabts, Furthermore, opportunities exist for engineered improvements based on the research and development in advancing the generic CANDU Technology. Final Large CANDU joint study report with technical deliverables will be issued 1994 October. Phase 2 R and D program of the joint studies will be determined this year and implemented in next year. CANDU neutron economy permits versatility in choices of fuel cycles. This allows a utility to choose fuel cycle options for lower fuelling cost, better security of supply, and ultimately for much lower spent-fuel volume, than with PWR's alone. To meet Korea's strategic requirements, CANDU should be an integral part of the electricity supply mix.

  5. Experimental evaluation of the heat transfer performance of sodium heated once through steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Vinod, V., E-mail: vvinod@igcar.gov.in; Sivakumar, L.S.; Kumar, V.A. Suresh; Noushad, I.B.; Padmakumar, G.; Rajan, K.K.

    2014-07-01

    Highlights: • PFBR has eight units of steam generators to transfer 1250 MWt power. • A model steam generator was tested for its heat transfer performance. • The model steam generator transferred 6.05 MWt power at nominal conditions. • To produce steam at nominal conditions 91.7% of area is sufficient. • The steam generator design for PFBR is validated by experiments. - Abstract: Steam generator is a crucial component in a nuclear power plant because its availability is directly linked to the availability of heat transport system and thus the plant availability. In Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction in India, eight number of steam generators each with a heat transfer capacity of 156 MWt transfers 1250 MW of heat from secondary sodium to the conventional steam/water system. The sodium heated once through steam generator with 23 m long seamless straight tubes produces super heated steam at 17.2 MPa pressure and 493 °C temperature. A model steam generator of 5.5 MWt power was tested in steam generator test facility of Indira Gandhi Center for Atomic research for validating the thermal hydraulic and mechanical design of the steam generator. The testing revealed the adequacy of heat transfer capability of the steam generator to transfer the intended power. From the experimental data it is estimated that the steam generator has 8.3% more tube surface area than the required to produce steam at nominal conditions. This paper gives the details of the model steam generator, heat transfer experiments conducted to validate the thermal design and the method for estimating the additional heat transfer area in once through type steam generator.

  6. Experimental evaluation of the heat transfer performance of sodium heated once through steam generator

    International Nuclear Information System (INIS)

    Highlights: • PFBR has eight units of steam generators to transfer 1250 MWt power. • A model steam generator was tested for its heat transfer performance. • The model steam generator transferred 6.05 MWt power at nominal conditions. • To produce steam at nominal conditions 91.7% of area is sufficient. • The steam generator design for PFBR is validated by experiments. - Abstract: Steam generator is a crucial component in a nuclear power plant because its availability is directly linked to the availability of heat transport system and thus the plant availability. In Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction in India, eight number of steam generators each with a heat transfer capacity of 156 MWt transfers 1250 MW of heat from secondary sodium to the conventional steam/water system. The sodium heated once through steam generator with 23 m long seamless straight tubes produces super heated steam at 17.2 MPa pressure and 493 °C temperature. A model steam generator of 5.5 MWt power was tested in steam generator test facility of Indira Gandhi Center for Atomic research for validating the thermal hydraulic and mechanical design of the steam generator. The testing revealed the adequacy of heat transfer capability of the steam generator to transfer the intended power. From the experimental data it is estimated that the steam generator has 8.3% more tube surface area than the required to produce steam at nominal conditions. This paper gives the details of the model steam generator, heat transfer experiments conducted to validate the thermal design and the method for estimating the additional heat transfer area in once through type steam generator

  7. Efficiency and accuracy of the perturbation response coefficient generation method for whole core comet calculations in BWR and CANDU configurations

    International Nuclear Information System (INIS)

    The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)

  8. CANDU 9 fuelling machine carriage

    International Nuclear Information System (INIS)

    Continuous, on-power refuelling is a key feature of all CANDU reactor designs and is essential to maintaining high station capacity factors. The concept of a fuelling machine carriage can be traced to the early CANDU designs, such as the Douglas Point Nuclear Generating Station. In the CANDU 9 480NU unit, the combination of a mobile carriage and a proven fuelling machine head design comprises an effective means of transporting fuel between the reactor and the fuel transfer ports. It is a suitable alternative to the fuelling machine bridge system that has been utilized in the CANDU 6 reactor units. The CANDU 9 480NU fuel handling system successfully combines features that meet the project requirements with respect to fuelling performance, functionality, seismic qualification and the use of proven components. The design incorporates improvements based on experience and applicable current technologies. (author). 4 figs

  9. Proceedings of steam turbine-generator developments for the power generation industry

    International Nuclear Information System (INIS)

    This book contains proceedings of Steam Turbine Generator developments for the power industry. Topics covered include: areas of current concentration and continuing expansion and development in the power generation industry; the combined cycle; the continuing development of more reliable and more efficient low-pressure steam turbine designs; several techniques and processes under development which may provide significant improvements in the future thermal performance of power generation systems

  10. Tests and analysis on steam generator tube failure propagation

    International Nuclear Information System (INIS)

    The understanding of leak enlargement and failure propagation behavior is essential to select a design basis leak (DBL) of LMFBR steam generators. Therefore, various series of experiments, such as self-enlargement tests, target wastage tests, failure propagation tests were conducted in a wide range of leak using test facilities of SWAT at PNC/OEC. Especially, in the large leak tests, potential of overheating failure was investigated under a prototypical steam cooling condition inside target tubes. In the small leak, the difference of wastage resistivity was clarified among several tube materials such as 9-chrome steels. In regard to an analytical approach, a computer code LEAP (Leak Enlargement and Propagation) was developed on the basis of all of these experimental results. The code was used to validate the previously selected DBL of the prototype reactor, Monju, steam generator. This approach proved to be successful in spite of somewhat over-conservatism in the analysis. Moreover, LEAP clarified the effectiveness of a rapid steam dump and an enhanced leak detection system. The code improvement toward a realistic analysis is desired, however, to lessen the DBL for a future large plant and then the re-evaluation of the experimental data such as the size of secondary failure is under way. (author). 4 refs, 8 figs, 1 tab

  11. Multifunctional Porous Graphene for High-Efficiency Steam Generation by Heat Localization.

    Science.gov (United States)

    Ito, Yoshikazu; Tanabe, Yoichi; Han, Jiuhui; Fujita, Takeshi; Tanigaki, Katsumi; Chen, Mingwei

    2015-08-01

    Multifunctional nanoporous graphene is realized as a heat generator to convert solar illumination into high-energy steam. The novel 3D nanoporous graphene demonstrates a highly energy-effective steam generation with an energy conversation of 80%. PMID:26079440

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

    Science.gov (United States)

    Bergunker, V. D.

    2011-03-01

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

  13. Gamma spectrometry application for steam generators radiological characterization

    International Nuclear Information System (INIS)

    Steam generators (SGs) are heat exchangers used to convert water into steam from heat produced in a nuclear reactor core. They are used in pressurized water reactors between the primary and secondary coolant loops. The preservation of the complete separation between the primary and secondary fluid is of capital importance in order to avoid radioactive contamination of secondary fluid and small loss of coolant also. Most French PWR suppliers use the vertical U-tube design with inverted tubes. The heat exchange section consists of a vertical, inverted U-tube bundle with the tube plate and the channel head. The steam drum portion consists of the internal moisture separating equipment and the enclosing pressure shell. In operation, primary coolant from the nuclear reactor vessel is circulated through the U-tubes. During this passage, the coolant gives off heat to the secondary water on the shell side of the steam generator, causing it to boil to steam. This steam, in turn, is passed through the moisture separating equipment in order to reduce the entrained moisture content and produce essentially dry steam. These U tubes have an important safety role because they constitute one of the barriers between the radioactive and non-radioactive sides of the power plant. For this reason, the integrity of the tubing is essential in minimizing the leakage of water between the two sides. The integrity of the primary system must be assured in any case. When power plants approach the end of life, decontamination and decommissioning of Steam Generators must be planned. It implies many issues: strategic, technological and scientific, measurement, environmental, legislation, and economic issues. The first step is the decontamination which can be performed by several techniques such as: washing, heating, chemical or electrochemical action, mechanical cleaning, or others... Prior to performing the decommissioning, the appropriate knowledge on the presence, kind and distribution of

  14. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    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

  15. 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

  16. Materials performance in CANDU reactors: The first 30 years and the prognosis for life extension and new designs

    International Nuclear Information System (INIS)

    A number of CANDU reactors have now been in-service for more than 30 years, and several are planning life extensions. This paper summarizes the major corrosion degradation operating experience of various out-of-core (i.e., excluding fuel channels and fuel) materials in-service in currently operating CANDU reactors. Also discussed are the decisions that need to be made for life extension of replaceable and non-replaceable components such as feeders and steam generators, and materials choices for new designs, such as the advanced CANDU reactor (ACR) and enhanced CANDU-6. The basis for these choices, including a brief summary of the R and D necessary to support such decisions is provided. Finally we briefly discuss the materials and R and D needs beyond the immediate future, including new concepts to improve plant operability and component reliability

  17. R and R programmes to advance CANDU technology

    International Nuclear Information System (INIS)

    A key characteristic of the CANDU reactor design is the ability to meet future requirements via incremental modifications as opposed to revolutionary design changes. The main objectives for advancing CANDU technology are 1) to reduce capital and operating costs, 2) to increase capacity factors, 3) to increase passive safety, and 4) to enhance fuel/fuel cycle flexibility. These objectives are being met by performing research and development in 6 key areas: fuel channels, fuel/fuel cycle technology, safety, heavy water production, plant systems and components, and information technology. Fuel channel improvements are gained through the elucidation and application of basic materials science for life extension. Fuel and fuel cycle work is focusing on advanced cycles, and on the development of a bundle to act as a carrier for advanced fuels that improves burnup and economics. In safety, the inherent features of CANDU are used to enhance passive or natural safety concepts, such as the use of the moderator as an effective heat sink, and the development of low temperature fuels. Heavy water processes are being developed that can be used with existing hydrogen sources (such as electrolytic hydrogen or steam reformers), or that can be used in a stand-alone mode. Plant systems and components work includes improvements to plant components such as steam generators, and the application of advanced control centre technology. Information technology is being developed to improve all aspects of CANDU design, construction, and operation. This paper gives an overview of some of the R and D in these areas that is supporting the incremental improvement of current and advanced CANDU designs. (author). 7 figs

