WorldWideScience

Sample records for steam generator sub-assemblies

  1. Environmental Assessment for DOE permission for off-loading activities to support the movement of Millstone Unit 2 steam generator sub-assemblies across the Savannah River Site

    International Nuclear Information System (INIS)

    1992-10-01

    The Department of Energy (DOE) has prepared an Environmental Assessment (EA), for the proposed granting of DOE permission of offloading activities to support the movement Millstone Unit 2 steam generator sub-assemblies (SGSAs) across the Savannah River Site (SRS). Based on the analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, an environmental impact statement is not required, and the Department is issuing this Finding of No Significant Impact. On the basis of the floodplain/wetlands assessment in the EA, DOE has determined that there is no practicable alternative to the proposed activities and that the proposed action has been designed to minimize potential harm to or within the floodplain of the SRS boat ramp. No wetlands on SRS would be affected by the proposed action

  2. Steam 80 steam generator instrumentation

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  3. Steam generator life management

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  4. 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.

  5. Steam generator tube failures

    International Nuclear Information System (INIS)

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

    1996-04-01

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

  6. French steam generator

    International Nuclear Information System (INIS)

    Remond, A.

    1986-01-01

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

  7. Steam generator sludge removal apparatus

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  8. Steam generators - problems and prognosis

    International Nuclear Information System (INIS)

    Tapping, R.L.

    1997-05-01

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

  9. Steam generator comprehensive testing facility

    International Nuclear Information System (INIS)

    Chen Changbing

    1998-12-01

    The main system and equipment of a capacity of 30 MW steam generator testing rig is briefly introduced. The thermal performance of steam generator simulator has been verified with a steam generator model comprehensive testing. The application and function of the Facility are analyzed. And the effective use of the equipment of the Facility is also proposed

  10. Steam generator management program

    International Nuclear Information System (INIS)

    Cho, Nam Cheoul; Kim, Moo Soo; Lee, Kwang Woo

    2003-01-01

    Recently, the common concern of nuclear power industry in the development of technology mitigating and preventing the aging of steam generator tubes prevails, because the trends of steam generator flaws at Uljin unit no. 1,2 and KSNP impose a burden on the operation of nuclear power plant. While the regulatory agency is demanding the establishment of the advanced general performance maintenance system, the steam generator management program adapting advanced technology is being developed which may comply with EPRI PWR SG guidelines based on NEI 97-06. General guidelines including all the maintenance aspects consist of the tube integrity assessment criteria, repair limit, allowable leakage level, water chemistry will be composed in order to obtain the approval of regulatory agency and be applied to nuclear power plant early 2005. This presentation is to introduce maintenance state including SG tube degradation and main contents of advanced SG management program being developed, and furthermore update present and future plan, and estimate the alternation after the completion

  11. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1984-10-01

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

  12. Steam generator materials

    International Nuclear Information System (INIS)

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

    2002-05-01

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

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

    International Nuclear Information System (INIS)

    1982-10-01

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

  14. Steam generator accident protection

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.

    1986-01-01

    The outbreak of an accident in separate heat exchange unit of a steam generator (all the heat exchange units being connected to an alkali metal circuit) will activate a stop valve and a relief valve. The two valves are placed at the input and output pipes of the damaged unit. The same valves will also close after a very short time the inflow of metal into the disturbed heat exchange unit and this will immediately, without any pressure shock, pipe the metal into an equalizer tank; they will also close the outflow of metal which together with the reaction products will be piped into the discharge tank. At the same time the stop valves on the feed water pipe are closed. The whole equipment shows excellent coordination and has standby elements which are put into operation when any of the elements fail. The failed unit is thus safely separated from the other units and may be independently repaired. (J.B.)

  15. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1983-08-01

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

  16. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  17. Steam generator reliability improvement project

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  18. Steam generator reliability improvement project

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. Steam generators of PWRs and plan of replacing steam generators

    International Nuclear Information System (INIS)

    Dodo, Takashi; Matsushita, Kiyohiko.

    1995-01-01

    Also in Japan, the plan of replacing steam generators which are one of most important equipments of PWRs is being advanced. In this report, the whole circumstances of the plan of replacing steam generators, such as the background that necessitated the replacement, the concrete plan of the replacement, the results of the replacement in foreign countries and others, are described in general. The function of steam generators is to form a part of the pressure boundary for confining reactor primary coolant, and to generate steam by the heat exchange between primary and secondary coolants. The basic structure is a shell and tube type heat exchanger having vertically installed, inverse U shape tubes, in which primary coolant flows. In Japan as of December, 1994, 22 PWR plants of 18,186 million kW output were in operation. The management of steam generators and the background of their replacement due to the fear of public about heating tube leak, the enormous cost of the inspection and repair, the restriction of power system operation and others are explained. As for the steam generators of Takahama No. 2, Oi No. 1 and Genkai No. 1 plants, the application for the replacement was made, and the breakdown of a tube occurred in Mihama No. 2 plant. The replacement of steam generators in Japan and foreign countries is reported. (K.I.)