  18. 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)

  19. Radiographic examination of pressured parts for heat recovery steam generator

    International Nuclear Information System (INIS)

    A larger Nuclear Power Generation and Non Nuclear Power Generation are shipped to the job sites in various stages of fabrication and subassembly. Welding and welding related processes are central to Power Generation component fabrication and assembly in the site. This papers presents some results of the investigation that was carried out to examine the welding results of the site construction of Heat Recovery Steam Generator Piping of Tanjung Priok Gas Fired Power Plant Extension Project (740 MW) using the Radiography Test Method based on the ASME Standard. From this investigation it could be concluded that there was no crack founded in the selected specimens of the piping. The rejectable Incomplete Penetration was found in the Hot Reheat Steam Piping HRSG1. Some Porosities and Slag Inclusion are rejected because their size and length are longer than acceptable value limits, therefore should be repaired. However some of the results are accepted and no need to be repaired. The rejected Worm Holes is found on IP Super Heater Inlet Piping of HRSG1 whereas the undercut occurred on HP Steam Drum of HRSG. (author)

  20. 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)

  1. Conceptual designs for very high-temperature CANDU reactors

    International Nuclear Information System (INIS)

    300 - 500 MWe). The steam cycle and coolant conditions are proposed to be the same as CANDU-X Mark I. The major difference between the reactors is that natural convection would be used to circulate the primary coolant around the heat transport system. This approach enhances cycle efficiency and safety, and is viable for reactors operating near the critical point of water because of the large increases in heat capacity and thermal expansion coefficient across the core. The third concept, CANDUal-X, is a dual cycle concept, with core conditions similar to the Mark 1 and NC. In this concept, coolant leaving the core is first expanded through a VHP turbine in a direct cycle. Employing a dual steam cycle avoids a high-pressure steam generator. The conditions of the core and the VHP expansion can be designed such that the exhaust from the turbine is used as the heat source for an indirect cycle; that is, the secondary side can be equivalent to that presently employed in conventional CANDU plants. An advantage of this concept over conventional direct cycle nuclear plants is that only one relatively small turbine is exposed to radioactive coolant, and it is located within containment. In summary, the reactors described above represent concepts that evolve logically from the current CANDU designs to higher efficiency, with only modest extensions of current technology. This paper presents a technical overview of the different conceptual designs, as well as a brief discussion of the enabling technologies that are common to each, which is the focus of current R and D. (author)

  2. Development of a computer program to predict structural integrity against fretting wear of steam generator tubes: PIAT (program for integrity assessment of steam generator tubes)

    International Nuclear Information System (INIS)

    Highlights: ► We develop a computer code to assess the structural integrity of steam generator tubes. ► Flow-induced vibration of whole steam generator tubes can be analyzed systematically. ► The wear map is obtained to predict the wear depth of whole steam generator tubes. ► The structural integrity of steam generator tubes can be improved significantly. -- Abstract: Flow induced vibration of steam generator tubes potentially causes excessive fretting wear at the supports such as anti-vibration bars and tube support plates. For a reliable design of tubes against the flow-induced vibration related failure, the prediction of vibration and wear of tubes should be performed through complicated steps including the thermal-hydraulic analysis, dynamic modal analysis, evaluation of fluid-elastic instability, prediction of turbulence-induced vibration and wear depth for thousands of tubes. However, entire tubes cannot be evaluated within a limited time of design engineering by the conventional analysis methodology. In this paper, we describe an efficient computer program to assess the structural integrity of steam generator tubes against the flow-induced vibration related failure in a very systematic way. The program contains all the necessary thermal-hydraulic database of typical steam generators. It has a very special function to perform modal analysis for all thousands of tubes of a steam generator much faster than the conventional method. The program also performs fluid-elastic instability analysis and calculates the vibrational response to the turbulent flow excitation, and then can predict the wear depth for all tubes of a steam generator. Finally, we can generate the wear prediction map for whole tubes so that an efficient and practical steam generator maintenance management program is feasible. The utilization of the developed computer program for the design and maintenance of steam generators can significantly increase the structural integrity of steam

  3. 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.

  4. 76 FR 74834 - Interim Staff Guidance on Aging Management Program for Steam Generators

    Science.gov (United States)

    2011-12-01

    ... using Revision 3 of the Nuclear Energy Institute's (NEI) document, NEI 97-06, ``Steam Generator Program... manage steam generator aging. The LR-ISG revises the NRC staff's aging management recommendations... licensees have adopted new steam generator technical specification requirements. Revision 3 of NEI 97-06...

  5. Experience in steam generator operation at NPPs. Main problems and ways for their solution

    International Nuclear Information System (INIS)

    The information concerning main difficulties connected with steam generator operation at foreign NPPs with PWRs is generalized. Typical causes leading to steam generator service life duration decrease are considered. The results of analysis of events, which take place during steam generator tube rupture at the Japanese Mihama NPP are given

  6. Detection and localisation of leaks in steam generators

    International Nuclear Information System (INIS)

    Steam generator experience in Netherlands is concentrated in NERATOM, the company that takes part n the construction of LMFBRs: SNR-300 in Kalkar in the framework of the international company INB (Internationale Natrium Brutreaktor Bau) together with the German company INTERATOM and the Belgian company BelgoNucleaire. Experience consists of designing constructing and testing of prototypes of the SNR-300 plant components, constructing and licensing of the components for SNR-300 proper and in conceptual design of components for larger power plants. Supporting research and testing of the prototypes is carried out by the organization for Industrial Research TNO. The two types studied are the straight tube type and the helical tube type steam generator. Descriptions of both types were published earlier. It should be pointed out here that the construction material is the ferritic 2% Cr 1 Mo 1NiNb-steel (DIN Wnr. 1.6770)

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

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. 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.

  10. Ultrasonic downcomer flow measurements for recirculating steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Janzen, Victor, E-mail: Victor.Janzen@cnl.ca [Canadian Nuclear Laboratories, Chalk River, ON, Canada K0 J 1J0 (Canada); Luloff, Brian [Canadian Nuclear Laboratories, Chalk River, ON, Canada K0 J 1J0 (Canada); Sedman, Ken [Nuclear Safety Analysis & Support Department, Bruce Power, Toronto, ON, Canada M5G 1X6 (Canada)

    2015-08-15

    Highlights: • Measuring recirculating flow in nuclear steam generators provides useful information. • Flow measurements shed light on component performance and degradation mechanisms. • Commonly used ultrasonic technology and application methods are described. • Results of measurements at several power reactors are summarized. • Potential improvements in reliability and flexibility of application are suggested. - Abstract: Measurements of downcomer flow in nuclear steam generators can provide unique fitness for service and performance indicators related to overall thermalhydraulic performance, safety related secondary-side setpoints and certain forms of degradation. This paper reviews the benefits of downcomer-flow measurements to nuclear power–plant operators, and describes methods that are commonly used. It summarizes the history and state-of-the-art of the most widely used technology, non-intrusive ultrasonic systems, including field applications at several nuclear power plants. It also describes the technical challenges that remain, and summarizes recent technical developments and future improvements.

  11. Modeling of eddy current probe response for steam generator tubes

    International Nuclear Information System (INIS)

    Sample calculations were performed with a three-dimensional (3-D) finite-element model analysis that describe the response of an eddy current (EC) probe to steam generator (SG) tubing artifacts. Such calculations could be very helpful in understanding and interpreting of EC probe response to complex tube/defect geometries associated with the inservice inspection (ISI) of steam generator (SG) tubing. The governing field equations are in terms of coupled magnetic vector and electric scalar potentials in conducting media and of total or reduced scalar potentials in nonconducting regions. To establish the validity of the model, comparisons of the theoretical and experimental responses of an absolute bobbin probe are given for two types of calibration standard defects. Preliminary results are also presented from a recent theoretical study of the effect of ligament size in axial cracks on EC indications with conventional ISI bobbin probes

  12. Failure of fretted steam generator tubes under accident conditions

    International Nuclear Information System (INIS)

    Tests were carried out with a bank of tubes in a water tunnel to determine the tolerance of flawed nuclear reactor steam generator tubes to accident conditions which would result in high cross-flow velocities. Fourteen specimen tubes were tested, each having one or two types of defect machined into the surface simulating fretting-wear type scars found in some operating steam generators. The tubes were tested at flow velocities sufficient to induce high fluid elastic-type vibrations. Seven of the tubes failed near the thinnest section of the defects during the one-hour tests, due to impacting and/or rubbing between the tube and the support. Strain gauges, displacement transducers, force gauges and an accelerometer were used on the target tube and/or the tube immediately downstream of it to measure their vibrational characteristics

  13. 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.