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

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

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

  1. 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.

  2. 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)

  3. CANDU steam generator life management

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  4. Specification of steam generator, condenser and regenerative heat exchanger materials for nuclear applications

    International Nuclear Information System (INIS)

    Jovasevic, J.V.; Stefanovic, V.M.; Spasic, Z.LJ.

    1977-01-01

    The basic standards specifications of materials for nuclear applications are selected. Seamless Ni-Cr-Fe alloy Tubes (Inconel-600) for steam generators, condensers and other heat exchangers can be employed instead of austenitic stainless steal or copper alloys tubes; supplementary requirements for these materials are given. Specifications of Ni-Cr-Fe alloy plate, sheet and strip for steam generator lower sub-assembly, U-bend seamless copper-alloy tubes for heat exchanger and condensers are also presented. At the end, steam generator channel head material is proposed in the specification for carbon-steel castings suitable for welding

  5. Steam generator blowdown system upgrades

    International Nuclear Information System (INIS)

    Lee, Tien P.; Kim, David H.; Jindal, Krishan K.

    2004-01-01

    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)

  6. Steam generator thermal-hydraulics

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  7. Research program plan: steam generators

    International Nuclear Information System (INIS)

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

    1985-07-01

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

  8. Options for Steam Generator Decommissioning

    International Nuclear Information System (INIS)

    Krause, Gregor; Amcoff, Bjoern; Robinson, Joe

    2016-01-01

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

  9. Model of reverse steam generator

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  10. Water box for steam generator

    International Nuclear Information System (INIS)

    Lecomte, Robert; Viaud, Michel.

    1975-01-01

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

  11. Design of SMART steam generator cassette

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  12. Steam generator tube integrity program

    International Nuclear Information System (INIS)

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

    1996-01-01

    The degradation of steam generator tubes in pressurized water nuclear reactors continues to be a serious problem, and the US Nuclear Regulatory Commission (NRC) is developing a performance-based rule and regulatory guide for steam generator tube integrity. To support the evaluation of industry-proposed implementation of these performance-based criteria, the NRC is sponsoring a new research program at Argonne National Laboratory on steam generator tubing degradation. The objective of the new program is to provide the necessary experimental data and predictive correlations and models that will permit the NRC to independently evaluate the integrity of steam generator tubes. The technical work in the program is divided into four tasks, (1) assessment of inspection reliability, (2) research on in-service inspection technology, (3) research on degradation modes and integrity, and (4) development of methodology and technical assessments for current and emerging regulatory issues. The objectives of and planned research activities under each of these four tasks are described here. (orig.)

  13. Steam generator waterlancing at DNGS

    International Nuclear Information System (INIS)

    Seppala, D.; Malaugh, J.

    1995-01-01

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

  14. Steam generator tube vibration study

    International Nuclear Information System (INIS)

    Enderlin, W.I.

    1986-01-01

    Chemical cleaning has been proposed to remove magnetite buildup in some pressurized water reactor steam generators. The US Nuclear Regulatory Commission (NRC) has expressed concern that such cleaning would combine with the tube denting caused by magnetite formation to enlarge tube/tube support plate clearances, increasing the level of flow-induced vibrations that could lead to unacceptably high tube wear and failure rates. In support of NRC, the Pacific Northwest Laboratory investigated whether such increased clearances would exacerbate tube fretting wear. Using a full-length scale model of a steam generator tube bundle, flow tests were conducted at an instrumented location through clearances representing as-built and post-cleaned tube conditions. Test results indicated little potential for increased tube wear as a result of chemical cleaning, under normal operating conditions at tube support locations similar to that tested

  15. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

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

  16. 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.

  17. Horizontal steam generator thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  18. Designing steam generators for nuclear power plants

    International Nuclear Information System (INIS)

    Hanak, D.

    1989-01-01

    The existing types, their performance, assets and shortcomings are given for vertical steam generators manufactured by Combustion Engineering and by Westinghouse, for through-flow steam generators with slight overheating of the withdrawn steam manufactured by Babcock, and for horizontally positioned WWER-440 and WWER-1000 steam generators. The steam generator for the WWER-1000 reactors of the Temelin nuclear power plant is dealt with in detail. Its design and structural materials used are given. The procedure for strength calculations of the pressure parts of the steam generator by the finite elements method, the in-service diagnostics system as well as the device for simulating corrosion phenomena in steam generators are described. (E.J.). 6 figs., 26 refs

  19. Operating experience with steam generators

    International Nuclear Information System (INIS)

    Bouecke, R.

    1991-01-01

    In contrast to steam generator tube degradation problems that have been widely encountered worldwide, steam generators of the Siemens/KWU design have proven by operating experience that they are very efficient in minimizing tube corrosion or any other SG related problems. The paper will substantiate this statement by addressing the performance characteristics of nearly 20 years of operation experience. Emphasis is put on evaluations comparing the heat transfer capacity of Incoloy 800 with that of Inconel 690 TT. Various tube support designs are discussed with respect to hide-out behaviour. Recent evaluations confirm the superiority of grid type tube support designs compared to tube support plates. Anti vibration bars acc. to Siemens/KWU design allow proper support even of the innermost U-tubes, by which excessive vibration induced tube failures due to fatigue are ruled-out. Steam generators are key components which can heavily effect plant safety and availability. This was recognized by Siemens/KWU at a very early stage of the SG design. A multi level concept was developed and consequently applied, the characteristics of which can be highlighted as follows: - Implementation of specific design features after careful experimental and/or analytical verification; - Material selection based on profound validation tests; - Stringent inspection requirements regarding control of manufacturing; - Deliberate specification and control of water chemistry guidelines; - Close feedback of operational problems to be readily considered in design improvements or remedial actions to be taken. A strict application of this concept has reached a stage which allows the following summarizing statements: - All main design features of the Siemens/KWU SG are in use - basically unchanged - for roughly 20 years; - All of them are verified by excellent operating experience and available for implementation in any replacement or new SG design; - No need for replacement of Siemens/KWU SG is

  20. Thermal-hydraulics in recirculating steam generators

    International Nuclear Information System (INIS)

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

    1981-04-01

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

  1. Forced circulation steam generators for SAGD applications

    Energy Technology Data Exchange (ETDEWEB)

    Vasudevan, M [Engineered Boiler Systems (United States)

    2011-07-01

    Steam for steam assisted gravity drainage (SAGD) is traditionally supplied by once-through steam generators (OTSG) or drum type boilers. Research by Cleaver Brooks aims to combine the quality features of OTSGs and drum boilers and to address issues encountered with both these types of machinery. The forced-circulation oil sands steam generator (FC-OSSG) resulted from the design of a flow management system which allowed water quality upsets to be handled successfully. Integration of the boiler and the burner design through computational fluid dynamics (CFD) led to a decrease in the environmental footprint of the steam generator. Circuit maintenance was also significantly facilitated by elaborating new piggable circuits. Moreover, the design of a water cooled furnace enabled the generator to work at constant heat. The resulting FC-OSSG produces cleaner steam than either OTSGs or drum boilers, and has a higher capacity than OTSGs. Additionally, turndown, blowdown and feedwater quality are similar to that obtained using drum boilers.

  2. Method to detect steam generator tube leakage

    International Nuclear Information System (INIS)

    Watabe, Kiyomi

    1994-01-01

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

  3. Steam generator leak detection using acoustic method

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  4. 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.

  5. LMR steam generator blowdown with RETRAN

    International Nuclear Information System (INIS)

    Wei, T.Y.C.

    1985-01-01

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

  6. Third international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

    1995-01-01

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

  7. CAREM-25 Steam Generator Stability Analysis

    International Nuclear Information System (INIS)

    Rabiti, A.; Delmastro, D.

    2003-01-01

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

  8. Steam generator tube denting simulation testing

    International Nuclear Information System (INIS)

    Battaglia, P.J.; Singley, W.J.

    1978-02-01

    Tube denting has been reported in steam generators in a number of commercial nuclear power plants in recent years. In order to aid in understanding of the mechanism leading to tube denting in the steam generators, a Bettis laboratory test program was initiated to attempt to reproduce tube denting and to investigate the effects of chemistry, design, and temperature. The results of the tests indicate that denting can be reproduced in the small model steam generator test apparatus devised for this testing. Denting was observed under carbon steel support plates in seawater-contaminated secondary water in a test with hydrazine-ammonia chemistry and a primary water inlet temperature of 590 0 F and in a test with hydrazine-morpholine chemistry and a primary water inlet temperature of 545 0 F. The parameters of these two tests simulate conditions in a typical commercial steam generator and the Shippingport Atomic Power Station steam generators, respectively

  9. Hockey-stick steam generator for LMFBR

    International Nuclear Information System (INIS)

    Hallinan, G.J.; Svedlund, P.E.

    1981-01-01

    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

  10. Circumferential cracking of steam generator tubes

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    1997-04-01

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

  11. Maintenance and repair of LMFBR steam generators

    International Nuclear Information System (INIS)

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

    1984-06-01

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

  12. Strategic maintenance plan for Cernavoda steam generators

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  13. Steam Generator Inspection Planning Expert System

    International Nuclear Information System (INIS)

    Rzasa, P.

    1987-01-01

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

  14. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

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

  15. 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

  16. In service inspection for steam generator tubes

    International Nuclear Information System (INIS)

    Comby, R.; Eyrolles, Ph.

    1988-01-01

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

  17. New steam generators slated for nuclear units

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

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

  18. Hydrogen generation from steam reaction with tungsten

    Science.gov (United States)

    Smolik, G. R.; McCarthy, K. A.; Petti, D. A.; Coates, K.