  14. In-service inspection of WWER steam generators

    International Nuclear Information System (INIS)

    In July 1990, INTERCONTROLE supplied Atomenergoremont with two eddy current systems (GIBBON robot) for inspecting steam generator tubes in Soviet WWER-type reactors. The robot accesses and inspects all tubes in a cylinder section of 800 mm over 90 degrees and gets round plugged tubes either by rotation or by vertical displacement. The eddy current inspection system is controlled from a mobile shelter located 200 m from the steam generator. The control system controls all subsystems on the GIBBON and axial rotating probe drives on an AT personal computer. The unique data acquisition features include the HARMONIC 210 eddy current instrument and data acquisition and verification software. Verification of axial and rotating probe data is based on AIDA, DESIREE and ESTELLE software, developed jointly by INTERCONTROLE and EdF. (Z.S.) 4 figs., 1 ref

  15. Leak detection and location in MONJU steam generators

    International Nuclear Information System (INIS)

    Leak detection system of MONJU steam generator depends mostly on in-sodium hydrogen detectors. The requirements on leak detector performance are determined from the point of view of protecting tube leak propagation due to wastage, and the process of determining the performance is shown briefly. Research and development activities on in-sodium hydrogen detectors are described and the specifications of leak detectors for MONJU are also presented. In-cover-gas hydrogen detector and acoustic detector are under development. Research and development activities on the leak location after steam generator shutdown by such methods as an electromagnetic method and ultrasonic method are described. The results of the research and development work on inserting the test probes into tubes are described also. An idea for finding the condition of tubes in the neighbourhood of the leak is also presented. (author)

  16. Radiation exposure management for the Point Beach steam generator replacement

    International Nuclear Information System (INIS)

    Replacement of the steam generators at the Point Beach Unit 1 Nuclear Plant was a turnkey project undertaken for Wisconsin Electric Power Company (WEPCO) by Westinghouse. Westinghouse had full responsibility for the program from project management and all steam generator removal/installation activities to health physics implementation. The replacement method associated with the replacement was a pipe cut approach as opposed to the channel head cut method. The replacement work began in early October 1983 and was completed on January 25, 1984 - 31 days ahead of schedule and with a record-low collective radiation exposure of 590 man-rem. The success of the operation, from a radiation exposure management point of view, is attributed primarily to extensive planning and implementation of innovative replacement techniques

  17. Homogenized behaviour of the steam generator perforated plates

    International Nuclear Information System (INIS)

    To determine the overall behaviour of structures such as multiperforated plates, that are found in the industrial components (for instance in the nuclear plant steam generators), we propose to apply the theory of the heterogeneous thermoelastic plates. First we begin by the formulation of the model, lying on an asymptotic expansion. Then we describe the application to the tube sheet and support plates case, for 900 MW and 1300 MW steam generators. Numerical values of the homogenized behaviour are provided (thermal conductivity and thermoelastic coefficients). These values are compared with those available in the literature. Some comments on the mechanical fields distribution are added, for instance: hole ovalization, stress concentrations... This study completes earlier EDF works on the thermal and mechanical homogenization of the tube sheets, which are realized before the theoretical formulation of the homogenization for plates and shells structures. (author). 16 figs., 21 tabs., 14 refs

  18. An innovative piping verification program for steam generator replacement

    International Nuclear Information System (INIS)

    The traditional programmatic approach to confirm the acceptability of piping thermal expansion has an impact on the schedule for the startup of nuclear plants. The process of obtaining, evaluating, and resolving critical measurements at pipe supports and plant structures is a critical path activity that extends the time required for the plant to obtain or resume full power operation. In order to support the schedule for and minimize the duration of the steam generator replacement (SGR) outage at North Anna Unit 1, an innovative piping verification program was developed and implemented. The approach used for the restart verification program involved a significant planning effort prior to the SGR outage and kept piping system commodity verification activities off of the critical path by performing a series of engineering evaluation tasks before and during the SGR outage. The lessons learned from the successful program is being revised and improved for implementation on the steam generator replacement project for North Anna Unit 2

  19. Life Assurance Strategy for CANDU NPP

    International Nuclear Information System (INIS)

    include design provisions to replace fuel channels and steam generators. Difficult to replace components such as reactor building structures and calandria/shield tank assembly are designed for much beyond 40 years. Given the performance of CND's to date and the successfully completed rehabilitations and the lessons learned from older plants, a newly committed CANDU will have an economic service life significantly longer than 40 years. The CANDU design life was initially set at thirty years. The key components of a CANDU nuclear steam plant are the calandria vessel, the fuel channels, the reactivity control mechanisms, and the primary heat transport components including piping and steam generators. The calandria vessel, a large stainless steel tank, experiences conditions of relatively low temperature and pressure and is designed for a very long life. Experience to date shows that of the remaining components, fuel channels and reactivity control mechanisms are replaceable. Given that other refurbishments and/or replacements can be done to existing plants, a minimum of 40 year operating life can be achieved. Large scale fuel channel replacement was dictated by Station Life Assurance rather than Life Extension considerations. This major rehabilitation program has been successfully implemented for three of the Pickering A reactors to achieve a minimum 40 year operating life. In this program steady flow of successful design and process improvements have contributed to the knowledge base and know how of the CANDU industry. Over the next few years, retuning of the fourth Pickering A unit and the first of the Bruce A units will be undertaken providing the opportunity for Life extension of these units. Steam Generators in most CANDU plants continue to perform, with relatively low tube failures and plugging rates. Remedial measures are being taken, with solutions being evaluated by Ontario Hydro to address current degradation problems due to tube fouling and sludge deposition. R

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

    International Nuclear Information System (INIS)

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

  1. Expert system for failure analysis of steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Buchmayr, B.; Cerjak, H.; Wakonig, H.; Kleemaier, R.; Nowotny, P. (Graz University of Technology, Graz (Austria))

    1990-09-01

    This paper introduces an expert system (ES) for prelocation damage analysis to steam generators. It has been developed in collaboration between Simmering-Graz-Pauker AG (SGP) and the Materials Information and Welding Technology Department of Graz Technical University. The paper indicates to power plant engineers the possibilities for the application of this kind of system and the basic features for realization of the project. 9 refs., 4 figs., 4 tabs.

  2. Thermal hydraulic analysis of Alfred bayonet tube steam generator

    OpenAIRE

    Caramello, Marco; Panella, Bruno; De Salve, Mario; Bertani, Cristina

    2015-01-01

    The paper analyzes the performance of ALFRED steam generator from the thermal-hydraulic point of view highlighting the effect of some design features. The parameters object of the study are the regenerative heat transfer, the dimension of the inner tube and the length of the bayonet. The system code RELAP5-3D/2.4.2 has been chosen for the analysis. Sensitivities analysis allowed the determination of the different design parameters influence, here briefly summarized. The increase of regenerati...

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

    Science.gov (United States)

    Hassanzadeh, Bazargan; Keshavarz, Ali; Ebrahimi, Masood

    2014-01-01

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

  4. MPGV: an expert system for steam generator maintenance

    International Nuclear Information System (INIS)

    With more than 223 SGs delivered or in the course of fabrication, FRAMATOME has acquired a wealth of experience in designing, fabricating and maintaining steam generators. this, in conjunction with the extensive FRAMATOME research and development program, makes FRAMATOME an ideal vendor of a structured program for assistance with SG operation. This service is provided by a group of FRAMATOME experts whose accumulated knowledge, competence, and experience permit a detailed analysis of all parameters affecting SG operation. (J.P.N.)

  5. Multifrequency eddy current testing of helical tubes of steam generators

    International Nuclear Information System (INIS)

    In the event of a water-sodium reaction in a steam-generator of a fast breeder reactor, it is necessary to test the tubes close to the leak to evaluate the damage. In SUPERPHENIX, the tubes are about 100m long and are coiled on a dead body. This report describes the equipment and the technic to test such tubes with multifrequency eddy current technics

  6. Failure of austenitic stainless steel tubes during steam generator operation

    Directory of Open Access Journals (Sweden)

    M. Głowacka

    2012-12-01

    Full Text Available Purpose: of this study is to analyze the causes of premature failure of steam generator coil made of austenitic stainless steel. Special attention is paid to corrosion damage processes within the welded joints.Design/methodology/approach: Examinations were conducted several segments of the coil made of seamless cold-formed pipes Ø 23x2.3 mm, of austenitic stainless steel grade X6CrNiTi18-10 according to EN 10088-1:2007. The working time of the device was 6 months. The reason for the withdrawal of the generator from the operation was leaks in the coil tube caused by corrosion damage. The metallographic investigations were performed with the use of light microscope and scanning electron microscope equipped with the EDX analysis attachment.Findings: Examinations of coil tubes indicated severe corrosion damages as pitting corrosion, stress corrosion cracking, and intergranular corrosion within base material and welded joints. Causes of corrosion was defined as wrong choice of austenitic steel grade, improper welding technology, lack of quality control of water supply and lack of surface treatment of stainless steel pipes.Research limitations/implications: It was not known the quality of water supply of steam generator and this was the reason for some problems in the identification of corrosion processes.Practical implications: Based on the obtained research results and literature studies some recommendations were formulated in order to avoid failures in the application of austenitic steels in the steam generators. These recommendations relate to the selection of materials, processing technology and working environment.Originality/value: Article clearly shows that attempts to increase the life time of evaporator tubes and steam coils by replacing non-alloy or low alloy structural steel by austenitic steel, without regard to restrictions on its use, in practice often fail.

  7. Nuclear power plant modeling and steam generator stability analysis

    International Nuclear Information System (INIS)

    This thesis describes the development of a computer model simulating the transient behavior of a pressurized water reactor (PWR) nuclear steam supply system (NSSS) and a stability analysis of steam generators in an overall NSSS structure. In the analysis of stream generator stability characteristics, an emphasis was placed on the physical interpretation of density wave oscillation (DWO) phenomena in boiling channels. The PWR NSSS code TRANSG-P is based on the nonlinear steam generator code TRANSG, in which the basic flow channel and heat-exchanger models were previously formulated. In addition to the steam generator, the TRANSG-P code includes models for the pressurizer, the pump, and the turbine. The mathematical model for fluid channels is based upon one-dimensional, nonlinear, single-fluid conservation equations of mass, momentum, and energy. Space and time discretization of these equations is accomplished using an implicit finite-difference formulation. The pressurizer is modeled as a nonequilibrium system at uniform pressure, consisting of vapor and liquid regions. Flashing and condensation are accounted for, and control elements are also modeled. The pump behavior is determined by making use of homologous curves, whereas simple energy conservation and choked flow equations are used to model the turbine. Efforts were made to assess the accuracy of the entire plant model of the TRANSG-P code through simulation of a loss-of-feedwater accident that occurred at a PWR plant. The TRANSG-P results are in reasonable agreement with the plant data, which inherently are subject to considerable uncertainties. In addition, once-through and natural-circulation boiling channel calculations, performed for the investigation of flow stability characteristics, showed good agreement with the test data

  8. 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.