    1998-10-01

    A LOCA in a fusion reactor involving an ingress of steam presents a safety concern due to hydrogen generated from steam reactions with plasma facing components. Hydrogen concentrations must be maintained below explosive levels. To support safety evaluations we have experimentally determined hydrogen generation rates when a tungsten alloy is exposed to steam from 400°C to 1200°C. We studied effects of steam pressure between 2.8 × 10 4 and 8.5 × 10 4 Pa, i.e., (0.28-0.84 atm) and gas velocity between 0.011 and 0.063 m/s. We present relationships for the reaction rates, oxidation phases, and mechanisms associated with the hydrogen generation.

  19. 49 CFR 229.105 - Steam generator number.

    Science.gov (United States)

    2010-10-01

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

  20. IAEA activities on steam generator life management

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  1. 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.

  2. Fourth international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  3. Program for parameter studies of steam generators

    International Nuclear Information System (INIS)

    Mathisen, R.P.

    1982-11-01

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

  4. Acoustic leak detector in Monju steam generator

    International Nuclear Information System (INIS)

    Wachi, E.; Inoue, T.

    1990-01-01

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

  5. Stability study in one step steam generators

    International Nuclear Information System (INIS)

    Abbate, P.

    1990-01-01

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

  6. 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)

  7. KNK steam generator leakage, evaluations and improvements

    International Nuclear Information System (INIS)

    Dumm, K.; Ratzel, W.

    1975-01-01

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

  8. Eddy current testing of steam generator tubes

    International Nuclear Information System (INIS)

    Neumaier, P.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

  10. Leakage experiences with 1 MW steam generator

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  11. Solar energy for steam generation in Brazil

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-05-01

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

  12. 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.

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

    International Nuclear Information System (INIS)

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

    1986-10-01

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

  14. Steam-generator replacement sets new marks

    International Nuclear Information System (INIS)

    Beck, R.L.

    1995-01-01

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

  15. Future aspects for liquid metal heated steam generators

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  16. Embalse steam generators - status in 2009

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  17. Acoustic leak detection of LMFBR steam generator

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi; Yoshida, Kazuo

    1993-01-01

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

  18. Solar steam generation by heat localization.

    Science.gov (United States)

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

    2014-07-21

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

  19. Three Steam Generator Replacement Projects in 1995

    International Nuclear Information System (INIS)

    Holz, R.; Clavier, G.

    1996-01-01

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

  20. A steam generator maintenance and repair robot

    International Nuclear Information System (INIS)

    Eccleston, M.J.

    1987-01-01

    A robot is described which allows the maintenance and repair of Model 44 and 51 steam generators with minimum operator received dose. The manipulator loads itself into the steam generator channel head through the manway, and locks itself to the tube plate, from where it can reach out to accept tools and inspection equipment. At no stage in the operation of the robot does the operator have to break the plane of the manway or stand in the direct shine path from the tube plate. The robot design is based on a manipulator successfully developed for use throughout the British nuclear industry. At the time of writing five such manipulators were in manufacture for applications in fuel element remote handling and fuel reprocessing plant repair. Proposals are being considered for further applications in reactor decommissioning, underwater fuel handling and general remote maintenance

  1. 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.

  2. EP 1000 steam generator tube rupture analyses

    International Nuclear Information System (INIS)

    Saiu, G.; Frogheri, M.; Schulz, T.L.

    2001-01-01

    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

  3. Fifth CNS international steam generator conference

    International Nuclear Information System (INIS)

    2006-01-01

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

  4. U-tube steam generator predictions

    International Nuclear Information System (INIS)

    Hassan, Y.A.; Kalyanasundaram, M.

    1991-01-01

    This paper reports on the development of a RELAP5/MOD2 computer code model for a Model Boiler-2 U-tube steam generator (UTSG) to predict the thermal-hydraulic response of a UTSG during steady-state operation and for a loss-of-feedwater (LOF) transient. Steady-state conditions calculated by RELAP5 are compared with the measured data. The calculated heat transfer from the primary to the secondary side of the steam generator is found to be underpredicted by 30%. The heat transfer correlations used in existing thermal-hydraulic codes are developed for flow inside individual tubes and not for flow around tube bundles. Consequently, the secondary convective heat transfer is not accurately predicted by the codes. A revised version of the RELAP5 code with modified heat transfer correlations reasonably predicts the primary to the secondary heat transfer in bundle environments. Improved heat fluxes and heat transfer coefficients are obtained during steady-state and LOF accident transients. Steady-state behavior of the Semiscale MOD-2C steam generator is also computed with both the original and the revised versions of the code. Good agreement is achieved between the predictions and the test data when the modified heat transfer correlations are utilized

  5. Fracture toughness determination in steam generator tubes

    International Nuclear Information System (INIS)

    Bergant M; Yawny, A; Perez Ipina, J

    2012-01-01

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

  6. Minimize corrosion degradation of steam generator tube materials

    International Nuclear Information System (INIS)

    Lu, Y.