  9. Performance demonstration requirements for eddy current steam generator tube inspection

    International Nuclear Information System (INIS)

    This paper describes the methodology used for developing performance demonstration tests for steam generator tube eddy current (ET) inspection systems. The methodology is based on statistical design principles. Implementation of a performance demonstration test based on these design principles will help to ensure that field inspection systems have a high probability of detecting and correctly sizing tube degradation. The technical basis for the ET system performance thresholds is presented. Probability of detection and flaw sizing tests are described

  10. Edf's experimental program on a retired steam generator

    International Nuclear Information System (INIS)

    In 1990 EDF replaced the steam generators of the DAMPIERRE unit 1 station (900 MWe, 3-loop PWR). One S.G. has been stored in a specially designed facility; component behaviours are evaluated, and the S.G. (in upright position) has been used as a full-size mockup for testing inspection and maintenance tools or processes. On-site investigations are confirmed by hot cell examinations. The program includes specific examinations that cannot be conducted during regular inspections. Particularly, investigations concerning the steam generator shells have been more extensive than during plant outages, including the upper internals support and the circular/conical section weld joint. The Inconel 182' weld between the tube sheet and partition plate has also been inspected. NDT measurements have provided a basis for the localization of the samples that have been taken to hot cells for destructive examinations. The results of this program should help the interpretation of on-site inspection results, and extend the choice of available tools/processes for inspection and maintenance. Particularly, the secondary side cleaning has been studied with sampling of the tube sheet and tube support plate before and after the operation, regarding the oxides and deposits between the tube and the tube sheet or tube support plate to determine the efficiency. S.G. components (welds, support structures, tubes,...) have been inspected to determine if any damage occurred. Investigations in the behaviour of the steam generator components, on the basis of site or hot cell examinations as well as experimental testing (corrosion, mechanical, hydraulic tests) should be helpful in predicting the performance of the steam generators, optimizing their maintenance policy, and ultimately extended their service life. (Author)

  11. CANDU combined cycles featuring gas-turbine engines

    International Nuclear Information System (INIS)

    In the present study, a power-plant analysis is conducted to evaluate the thermodynamic merit of various CANDU combined cycles in which continuously operating gas-turbine engines are employed as a source of class IV power restoration. It is proposed to utilize gas turbines in future CANDU power plants, for sites (such as Indonesia) where natural gas or other combustible fuels are abundant. The primary objective is to eliminate the standby diesel-generators (which serve as a backup supply of class III power) since they are nonproductive and expensive. In the proposed concept, the gas turbines would: (1) normally operate on a continuous basis and (2) serve as a reliable backup supply of class IV power (the Gentilly-2 nuclear power plant uses standby gas turbines for this purpose). The backup class IV power enables the plant to operate in poison-prevent mode until normal class IV power is restored. This feature is particularly beneficial to countries with relatively small and less stable grids. Thermodynamically, the advantage of the proposed concept is twofold. Firstly, the operation of the gas-turbine engines would directly increase the net (electrical) power output and the overall thermal efficiency of a CANDU power plant. Secondly, the hot exhaust gases from the gas turbines could be employed to heat water in the CANDU Balance Of Plant (BOP) and therefore improve the thermodynamic performance of the BOP. This may be accomplished via several different combined-cycle configurations, with no impact on the current CANDU Nuclear Steam Supply System (NSSS) full-power operating conditions when each gas turbine is at maximum power. For instance, the hot exhaust gases may be employed for feedwater preheating and steam reheating and/or superheating; heat exchange could be accomplished in a heat recovery steam generator, as in conventional gas-turbine combined-cycle plants. The commercially available GateCycle power plant analysis program was applied to conduct a

  12. 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.

  13. Steam generator primary manway modifications for Wolsong unit 1

    International Nuclear Information System (INIS)

    'Full text:' The steam generator primary head manway covers and manipulators in Wolsong Unit 1 had features that were problematic when opening and closing the manway covers. Some of the difficulties experienced include manipulation of the covers, ensuring that the inner manway cover gasket has seated properly and ensuring that the stud seals were leak tight. A combination of deficiencies were increasing the risk of future primary head manway cover leakage. Since this risk had the potential to cause an unscheduled shutdown, the Wolsong 1 steam generators were modified with new components that incorporated the following changes: Wider spiral wound gaskets for both inside and outside cover. New Manway Covers utilizing fully captured gasket design. Primary head machining to provide the wider seat required by the new gasket designs. New studs to accommodate the change in required length. New nut design using spiral wound gasket for sealing the nuts/studs. New cover handling device (Manway manipulators), with sufficient control to improve manway cover positioning; ease of operation and accuracy. As a Steam Generator designer and active manufacturer B and W introduced proven technology based on many years of operating experience. B and W provided the engineering analysis, replacement components and installation services. (author)

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

    International Nuclear Information System (INIS)

    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

  15. Thermal Sizing of Printed Circuit Steam Generator for Integral Reactor

    International Nuclear Information System (INIS)

    SMART aims at achieving enhanced safety and improved economics; the enhancement of safety and reliability is realized by incorporating inherent safety-improving features and reliable passive safety systems. The improvement in the economics is achieved through a system simplification, component modularization, reduction of construction time, and high plant availability. The standard design approval assures the safety of the SMART system. The capital cost of the major plant equipment has a significant effect on the overall economics of the nuclear plant. Minimizing the cost of manufacturing of the nuclear plant components is important to reduce the cost of the reactor. It is necessary to reduce the size of the steam generator in order to design a smaller reactor vessel, which is substantial for the overall construction cost, with the required thermal capacity preserved. The Printed Circuit Heat Exchanger is a type of compact heat exchangers that provides high power density along with a low pressure drop and reduced maintenance requirements. This paper describes the approach we used while determining the size of the Printed Circuit Steam Generator (PCSG) and resultant smaller reactor vessel. Thermal hydraulic and geometric parameters for the PCSG were studied. The results show that the overall volume of the steam generator can be significantly reduced. On the basis of this calculation, we can design a smaller reactor vessel with the PCSG

  16. 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.

  17. Corrosion risk of horizontal steam generators of VVER-440 units

    International Nuclear Information System (INIS)

    In the stress corrosion cracks the material inspections always indicated the presence of iron corrosion product deposits. 70-90 % of the deposits found in the vicinity of the cracks is magnetite (Fe3O4), 10 % is hematite (H-Fe2O3) and 10 % is goethite (H-FeO(OH)). The change of the water chemistry regime was successful, because the water chemical stress corrosion risk of the steam generators significantly decreased, but it has not become minimal. The design philosophy of the water chemistry regime of the secondary system for the extended operational lifetime can be summarized in the following way: the water chemistry regime must ensure the minimal stress corrosion risk of the heat-transfer tubes of the steam generators, which is acceptable according to our current knowledge in order to have as lower number of plugged steam generator tubes as possible by the end of the extended operational lifetime (the year 50). During the extended operational lifetime the reactor is operated at power of 108 %, i.e. its nominal thermal power is increased from the current 1375 MW to 1500 MW, but the increased power does not require any changes in the water chemistry regime of the secondary circuit. (O.M.)

  18. Acoustic detection of steam-water in a model of steam generator with a helicoidal tube bundle

    International Nuclear Information System (INIS)

    The study of mechanical vibrations of the wall of a simulated steam generator allows the detection of steam-water injection in sodium. Measurements carried out in this test showed that it is possible to reveal this injection and secondary leaks created by wastage

  19. 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.

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

    International Nuclear Information System (INIS)

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

  1. Steam-Generator Dilute-Chemical-Cleaning Program: steam-generator chemical-cleaning project. Annual report for 1981

    International Nuclear Information System (INIS)

    The dilute chemical cleaning program evaluates the feasibility of using low-concentration, regenerable solvents to maintain the secondary side of PWR steam generators in a clean condition. The experimental work carried out during this report period identified an acceptable dilute cleaning solvent formulated with 0.1 wt % each of citric acid, gluconic acid and ascorbic acid. Corrosion rates for the major steam generator construction materials can be limited to + or NH4+ form. Solvent pH in the range of 3.4 to 3.8 was maintained during the cleaning operations with chemical additions. It was also demonstrated that mixed-bed resins in the H-OH form are capable of removing residual chemicals after cleaning and restoring coolant quality to a conductivity level of less than 10 μmhos

  2. Loss-of-feedwater, steam generator tube rupture, and steam line break experiments: Steam generator transient response test program: Interim report

    International Nuclear Information System (INIS)

    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. Two LOF tests were analyzed in detail. Both tests were initiated from 100% power condition by shutting off the main feedwater flow. In LOF Test No. 1, the remaining boundary conditions were kept constant while in LOF Test No. 2, the power was rapidly reduced to 3%. The results show that the primary to secondary heat transfer becomes degraded when the collapsed water liquid level in the bundle region falls below approximately 50 inches. The SGTR test analyzed in detail - SGTR Test No. 2 - simulated the post-reactor-trip portion of the SGTR transient (T/sub prim/ = 5600F). The transient was initiated by starting the SGTR flow injection and simultaneously shutting off the auxiliary feedwater. The water level rose and flooded the dryer to its mid-elevation by the end of the test. The primary carry-over was shown to be less than 0.4% of the tracer mass injected into the secondary side by the SGTR flow. SGTR Test No. 3 investigated the response of the intact steam generator. Reverse heat transfer and low heat flow conditions were simulated. The results have demonstrated the occurrence of temperature stratification in the secondary water which lasted for about 800 seconds

  3. Experience with modular steam generator production and application of new testing methods

    International Nuclear Information System (INIS)

    Experience is reviewed gained at the Trebic IBZKG plant with the production of modular steam generators. The plant started producing steam generators for the Jaslovske Bohunice nuclear power plant in 1965. In addition to the steam generator for the A-1, the plant also produced a loop for the Melekess power plant and a steam generator for the BOR-60 reactor. Operating experience gained so far allowed improving the quality of the BOR steam generator, especially in the tube-tube plate joint. A double tube plate was used and the welded joint shape was changed. As a result of high requirements on the quality of welded joints, the steam generator has successfully been in operation for more then 10,000 hours. The existing experience was utilized in designing a new steam generator named Nadya. Many design and technological requirements were presented concerning the Nadya generator and many new checking operations have been included in technology. (Kr)

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. Containment internal concrete modifications during steam generator replacement at North Anna Unit 1

    International Nuclear Information System (INIS)

    North Anna Unit 1's three Westinghouse Model 51 steam generators had experienced corrosion-related degradation that required periodic inspection and plugging of steam generator tubes to ensure their continued safe and reliable operation. Despite improvements in secondary water chemistry, tube degradation had continued in the steam generators which resulted in extensive tube inspections and significant dose to personnel. It was therefore decided to replace the bottom part of the steam generator along with the tubes in January 1993. This paper presents the various containment internal concrete modifications that were done to facilitate the movement of the old and new steam generator lower assemblies out of and into containment

  7. Steam Generator Group Project: Task 10, Secondary side examination

    International Nuclear Information System (INIS)

    This report concludes an effort to examine and assess from the secondary side, the condition of the retired-from-service Surry 2A steam generator. It is includes photographs of degradation of various components or regions in a generic recirculating type steam generator. The photographic detail given in the text (and the previous report NUREG/CR-3843, PNL-5033) have not been readily available to many investigators outside the Nuclear Steam Supply Vendors and Users. The photographs include views of Inconel 600 heat exchanger tubes (0.875 diameter [nominal] x 0.050 inch wall) showing deformed and intergranularly stress-corrosion cracked U-bends, tube denting in the support plate, intergranular attack and thinning (both in the tube sheet region), support plat deformation and cracking at flow slots and in ligaments between flow holes and tube holes. In addition, photographs of tube pitting, anti-vibration bar fretting, and the sludge pile are presented. An experimental stress analysis was conducted on a distorted Row 1 tube in the region of a compressed flow slot between the 6th and 7th (top) support plates. During removal of the tube, relaxation strains were measured and residual stresses calculated. Finally a cursory metallurgical failure analysis was conducted on a broken U-bend (R1C91) to determine its mode of failure. A rotabroach boring technique was used to make multiple penetrations with minimal damage to the secondary side, at various locations on the shell

  8. Inverted Steam Generators for Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Inverted steam generators (ISG) are tube bundle heat exchangers with sodium flow on the inside the tubes and water and steam flow on the shell side. Two types of inverted steam generators, one so called micro modular (MMISG) and one modular (MISG), were developed by Czech institutes in cooperation with Russian organizations and designed and manufactured by Czech organizations in the past. Both still operate successfully at BOR 60 in RIAR Dimitrovgrad, Russian Federation. The MMISG has operated since 1981 (thus 31 year of operation) and the MISG has operated since 1991 (thus 21 year of operation). Design studies of ISG modules with 100 MW thermal power have been performed in the framework the CP European Sodium Fast Reactor (CP ESFR) project (7th EUFP). In the paper selected design parameters of Czech provenience ISGs that are in operation at the BOR 60 as well as design and safety and reliability features if the ISG 100 MW modules for CP ESFR are described and discussed. (author)

  9. Probabilistic methodology for assessing steam generator tube inspection - Phase II: CANTIA - a probabilistic method for assessing steam generator tube inspections

    International Nuclear Information System (INIS)

    The objectives of this project were to develop a computer-based method for probabilistic assessment of inspection strategies for steam generator tubes, and to document the source code and to provide a user's manual for it. The program CANTIA was created to fulfill this objective, and the documentation and verification of the code is provided in this volume. The user's manual for CANTIA is provided as a separate report. CANTIA uses Monte Carlo techniques to determine approximate probabilities of steam generator tube failures under accident conditions and primary-to-secondary leak rates under normal and accident conditions at future points in time. The program also determines approximate future flaw distributions and non-destructive examination results from the input data. The probabilities of failure and leak rates and the future flaw distributions can be influenced by performing inspections of the steam generator tubes at some future points in time, and removing defective tubes from the population. The effect of different inspection and maintenance strategies can therefore be determined as a direct effect on the probability of tube failure and primary-to-secondary leak rate

  10. Design, Construction and Testing of a Parabolic Solar Steam Generator

    Directory of Open Access Journals (Sweden)

    Joshua FOLARANMI

    2009-07-01

    Full Text Available This paper reports the design, construction and testing of a parabolic dish solar steam generator. Using concentrating collector, heat from the sun is concentrated on a black absorber located at the focus point of the reflector in which water is heated to a very high temperature to form steam. It also describes the sun tracking system unit by manual tilting of the lever at the base of the parabolic dish to capture solar energy. The whole arrangement is mounted on a hinged frame supported with a slotted lever for tilting the parabolic dish reflector to different angles so that the sun is always directed to the collector at different period of the day. On the average sunny and cloud free days, the test results gave high temperature above 200°C.

  11. The market for steam turbine generators around the world

    International Nuclear Information System (INIS)

    As a discrete market (in the mathematical meaning of the word) with irregular sales from one year to the next, the market for steam turbine generators in nuclear plants requires working out a strategy adapted to each project. The diversity of the reactors proposed (technology, thermal power, the thermodynamic characteristics of the steam supplied), the variety of the cold sources to be used (ranging from the Baltic Sea to the Indian Ocean) and the different frequencies of electricity grids (50 or 60 Hz) necessitate developing platforms of solutions. Furthermore, the requirement that local businesses have a share in contracts often entails partnerships. After pointing out the diversity of this market, the effort is made to point out its principal characteristics. (authors)

  12. Internal oxidation as a mechanism for steam generator tube degradation

    International Nuclear Information System (INIS)

    Internal oxidation has been proposed as a plausible mechanism for intergranular stress corrosion cracking (IGSCC) of alloy 600 steam generator tubing. This theory can reconcile the main thermodynamic and kinetic characteristics of the observed cracking in hydrogenated primary water. Although secondary side IG attack or IGSCC is commonly attributed to the presence of strong caustic or acidic solutions, more recent evidence suggests that this degradation takes place in a near-neutral environment, possibly dry polluted steam. As a result, internal oxidation is also a feasible mechanism for secondary side degradation. The present paper reviews experimental work carried out in an attempt to determine the validity of this mechanism. The consequences for the expected behaviour of alloys 690 and 800 replacement materials are also described. (author)

  13. Fabrication and inspection development for CRBRP steam generators

    International Nuclear Information System (INIS)

    One of the critical nonnuclear elements of the CRBRP is the steam generator that transfers the heat from the sodium system to the high-pressure steam system but must maintain integrity and separation of the two fluids. The construction material is 21/4 Cr--1 Mo alloy steel with high-purity (e.g. vacuum arc remelt) material being used for the tubing and tubesheets. For confidence in successful manufacturing of the several evaporator and superheater modules, key development activities are under way (1) for procurement of high-quality components, (2) to assure proper assembly (with emphasis on welding), and (3) to assure that adequate nondestructive testing methods are available to examine the units. (auth)

  14. Alternative method for steam generation for thermal oxidation of silicon

    Science.gov (United States)

    Spiegelman, Jeffrey J.

    2010-02-01

    Thermal oxidation of silicon is an important process step in MEMS device fabrication. Thicker oxide layers are often used as structural components and can take days or weeks to grow, causing high gas costs, maintenance issues, and a process bottleneck. Pyrolytic steam, which is generated from hydrogen and oxygen combustion, was the default process, but has serious drawbacks: cost, safety, particles, permitting, reduced growth rate, rapid hydrogen consumption, component breakdown and limited steam flow rates. Results from data collected over a 24 month period by a MEMS manufacturer supports replacement of pyrolytic torches with RASIRC Steamer technology to reduce process cycle time and enable expansion previously limited by local hydrogen permitting. Data was gathered to determine whether Steamers can meet or exceed pyrolytic torch performance. The RASIRC Steamer uses de-ionized water as its steam source, eliminating dependence on hydrogen and oxygen. A non-porous hydrophilic membrane selectively allows water vapor to pass. All other molecules are greatly restricted, so contaminants in water such as dissolved gases, ions, total organic compounds (TOC), particles, and metals can be removed in the steam phase. The MEMS manufacturer improved growth rate by 7% over the growth range from 1μm to 3.5μm. Over a four month period, wafer uniformity, refractive index, wafer stress, and etch rate were tracked with no significant difference found. The elimination of hydrogen generated a four-month return on investment (ROI). Mean time between failure (MTBF) was increased from 3 weeks to 32 weeks based on three Steamers operating over eight months.

  15. Study of Constant Voltage Control on Small Steam Generator Based on PID Algorithm

    Directory of Open Access Journals (Sweden)

    Yanjun Xiao

    2014-03-01

    Full Text Available The object of this study is a kind of 3 kW small steam generator, which can recover waste heat through making use of 0.1~0.3 MPa steam. This can exploit secondary energy efficiently. The electricity generated can be commonly used as factory lighting, heating, fan and emergency power supply. But the generation voltage of the existed steam turbine is instable, especially when the steam pressure and the load of the generator changes suddenly. This can pose a threat to electrical safety and greatly limit the market of small steam generator. In this study, PID control algorithm is used to control the amount of steam into the turbine of generator system. And the closed-loop control system can make a real-time feedback regulation to the steam, so that the generator voltage can be stable. The user's electrical safety requirements are satisfied as well.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

  17. Thermal design analysis of triple-pressure heat recovery steam generator and stream turbine systems

    International Nuclear Information System (INIS)

    A computation routine, capable of performing thermal design analysis of the triple-pressure bottoming System (heat recovery steam generator and steam turbine) of combined cycle power plants, is developed. It is based on thermal analysis of the heat recovery steam generator and estimation of its size and steam turbine power. It can be applied to various parametric analyses including optimized design calculation. This paper presents analysis results for the effects on the design performance of heat exchanger arrangements at intermediate and high temperature parts as well as steam pressures. Also examined is the effect of steam sources for de-aeration on design performance

  18. Steam generator and preheater tube ID fouling and the impact on reactor inlet header temperature and eddy current inspections

    International Nuclear Information System (INIS)