    2006-01-01

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

  7. Ferritic steels for French LMFBR steam generators

    International Nuclear Information System (INIS)

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

    1983-06-01

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

  8. Steam generator of FBR type reactor

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    DeFur, D.D.

    1975-04-01

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

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

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.

    1986-01-01

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

  11. Analysis of fast reactor steam generator performance

    International Nuclear Information System (INIS)

    Hulme, G.; Curzon, A.F.

    1992-01-01

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

  12. 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.

  13. Corrosion Evaluation and Corrosion Control of Steam Generators

    International Nuclear Information System (INIS)

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

    2008-06-01

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

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

    International Nuclear Information System (INIS)

    1984-04-01

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

  15. Structural considerations in steam generator replacement

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  16. LMFBR steam generators in the United Kingdom

    International Nuclear Information System (INIS)

    Anderson, R.; Hayden, O.

    2002-01-01

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

  17. Nuclear fuel sub-assemblies

    International Nuclear Information System (INIS)

    Dodd, J.A.; Butterfield, C.E.; Waite, E.

    1979-01-01

    A fast reactor fuel sub-assembly has honeycomb grids for laterally supporting the fuel pins. The grids are of two series and are arranged alternately along the bundle. The grids of a first series provide a discrete cell for each pin but the grids of the second series have a peripheral group of cells only. The grids of the second series provide intermediate support of the edge pins to restrain bow. (author)

  18. Steam generator replacement from ALARA aspects

    International Nuclear Information System (INIS)

    Terry, I.; Breznik, B.

    2003-01-01

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

  19. Steam generator of the forced circulation type

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Ramaley, D.

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  2. Hideout in steam generator tube deposits

    International Nuclear Information System (INIS)

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

    1998-05-01

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

  3. 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.

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

    International Nuclear Information System (INIS)

    Ding Xunshen; Cui Baoyuan; Xue Yunkui; Liu Shixun

    1989-01-01

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

  5. Heysham II/Torness AGR steam generator

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  6. Steam generator design requirements for ACR-1000

    International Nuclear Information System (INIS)

    Subash, S.; Hau, K.

    2006-01-01

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

  7. Remotely installed steam generator nozzle dam system

    International Nuclear Information System (INIS)

    Mc Donald, F.X.; Weisel, E.M.; Schukei, G.E.

    1990-01-01

    This patent describes a method for remotely installing a dam unit in a nozzle or a nuclear steam generator head, the head including a manway. It comprises: mounting an articulated manipulator to an internal surface of the head, the manipulator having a free end which carries a jaw member; positioning the manipulator so that the jaw member is adjacent the manway and substantially on the manway axis; passing a first dam segment through the manway and attaching the jaw member to the first segment; positioning the manipulator so that the jaw member holds the first dam segment on one side of the manway axis; passing a second dam segment through the manway into engagement with the first dam segment to form a dam subassembly; translating the manipulator through the head until the dam subassembly is adjacent the nozzle; advancing the jaw member toward the nozzle until the cam subassembly is positioned substantially at the desired location of the dam unit with respect to the nozzle; and deploying the manipulator to install dam support structure between the dam subassembly and the steam generator, thereby forming an installed dam unit

  8. Repair technology for steam generator tubes

    International Nuclear Information System (INIS)

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

    2001-02-01

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

  9. 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.

  10. Research on safety of FBR steam generators

    International Nuclear Information System (INIS)

    Tanabe, Hiromi; Himeno, Yoshiaki; Kuroha, Mitsuo.

    1988-01-01

    At the initial stage, the research on ensuring the soundness of plants at the time of the sodium-water reaction due to large leak was a main purpose, but it was elucidated that the sodium-water reaction in steam generators does not threaten the safety of reactors, and the soundness of plants can be ensured even at the time of the conceivable largest leak. Recently, from the viewpoint of improving the economical efficiency of plants, the research has been changed to the elucidation of the behavior of small scale leak. The classification of sodium-water reaction phenomena, the investigation into the cases of accidents in foreign countries, and the setting-up of an accident scenario are described. The research on sodium-water reaction in the world has been carried out for four purposes, that is, to ensure the soundness of secondary systems against large leak accidents, to grasp leak propagation phenomena and to select the design base leak, to develop leak detection system and to learn the method of operation and taking countermeasures at the time of leak accidents. The present status of the research on these items in Japan and foreign countries is reported. Aiming at the practical use of FBRs, as the means of the cost reduction in plant construction, the omission of intermediate system is planned, and double walled tubes may be used for steam generators. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

    The Steam Generator (SG) tubing degradation caused by corrosion and other age-related mechanisms continues to be a significant safety and cost concern for many Nuclear Power Plants (NPP). The understanding of the steam generator ageing mechanisms is the key to effective management of steam generator ageing and consists of the knowledge of steam generator materials and these one properties, stressors and operating conditions, like degradation sites and wear mechanisms. The principal types of corrosion are presented which can occur in CANDU steam generator. There are also presented the operation conditions, the specifications referring to the water chemistry and the construction materials of Steam Generator, the factors that have a great influence on the corrosion behaviour during the whole exploitation period of this equipment. (R.P.)

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

    International Nuclear Information System (INIS)

    Ubra, O.; Doubek, M.

    1995-01-01

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

  13. Future development LMFBR-steam generators SNR2

    International Nuclear Information System (INIS)

    Essebaggers, J.; Pors, J.G.

    1975-01-01

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

  14. Dismantling of the 50 MW steam generator test facility

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  15. Corrosion aspects in steam generators of nuclear power plants

    International Nuclear Information System (INIS)

    Visoni, E.; Santos Pinto, M. dos

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  17. 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

  18. 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)

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

    2002-01-01

    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

  19. Comments on US LMFBR steam generator base technology

    International Nuclear Information System (INIS)

    Simmons, W.R.

    1984-01-01

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

  20. Internal ultrasonic testing of steam generator tubes

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  1. Radioactivity transport following steam generator tube rupture

    International Nuclear Information System (INIS)

    Hopenfeld, J.

    1985-03-01

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

  2. Development of advanced PWR steam generator

    International Nuclear Information System (INIS)

    Saito, Itaru; Nakamura, Tomomichi

    1999-01-01

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

  3. Nuclear-reactor steam-generator shut-off valve

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1978-01-01

    In a nuclear reactor installation a live-steam line leading from each steam generator through the containment is equipped with a fast-acting valve. If the pressure in the generator rises the valve operates as a safety valve and releases up to one-half of the aperture cross section of the valve, so that there is no danger of damage to the steam generators. According to the invention, the valve is operated for this purpose by two pistons by means of steam. The invention is of interest particularly for light-water reactors, e.g., pressurized-water reactors

  4. Modeling and Simulation of U-tube Steam Generator

    Science.gov (United States)

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

    2018-03-01

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

  5. Comparison of steam generator methods in PISC

    International Nuclear Information System (INIS)

    Lahdenperae, K.; Kankare, M.

    1996-01-01

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

  6. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

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

  7. 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.

  8. 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...

  9. Optimal design of vertical natural circulation steam generator weight

    International Nuclear Information System (INIS)

    Qin Huimin; Yan Changqi; Wang Jianjun; Wang Meng; He Shijing

    2011-01-01

    The steam generator is one of the key equipment in the PWR nuclear power plant, which has relative large weight and dimension, nowadays, it tends to using high capacity steam generators, which makes the steam generators to be heavier and larger, and it causes difficulties to the transport and arrangement of steam generators. Therefore, it is necessary to apply optimization techniques to the design of the steam generator in order to achieve the minimum weight or Janume by means of finding the optimum combination of design parameters while satisfying the heat exchange requirements and safety constraints. In this work, the evaluation model and corresponding codes of a vertical natural circulation steam generator were established, and based on the sensitivity analysis of thermohydraulic parameters and structural parameters which influence steam generator weight, the optimal design of steam generator, taking the net weight as objective, was carried out using a complex-genetic algorithm. The optimal results indicate that the net weight of optimum design is 17.16%, less than original design. The result shows an obvious effect which proves the possibility of optimization and can provide reference for real engineering design. (authors)

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

    International Nuclear Information System (INIS)

    Essebaggers, J.

    1975-01-01

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

  11. Ultrasonic enhancement of chemical cleaning of steam generators. Final report

    International Nuclear Information System (INIS)

    Scharton, T.

    1983-04-01

    This report presents the results of an investigation into the use of ultrasound to enhance the chemical cleaning of steam generator tube and support crevices. Primary attention was focused on a configuration with ultrasonic transducers in the downcomer region of the steam generator in conjunction with the EPRI Steam Generator Owners Group (SGOG) crevice solvent at 200 0 F. The investigation consisted primarily of experiments conducted in facilities designed to simulate the geometry and acoustics of a steam generator. The largest facility holds approximately 1000 gallons of solvent and simulates a 40 0 sector of a steam generator with two support plates. The testing demonstrated that ultrasonics is indeed an effective means of enhancing the crevice cleaning if sound levels sufficient to cause cavitation can be transmitted to the crevices. The effort focused on the coupling of the transducers to the wrapper plate, on the transmission of sound through the tube bundle, and on the determination of cavitation threshold levels

  12. Equipment transporter for nuclear steam generator

    International Nuclear Information System (INIS)

    Hayes, L.R.

    1987-01-01

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

  13. Nuclear reactor fuel sub-assemblies

    International Nuclear Information System (INIS)

    Dodd, J.A.

    1981-01-01

    An improved fuel sub-assembly for a liquid metal cooled fast breeder reactor, is described, in which fatigue damage due to buffeting by cross-current flows is reduced and protection is provided against damage by contact with other reactor structures during loading and unloading of the sub-assembly. (U.K.)