    Materials selection is an important consideration in new build and refurbishment of Heavy Water Reactors (HWR). This paper will focus on the impacts of the deposit of magnetite on the tube ID of steam generators and preheater. Bruce Power OPEX is being shared to illustrate the importance of materials selection. The deposit of magnetite on the tube ID of steam generators (SG) and preheater (PH) has two significant impacts that will be presented. Firstly, the degradation in SG and PH thermal performance causes a rise in the reactor inlet header temperature (RIHT). This rising trend continues unabated as long as deposits on the tube ID continues. If not managed this may result in loss of production due to the RIHT limits being reached. Mitigating actions such as tube ID cleaning is only a temporary solution as it does not stop the root cause which is feeder flow accelerated corrosion (FAC). Secondly, deposit of magnetite on the tube ID of steam generators (SG) and preheater (PH) has an impact on tube inspections as required by CSA N285.4. There are two impacts on SG and PH inspections. ID deposits reduces the clearance for eddy current probes in the tubes and make it more difficult to acquire inspection data. Additionally, tube ID deposits can reduce the effectiveness to detect and size flaws in the SG and PH tubes. Both issues make eddy current inspection a challenge for the utilities. These impacts affect the operation and inspection and maintenance of CANDU nuclear power plants at Bruce Power. Where possible these issues should be addressed in any future new build or refurbishment of HWR power plants. (author)

  19. Thermoelectric generation coupling methanol steam reforming characteristic in microreactor

    International Nuclear Information System (INIS)

    Thermoelectric (TE) generator converts heat to electric energy by thermoelectric material. However, heat removal on the cold side of the generator represents a serious challenge. To address this problem and for improved energy conversion, a thermoelectric generation process coupled with methanol steam reforming (SR) for hydrogen production is designed and analyzed in this paper. Experimental study on the cold spot character in a micro-reactor with monolayer catalyst bed is first carried out to understand the endothermic nature of the reforming as the thermoelectric cold side. A novel methanol steam reforming micro-reactor heated by waste heat or methanol catalytic combustion for hydrogen production coupled with a thermoelectric generation module is then simulated. Results show that the cold spot effect exists in the catalyst bed under all conditions, and the associated temperature difference first increases and then decreases with the inlet temperature. In the micro-reactor, the temperature difference between the reforming and heating channel outlets decreases rapidly with an increase in thermoelectric material's conductivity coefficient. However, methanol conversion at the reforming outlet is mainly affected by the reactor inlet temperature; while at the combustion outlet, it is mainly affected by the reactor inlet velocity. Due to the strong endothermic effect of the methanol steam reforming, heat supply of both kinds cannot balance the heat needed at reactor local areas, resulting in the cold spot at the reactor inlet. When the temperature difference between the thermoelectric module's hot and cold sides is 22 K, the generator can achieve an output voltage of 55 mV. The corresponding molar fraction of hydrogen can reach about 62.6%, which corresponds to methanol conversion rate of 72.6%. - Highlights: • Cold spot character of methanol steam reforming was studied through experiment. • Thermoelectric generation Coupling MSR process has been

  20. 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.

  1. Control concepts for direct steam generation in parabolic troughs

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-02-01

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

  2. Thermal stratification damage to the steam generator water supply pipes

    International Nuclear Information System (INIS)

    International experience has shown that there is a risk of a loss of the steam generator's normal feedwater (FFCS) and auxiliary feedwater (AFS) circuits due to thermal fatigue during hot standby or shutdown. Cases of large cracks, which could have led to leaks in a cycle, in the cases of certain generating units, have been reported in the worldwide installed base of PWR power stations. EDF was concerned by this risk and seized on the opportunity presented by the programmed replacement of the steam generators to carry out expert examinations of the common FFCS/AFS section. The examinations of 14 dismantled sections have revealed no faults greater than 1 mm in depth. The cracks are only located on the weld connecting the nozzle to the piping. They are mainly the results of welding faults. In order to detect the faults, and monitor any developments in them, EDF has developed an automatic control method using ultrasonics based on the principle of the measurement of time of flight. This technique was refined on a representative sample and can now determine the dimensions of cracks more than 3 mm deep. The crack stability analyses, carried out under faulted conditions which take the occurrence of an earthquake into account, show that a fault affecting half the thickness (10 mm) over a length of 30 mm is stable. No detectable fault was found on the PWR generating unit which has experienced the greatest number of hours of trot shutdown, which was checked in 1997. (authors)

  3. Device indicating start of steam or water reaction with sodium and damage of steam generator heat exchange tube wall

    International Nuclear Information System (INIS)

    Eddy currents induced by the alternating current of an exciting coil in the vicinity of steam or water leakage are used for indication. The coil is supplied from a power amplifier whose input is connected to an exciting generator by two measuring coils connected across each other. Their voltage is applied to a differential amplifier with an indicator. The equipment may be used for steam generators of nuclear power plants with sodium cooled reactors. (E.F.)

  4. Cost comparison of 4x500 MW coal-fuelled and 4x850 MW CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    The lifetime costs for a 4x850 MW CANDU generating station are compared to those for 4x500 MW bituminous coal-fuelled generating stations. Two types of coal-fuelled stations are considered; one burning U.S. coal which includes flue gas desulfurization and one burning Western Canadian coal. Current estimates for the capital costs, operation and maintenance costs, fuel costs, decommissioning costs and irradiated fuel management costs are shown. The results show: (1) The accumulated discounted costs of nuclear generation, although initially higher, are lower than coal-fuelled generation after two or three years. (2) Fuel costs provide the major contribution to the total lifetime costs for coal-fuelled stations whereas capital costs are the major item for the nuclear station. (3) The break even lifetime capacity factor between nuclear and U.S. coal-fuelled generation is projected to be 5%; that for nuclear and Canadian coal-fuelled generation is projected to be 9%. (4) Large variations in the costs are required before the cost advantage of nuclear generation is lost. (5) Comparison with previous results shows that the nuclear alternative has a greater cost advantage in the current assessment. (6) The total unit energy cost remains approximately constant throughout the station life for nuclear generation while that for coal-fuelled generation increases significantly due to escalating fuel costs. The 1978 and 1979 actual total unit energy cost to the consumer for several Ontario Hydro stations are detailed, and projected total unit energy costs for several Ontario Hydro stations are shown in terms of escalated dollars and in 1980 constant dollars

  5. Sound propagation in the steam generator - A theoretical approach

    International Nuclear Information System (INIS)

    In order to assess the suitability of acoustic tomography in the steam generator, detailed information on its acoustic transmission properties is needed. We have developed a model which allows one to calculate the sound field produced by an incident wave in the steam generator. In our model we consider the steam generator as a medium consisting of a two-dimensional array of infinitely long cylindrical tubes. They are thin-walled, made of metal and are immersed in a liquid. Inside them there is a liquid or a gas. The incident wave is plane and perpendicular to the cylindrical tubes. When a sound wave crosses the tube bundle, each individual tube is exposed to a fluctuating pressure field and scatters sound which, together with the incident wave, influences the pressure at all surrounding tubes. The motion of an individual tube is given by differential equations (Heckl 1989) and the pressure difference between inside and outside. The interaction of a tube wall with the fluid inside and outside is treated by imposing suitable boundary conditions. Since the cylinder array is periodic, it can be considered as consisting of a large number of tube rows with a constant distance between adjacent cylinders within a row and constant spacing of the rows. The sound propagates from row to row, each time getting partly transmitted and partly reflected. A single row is similar to a diffraction grating known from optics. The transmission properties of one row or grating depend on the ratio between spacing and wavelength. If the wavelength is larger than the spacing, then the wave is transmitted only in the original direction. However, for wavelengths smaller than the spacing, the transmitted wave has components travelling in several discrete directions. The response of one row to sound scattered from a neighbouring row is calculated from Kirchhoff's theorem. An iteration scheme has been developed to take the reflection and transmission at several rows into account. 7 refs, figs and

  6. Removal of deposited copper from nuclear steam generators

    International Nuclear Information System (INIS)

    A review of the copper-removal process implemented during the cleaning of the NPD nuclear steam generator in Ontario revealed that major shortcomings in the process were depletion of the strong ammonia solution and relatively poor copper removal. Tests have shown that the concentration of the ammonia solution can be preserved close to its initial value, and high concentrations of complexed copper obtained, by sparging the ammonia solution with oxygen recirculating through a gas recirculation loop. Using recirculating oxygen for sparging at ambient air temperature, approximately 11 g/l of copper were dissolved by 100 g/l ammonia solution while the gaseous ammonia content of the recirculating gas remained well below the lower flammability limit. The corrosion rates of mild steel and commonly used nuclear steam generator tube materials in oxygenated ammonia solution were less than 30 mil/yr and no intergranular attack of samples was observed during tests. A second technique studied for the removal of copper is to ammoniate the spent iron-removal solvent to approximately pH 9.5 and sparge with recirculating oxygen. Complexed ferric iron in the spent iron-removal solvent was found to be the major oxidizing agent for metallic copper. The ferric iron can be derived from oxidation of dissolved ferrous iron to the ferric state or from dissolved oxides of iron directly. To extract copper from the secondary sides of nuclear steam generators, strong ammonia solution sparged with recirculating oxygen is recommended as the first stage, while ammoniated spent iron-removal solvent sparged with recirculating oxygen may be used to remove the copper freshly exposed during the removal of iron

  7. 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.

  8. Influence of sodium water reaction on MONJU steam generator

    International Nuclear Information System (INIS)

    Despite the strenuous efforts improving the reliability of steam generators, it is required to ascertain the safe shutdown at Design Basis Leak and also to take the necessary actions to minimize the plant damage for more realistic small leaks. The process of Monju DBL selection and its supporting R and D works are included in this paper, together with the evaluation of system and critical components in direct connection with DBL. The detail plant shutdown procedures (including auxiliary system sequential action) at the time of water leaks are also explained. (author)

  9. Criteria for maintenance and repair - LMFBR steam generators

    International Nuclear Information System (INIS)

    The maintenance and repair criteria will be reviewed with respect to the designs presently under construction for the SNR-300 plant. This criteria shall be based upon the philosophy that safety and reliability are of the highest importance at all operating modes, while availability shall be maximized. To maximize the safety of the steam generator, measures have been taken to reduce the possibilities of failure by simplicity in design, choice of material, methods of fabrication and high quality assurance of critical parts of the pressure boundaries. The maintenance and repair program shall meet the same criteria or the intent of these criteria as applied for the original product. (author)

  10. Failure of austenitic stainless steel tubes during steam generator operation

    OpenAIRE

    M. Głowacka; J. Łabanowski; S. Topolska

    2012-01-01

    Purpose: of this study is to analyze the causes of premature failure of steam generator coil made of austenitic stainless steel. Special attention is paid to corrosion damage processes within the welded joints.Design/methodology/approach: Examinations were conducted several segments of the coil made of seamless cold-formed pipes Ø 23x2.3 mm, of austenitic stainless steel grade X6CrNiTi18-10 according to EN 10088-1:2007. The working time of the device was 6 months. The reason for the withdrawa...

  11. Process and device to decontaminate a nuclear reactor steam generator

    International Nuclear Information System (INIS)

    The electropolishing technique is used to decontaminate the microtubes of the tube-plate of a steam generator. The present invention proposes and describes a tool; this tool is adapted to a spider type support or another one, and, with the aid of four controled heads with mobile hollow electrode and associated pipes, allow to insert and position an electrode per head inside each of the four microtubes then to inject and extract the electrolyte, the rinsing solution and the contaminated effluents. The tool can be adapted on any handling equipment to treat the surface of any tube

  12. Application of ultrasonic shot peening to steam generator nozzles

    International Nuclear Information System (INIS)

    An effective countermeasure against stress corrosion cracks in nozzle welds is to improve the surface residual stress. A new technique of the ultrasonic shot peening (USP) for steam generator (SG) nozzles will be introduced as a method to improve the residual stress on Alloy 600 Welds. This method changes the compressive stress by applying plastic strain to the surface via the impact force of the shot material during the shot peening. We have successfully performed 14 USP operations in actual plants in Japan. (author)

  13. Steam generator leak detection at Bruce A Unit 1

    International Nuclear Information System (INIS)

    A new steam generator leak detection system was recently developed and utilized at Bruce A. The equipment is based on standard helium leak detection, with the addition of moisture detection and several other capability improvements. All but 1% of the Unit 1 Boiler 03 tubesheet was inspected, using a sniffer probe which inspected tubes seven at a time and followed by individual tube inspections. The leak search period was completed in approximately 24 hours, following a prerequisite period of several days. No helium leak indications were found anywhere on the boiler. A single water leak indication was found, which was subsequently confirmed as a through-wall defect by eddy current inspection. (author)

  14. Flaw tolerance of steam generator tubes under accident conditions

    International Nuclear Information System (INIS)

    Tests were carried out with a bank of tubes in a water tunnel to determine the tolerance of flawed nuclear reactor steam generator tubes to accident conditions which could result in high cross-flow velocities. Fifteen specimen tubes in all were tested, each having one of five types of circumferential slots machined into the outside wall near one end. The tubes were tested at flow velocities sufficient to induce high fluidelastic-type vibrations. All tubes were tested to failure, either until a leak occurred or to complete severance. Failure surfaces were characterized after testing. (author). 5 refs., 2 tabs., 5 figs

  15. Acceptance limits for steam generator tube loss of wall thickness

    International Nuclear Information System (INIS)

    This report presents a review of acceptance limits for steam generator tube loss of wall thickness for Bruce B. The limits are directed at loss of wall resulting from fretting wear experienced at U-bend supports. The report also reviews experience in other countries regarding occurrence of fretting wear and other flow-induced vibration problems, such as fatigue and corrosion fatigue. It reviews contemplated remedial fixes for the Bruce B fretting problems, and develops conclusions and recommendations regarding the remedial approaches and supporting inspection plans. (Author) 93 refs., 9 figs

  16. Steam generator tube rupture: studies to improve plant procedure

    International Nuclear Information System (INIS)

    These accidents have the particularities to lead to atmospheric radioactive release and to be able to be determinated with appropriate operator actions. These radioactive releases are function of several parameters of which sensitivity is analyzed. The major part of the calculations were performed by EDF with an home made code called ''AXEL''. The main conclusions are: - the optimization of the safety injection monitoring to minimize radioactive releases to atmosphere, while ensuring the cooling of the core; - the radioactive releases to atmosphere are very low in any case but much more important if the filling of the steam generator secondary side cannot be avoided

  17. Research of electrochemistry behavior of slower for nuclear steam generator

    International Nuclear Information System (INIS)

    Corrosion of construction material in high temperature water is one of the important problems of nuclear reactor. Lately research shows that TiO2 inhabiting species can alleviate local corrosion, such as stress corrosion cracking (SCC) and intergranular attack (IGA) of steam generator tubes, and thus prolong the life of SG tube. This paper has studied the corrosion potentials and polarization carve of steel stainless in caustic solution with TiO2 inhabiting species, and it shows that TiO2 have positive effect. (authors)

  18. Quadratic controller syntheses for the steam generator water level

    International Nuclear Information System (INIS)

    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)

  19. 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.

  20. Polymeric dispersants for control of steam generator fouling

    International Nuclear Information System (INIS)

    Fouling of steam generators by corrosion products from the feedtrain leads to loss of heat-transfer efficiency, disturbances in thermalhydraulics, and potential corrosion problems resulting from the development of sites for localized accumulation of aggressive chemicals. This report summarizes studies of the use of polymeric dispersants for the control of fouling, which were conducted at the Chalk River Laboratories. High-temperature settling studies on magnetite suspensions were performed to screen available generic dispersants, and the dispersants were ranked in terms of their dispersion efficiency; polyacrylic acid (PAA) and the phosphonate, HEDP, were ranked as the most efficient. Polyacrylic acid was considered more suitable than HEDP for nuclear steam generators, and more emphasis was given to the former in these studies. The dispersants had no effect on the particle deposition rates under single-phase forced-convective flow, but did reduce the deposition rates under flow-boiling conditions. The extent to which the deposition rates were reduced increased in proportion to the dispersant concentration. Preliminary corrosion tests indicated that pitting or general corrosion of steam generator tube materials in the presence of PAA was negligible. Corrosion of carbon steel, although higher in a magnetite-packed crevice under heat flux than in bulk water, was lower in the presence of PAA than in its absence. Some impurities (e.g., sulphate, sodium) were observed in commercially available PAA products at small, though significant concentrations, making these products unacceptable for use in nuclear plants. However, the PAA could be purified by ion exchange. Preliminary experiments, to assess the thermal stability of PAA at steam generator operating temperature, showed the polymer to break down in deaerated solutions and under argon cover to give hydrogen and carbon dioxide as the two major products in the gas phase and variable concentrations of acetate and formate

  1. Radiological assessment of steam generator repair and replacement

    International Nuclear Information System (INIS)

    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

  2. The Economic Evaluation for Kori-1 Steam Generator Replacement

    International Nuclear Information System (INIS)

    The economic evaluation was performed for Kori-1 steam generator(SG) replacement, in which the six senarios were evaluated for a 30, 40 and 50 year plant operating period : Scenario 1-Current Maintenance Approach : Scenario 2-SG Replacement as Early as Possible(1998) : Scenario 3-Scenario 2 + 4.8% Rerate :Scenario 4-18% Plugging Limit : Scenario 5-SG Replacement when Plugging Rate exceeds 15% : Scenario 6-Scenario 5 + 4.8% Rerate. The results of the evaluation indicate that immediate replacement of existing SGs was the most profitable alternative, especially in combination with a 4.8% rerate

  3. Form optimization thermomechanical studies on FBR steam generator steam box made of steel 2 1/4 Cr 1 Mo

    International Nuclear Information System (INIS)

    In breeder power stations the steam generator produces steam at a high temperature and pressure (4950C and 195 bar). Because of the mediocre mechanical characteristics of steel containing 2 1/4 Cr 1 Mo at high temperatures, the use of such a steel for steam generator steam boxes in this reactor system involves a compromise between: - the advantage of large thicknesses, which allow a reduction of the high stresses due to elevated operating pressures, - the rapid penalization on the thermal stresses which results from increasing the thicknesses. The main loads to which the steam outlet box is subjected in modular steam generators of Stein Industrie design are presented, and the steps leading to a satisfactory definition of the structure are explained. It is shown how the choice of rules of the ASME coce N47 leads to the definition of simple stress limiting criteria using Elastic Analysis. Some indications are given on the stresses and principles which motivated the choice of the general form of the structure, and an explanation is given of the simplified method of studying the effect of the head thickness which lay behind the design orientation. Finally, it is shown how the choice of the final design of the steam box was able to be made only after investigation of the best forms by two-dimensional thermomechanical modelling by the finite element method. (orig.)

  4. Design features of Candu 9

    International Nuclear Information System (INIS)

    Thirty-two nuclear generating units with an aggregate installed capacity of 19,119 MWe worldwide are equipped with heavy water moderated and cooled pressure tube reactors of the Canadian Candu line. The list includes nine reactors of the 700 MWe category, and twelve reactors of the 900 MWe category in the Candu 6 series. On the basis of the 900 MWe units, Atomic Energy of Canada Ltd. (AECL) developed the advanced Candu 9 series by evolution. This series has been designed for a service life of sixty years. The use of modular, simplified units and systems in the Candu 9 design is to shorten the planning and construction phase, increase safety, and improve plant operation. AECL will offer this reactor on the world market, first to its customers in (South) Korea, which is one of the reasons why the safety parameters have been chosen especially under the aspect of seismic characteristics. (orig.)