  14. Steam generator for FBR type reactor

    International Nuclear Information System (INIS)

    Watabe, Ichiro

    1998-01-01

    In a steam generator for an FBR type reactor, a heating fluid is introduced from outer circumferential one end of a cylindrical main body by way of a heating fluid distribution structure main body, put to heat exchange with helical coils disposed inside of the main body and then discharged from the other end of the main body, the distribution structure comprises a heating fluid entrance connected to and passed through the outer one circumferential end of the main body and an inner distribution duct comprising a cylindrical body having a large number of small distribution apertures on the circumference and having ring-like flanges protruded outwardly on both ends, with the end of one flange being secured to the inner surface of the main body so that an entrance for heating fluid is opened between both of the flanges. Since the other flange is left free, there is no restriction between the main body and the inner distribution duct, and the heating fluid distribution structure can be free from thermal stresses. (N.H.)

  15. Steam generator for FBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Watabe, Ichiro

    1998-11-24

    In a steam generator for an FBR type reactor, a heating fluid is introduced from outer circumferential one end of a cylindrical main body by way of a heating fluid distribution structure main body, put to heat exchange with helical coils disposed inside of the main body and then discharged from the other end of the main body, the distribution structure comprises a heating fluid entrance connected to and passed through the outer one circumferential end of the main body and an inner distribution duct comprising a cylindrical body having a large number of small distribution apertures on the circumference and having ring-like flanges protruded outwardly on both ends, with the end of one flange being secured to the inner surface of the main body so that an entrance for heating fluid is opened between both of the flanges. Since the other flange is left free, there is no restriction between the main body and the inner distribution duct, and the heating fluid distribution structure can be free from thermal stresses. (N.H.)

  16. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Smith, J.C.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Ding Xunshen

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Kawamura, H.; Meister, G.

    1982-10-01

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

  20. Coal fired steam generation for heavy oil recovery

    International Nuclear Information System (INIS)

    Firmin, K.

    1992-01-01

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

  1. Operational studies of thermal-hydraulic unit steam generators

    International Nuclear Information System (INIS)

    Shejnkman, A.G.; Bel'tyukov, A.I.; Vylomov, V.V.; Karpenko, A.I.; Lyzhin, A.A.; Smirnov, M.V.; Khaletskij, Eh.Eh.

    1989-01-01

    Results of the complex of tests, experimental studies and operational measurements of thermal-hydraulic characteristics of the PGM-200M steam generators conducted in the period of start-up and bringing-up to full capability at the BN-600 power unit are presented. Analysis of the data obtained shows that the steam generator comprising eight sections ensures heat removal at the nominal power with seven operating sections. The uniformity of coolant flow rate distribution over the steam generator sections practically does not depend on the number of operating sections

  2. Three Mile Island Nuclear Station steam generator chemical cleaning

    International Nuclear Information System (INIS)

    Hansen, C.A.

    1992-01-01

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

  3. A reflux capsule steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Lantz, E.

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

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

    International Nuclear Information System (INIS)

    Rosengaus, J.

    1987-01-01

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

  5. Prairie Island Nuclear Generating Plant steam generator owners group II: examination of 3 tubes removed from steam generator No. 12

    International Nuclear Information System (INIS)

    Kuchirka, P.; Madeyski, A.; Pearson, R.

    1986-01-01

    The No. 12 steam generator at Prairie Island Unit 1 has experienced some degradation of the hot leg tubes within the tube sheet region. The history of the degradation is given. A comparison of the rate of degradation of No. 12 steam generator with similar rates at Point Beach and Ginna is presen. The No. 11 steam generator has not experienced similar degradation. To investigate the cause and degree of degradation and to obtain correlation of field eddy current (EC) indications with actual conditions, the three hot leg tube segments were removed from the No. 12 steam generator for laboratory EC testing and destructive examination. Two of the tubes had field EC indications in the tube sheet region and one was apparently free of indications. The results of the testing and examinations are given. These tests showed that the lab EC correlated well with the destructive exam, but the field EC did not. The lab EC detected defects that are >35% thru wall, whereas the field EC detected defects that are >80%. The principle degradation was intergranular stress corrosion cracking in the tube sheet region with some uniform intergranular attack. The clean tube had randomly distributed IGA with a maximum depth of 15%. tube/tube sheet deposit chemical analysis does not support the existence of a caustic environment. The conclusions of this work are given

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-12-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  9. Devices for investigation and intervention on steam generators tubes bundles

    International Nuclear Information System (INIS)

    Launay, J.P.; Sort, M.

    1986-01-01

    After a brief recall on the French regulation concerning pressure vessels, the authors describe the experience and the devices used by Framatome for closing, repairing, sleeving and shot peening for steam generators tubes bundles [fr

  10. Technical development and its application on steam generator replacement

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. Improvements in steam cycle electric power generating plants

    International Nuclear Information System (INIS)

    Bienvenu, Claude.

    1973-01-01

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

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

    International Nuclear Information System (INIS)

    Frank, L.; Stokley, J.