  5. CANDU passive shutdown systems

    International Nuclear Information System (INIS)

    CANDU incorporates two diverse, passive shutdown systems (Shutdown System No. 1 and Shutdown System No. 2) which are independent of each other and from the reactor regulating system. Both shutdown systems function in the low pressure, low temperature, moderator which surrounds the fuel channels; the shutdown systems do not penetrate the heat transport system pressure boundary. The shutdown systems are functionally different, physically separate, and passive since the driving force for SDS1 is gravity and the driving force for SDS2 is stored energy. The physics of the reactor core itself ensures a degree of passive safety in that the relatively long prompt neutron generation time inherent in the design of CANDU reactors tend to retard power excursions and reduces the speed required for shutdown action, even for large postulated reactivity increases. All passive systems include a number of active components or initiators. Hence, an important aspect of passive systems is the inclusion of fail safe (activated by active component failure) operation. The mechanisms that achieve the fail safe action should be passive. Consequently the passive performance of the CANDU shutdown systems extends beyond their basic modes of operation to include fail safe operation based on natural phenomenon or stored energy. For example, loss of power to the SDS1 clutches results in the drop of the shutdown rods by gravity, loss of power or instrument air to the injection valves of SDS2 results in valve opening via spring action, and rigorous self checking of logic, data and timing by the shutdown systems computers assures a fail safe reactor trip through the collapse of a fluctuating magnetic field or the discharge of a capacitor. Event statistics from operating CANDU stations indicate a significant decrease in protection system faults that could lead to loss of production and elimination of protection system faults that could lead to loss of protection. This paper provides a comprehensive

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

    OpenAIRE

    Vladimir Melikhov; Oleg Melikhov; Yury Parfenov; Alexey Nerovnov

    2011-01-01

    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 dete...

  7. CANDU development

    International Nuclear Information System (INIS)

    Evolution of the 950 MW(e) CANDU reactor is summarized. The design was specifically aimed at the export market. Factors considered in the design were that 900-1000 MW is the maximum practical size for most countries; many countries have warmer condenser cooling water than Canada; the plant may be located on coastal sites; seismic requirements may be more stringent; and the requirements of international, as well as Canadian, standards must be satisfied. These considerations resulted in a 600-channel reactor capable of accepting condenser cooling water at 320C. To satisfy the requirement for a proven design, the 950 MW CANDU draws upon the basic features of the Bruce and Pickering plants which have demonstrated high capacity factors

  8. Structural analysis of steam generator internals following feed water main steam line break: DLF approach

    International Nuclear Information System (INIS)

    In order to evaluate the possible release of radioactivity in extreme events, some postulated accidents are analysed and studied during the design stage of Steam Generator (SG). Among the various accidents postulated, the most important are Feed Water Line Break (FWLB) and Main Steam Line Break (MSLB). This report concerns with dynamic structural analysis of SG internals following FWLB/MSLB. The pressure/drag-force time histories considered were corresponding to the conditions leading to the accident of maximum potential. The SG internals were analysed using two approaches of structural dynamics. In first approach simplified DLF method was adopted. This method yields an upper bound values of stresses and deflection. In the second approach time history analysis by Mode Superposition Technique was adopted. This approach gives more realistic results. The structure was qualified as per ASME B and PV Code SecIII NB. It was concluded that in all the components except perforated flow distribution plate, the stress values based on elastic analysis are within the limits specified by ASME Code. In case of perforated flow distribution plate during the MSLB transient the stress values based on elastic analysis are higher than the ASME Code limits. Therefore, its limit load analysis had to be done. Finally, the collapse pressure evaluated using limit load analysis was shown to be within the limits of ASME B and PV Code SecIII Nb. (author). 31 refs., 94 figs., 16 tabs

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

    International Nuclear Information System (INIS)

    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

  10. Investigation of techniques for the application of safeguards to the CANDU 600 MW(e) nuclear generating station

    International Nuclear Information System (INIS)

    A cooperative program with the Canadian Atomic Energy Control Board, Atomic Energy of Canada Limited and the IAEA was established in 1975: to determine the diversion possibilities at the CANDU type reactors using a diversion path analysis; to detect the diversion of nuclear materials using material accountancy and surveillance/containment. Specific techniques and instrumentation, some of which are unique to the CANDU reactor, were developed. 10 appendices bring together the relevant reports and memoranda of results for the Douglas Point Program

  11. 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.

  12. Corrosion evaluation of alternate nuclear steam generator tubing materials

    International Nuclear Information System (INIS)

    Several materials were evaluated for use in nuclear steam generators (NSG). These materials were exposed to corrosive conditions representative of those found in nuclear steam generators. The materials evaluated were gold, titanium, tantalum, niobium, Hastelloy C-276, Hastelloy G. Nickel 200, nickel-base Alloy 625, and heat-tracked nickel-base Alloy 600. The test environments simulated acid pitting attack, caustic stress corrosion cracking and reduced sulfur attack. In the pitting environment, the monolithic materials did well, however Nickel 200, nickel-base Alloy 600 and Hastelloy G3 did poorly. The remaining alloys, nickel-base Alloy 625 and Hastelloy C-276 were relatively unaffected in the pitting environment. Tantalum, titanium, niobium, nickel-base Alloy 625 performed poorly in the environment designed to evaluate resistance to caustic cracking. Nickel-base Alloy 600 (stress-relieved), Hastelloy C-276, Hasteloy G-3 and Nickel 200 compared fair to good in the caustic sodium. The gold was unaffected in the hot caustic solution. In the environment selected to represent a reduced sulfur environment, nickel-base Alloy 625 and Hastelloy C-276 exhibited considerable resistance. The nickel-base Alloy 600 was attacked within a relatively short period of time

  13. Experience on detection of leakages in LMFBR-steam generators

    International Nuclear Information System (INIS)

    One of the advantages of long time on full size LMFBR-components is that experience is gained nut only or, the behaviour of components at normal conditions, but also on the operational consequences (real or imaginary) disturbances. One of the most difficult situations that do occur during steam generator operation is the sudden appearance of a leak indication on the hydrogen detectors. It is possible to connect an automatic trip action to the hydrogen detector however, there are reasons not to do so. Spurious signals, which unfortunately do occur rather frequently, can cause unnecessary shut downs. In the case of a very small leak it can be very difficult to locate the leaking steam generator module and to get an impression of the size of the leak. The time available to confirm the leak, locate the component and to take the proper measures is strongly dependent on the leaking rate or translated into a visual signal, on the rate of rise of the hydrogen level shown on the instrument. During the operation of the 50 MW-SCTF at Hengelo experience was obtained with leak indications caused by real and imaginary leaks

  14. MPGV: an expert system for steam generator maintenance

    International Nuclear Information System (INIS)

    Feedback from operating PWR power plants around the world demonstrates that steam generator (SG) tube bundles are among the most sensitive components. They are responsible for a large fraction of forced plant outages and also require and will continue to require a significant maintenance investment. All utilities operating PWR plants are confronted with a number of diverse problems. Only a continuous and sustained strategy can help those utilities maintain SG performance and obtain high reliability and availability at a reasonable cost. The SG maintenance strategy must: . Optimize inspection schedules, . Anticipate potential damage, . Define appropriate remedial actions, . Plan and organize maintenance operations, . Implement preventive action rather than corrective action. It was a logical step to codify this maintenance strategy in a computer code which could be used in lieu of the experts. The use of artificial intelligence (AI) and expert system techniques seemed well adapted since the experts used nonalgorithmic reasoning to solve the problems. Once FRAMATOME had made a final decision, the first step was a feasibility study to decide whether AI was the best solution. The feasibility study produced a demonstration program EXPERTGV which was limited to diagnosing SG tube leaks. The success of the program prompted a second phase during which MPGV, the French acronym for steam generator preventive maintenance, was developed. The current version is limited to SG tube bundles

  15. 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.

  16. Turbine cycle thermal performance analysis according to replacement of steam generator for Kori unit 1

    International Nuclear Information System (INIS)

    After replacement of steam generator of Kori unit 1, electrical output is recovered with increasing of main steam presure and main steam flowrate. For this review, heat balance calculation and performance data review is conducted by using PEPSE code. The Increasement of electrical output is about 10MW. The energy theory is used for analysis of electrical power improvement. As a result, the amount of energy input by main steam is more than amount of irreversivility on this turbine cycle

  17. Local two-phase modeling of the water-steam flows occurring in steam generators

    International Nuclear Information System (INIS)

    The present study is related to the need of modeling the two-phase flows occurring in a steam generator (liquid at inlet and vapour at outlet). The choice is made to investigate a hybrid modeling of the flow, considering the gas phase as two separated fields, each one being modeled with different closure laws. In so doing, the small and spherical bubbles are modeled through a dispersed approach within the two-fluid model, and the distorted bubbles are simulated with an interface locating method. The main outcome is about the implementation, the verification and the validation of the model dedicated to the large and distorted bubbles, as well as the coupling of the two approaches for the gas, allowing the presentation of demonstration calculations using the so-called hybrid approach. (author)

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

    International Nuclear Information System (INIS)

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

  19. Robotized system for removal of slime from the bottom of steam generators

    Science.gov (United States)

    Kucherenko, O. V.; Shvarov, V. A.

    2014-02-01

    Reliability of steam generators depends not only on the main technical characteristics and correctness of the operational mode but also on the cleanliness of the heat-exchange surface and the presence of slime precipitated on the bottom. To provide the cleanliness, chemical methods of cleaning the heatexchange surfaces are used. In this article, we consider the process of removal of sediments that are formed precisely on the bottom of the steam generator from its volume. Possible mechanical methods for removal of sediments are presented. The consideration of variants of cleaning approved for acting steam generators showed the efficiency and applicability of the developed installation for the slime removal from steam generators. The main principles of construction of the system for slime removal from the steam generator bottom and constructive features of the installation, which make it possible to implement the stated tasks on the slime removal from the steam generator bottom, are given.

  20. Crack Growth Prediction of the Steam Turbine Generator Shaft

    International Nuclear Information System (INIS)

    The power network in China is encountering great changes and large-scale network is increasingly implemented for long distance power transmission as well as various kinds of power electronic devices, which bring in the risk of the torsional vibration of the turbine generator shafts, may cause the fatigue damage and cracks in the product life cycle. The paper analyzed the failed coupling of some 600MW steam turbine generator and calculated the local stress of the assembly under torsional load caused by the network disturbance. Then the crack propagation was analyzed with the predicted crack initiation position and crack propagation routine. The assembled coupling contains shaft, coupling and keys with interferences between the parts. Therefore the contact analysis was included. Extended Finite Element Method (X-FEM) is used to calculate the crack propagation and that the mesh needs not to be regenerated with the crack propagation, which is beneficial for engineering applications.