    1988-06-01

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

  13. Improvements during fabrication of the Spanish steam generators

    International Nuclear Information System (INIS)

    Alvarez Miranda, A.

    1994-01-01

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

  14. Steam generator tube integrity program. Phase I report

    International Nuclear Information System (INIS)

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

    1979-09-01

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

  15. 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.

  16. In-service chemical diagnostics of WWER-1000 steam generator

    International Nuclear Information System (INIS)

    Cervinka, J. et al.

    1989-01-01

    A chemical diagnostics subsystem is proposed for WWER-1000 steam generators. It consists of two parts. The analytical module serves indirect examination of the effect of the water chemistry of the steam generator secondary side on the structural materials involved. Samples are analyzed for the presence of corrosion-aggressive substances, which can cause or accelerate corrosion damage, and of corrosion products. The quantities measured include electrolytic conductance, pH value, redox potential and oxygen content and concentrations of selected anions. The other part of the chemical diagnostics subsystem provides direct measurement of the degree of salt concentration in the steam generator water by means of microconductivity sensors. Each steam generator should be equipped with three sensors. (Z.M.). 2 figs

  17. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

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

    1997-01-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

  18. Dynamic and control of a once through steam generator

    International Nuclear Information System (INIS)

    Gomes, Arivaldo Vicente

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    1980-03-01

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

  20. Acoustic detection for water/steam leak from a tube of LMFBR steam generator

    International Nuclear Information System (INIS)

    Sonoda, Masataka; Shindo, Yoshihisa

    1989-01-01

    Acoustic leak detector is useful for detecting more quickly intermediate leak than the existing hydrogen detector and is available for identification of leak location on the accident of water/steam leak from a tube of LMFBR steam generator. This paper presents the overview of HALD (High frequency Acoustics Leak Detection) system, which is more sensitive for leak detection and lower cost of equipment for identification of leak location than a low frequency type detector. (author)

  1. Automated Diagnosis and Classification of Steam Generator Tube Defects

    International Nuclear Information System (INIS)

    Garcia, Gabe V.

    2004-01-01

    A major cause of failure in nuclear steam generators is tube degradation. Tube defects are divided into seven categories, one of which is intergranular attack/stress corrosion cracking (IGA/SCC). Defects of this type usually begin on the outer surface of the tubes and propagate both inward and laterally. In many cases these defects occur at or near the tube support plates. Several different methods exist for the nondestructive evaluation of nuclear steam generator tubes for defect characterization

  2. MINET validation study using steam generator test data

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  3. 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.

  4. Steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

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

    1995-07-01

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

  5. Steam generator replacement at the Obrigheim nuclear power station

    International Nuclear Information System (INIS)

    Pickel, E.; Schenk, H.; Huemmler, A.

    1984-01-01

    The Obrigheim Nuclear Power Station (KWO) is equipped with a dual-loop pressurized water reactor of 345 MW electric power; it was built by Siemens in the period 1965 to 1968. By the end of 1983, KWO had produced some 35 billion kWh in 109,000 hours of operation. Repeated leaks in the heater tubes of the two steam generators had occurred since 1971. Both steam generators were replaced in the course of the 1983 annual revision. Kraftwerk Union AG (KWU) was commissioned to plant and carry out the replacement work. Despite the leakages the steam generators had been run safely and reliably over a period of 14 years until their replacement. Replacing the steam generators was completed within twelve weeks. In addition to the KWO staff and the supervising crew of KWU, some 400 external fitters were employed on the job at peak work-load periods. For the revision of the whole plant, work on the emergency systems and replacement of the steam generators a maximum number of approx. 900 external fitters were employed in the plant in addition to some 250 members of the plant crew. The exposure dose of the personnel sustained in the course of the steam generator replacement was 690 man-rem, which was clearly below previous estimates. (orig.) [de

  6. Destructive examination and analyses of pulled steam generator tubes

    International Nuclear Information System (INIS)

    Kim, Dong Jin; Kim, H. P.; Kim, J. S.; Lim, Y. S.; Hwang, S. S.; Kim, S. W.; Jeong, M. K.; Hong, J. H.; Kim, W. W.

    2011-07-01

    Steam generator model F in Kori 3, Younggwang 2 and Younggwang 1 as 950MWe PWR was provided by Westinghouse. Steam generator tube made of Alloy 600TT material (outer diameter 0.688'(17.475 mm), thickness 0.04'(1.016mm) in Blairsville was provided by Huntington alloys. Steam generator in Ulchin 4 as 1000MWe KHNP PWR was manufactured by Doosan heavy industry and steam generator tubes were manufactured by B and W (Bobcock and Wilcox). Alloy 600 MA was used as steam generator tubing material of outer diameter 19.05mm and thickness 1.07mm. Five tubes of Alloy 600TT which showed crack signal from non-destructive examination were pulled from Kori 3, Younggwang 2 and Younggwang 1. Two tubes which showed crack signal from non-destructive examination were pulled from Ulchin 4. For the pulled tubes, KAERI performed destructive examination. Through the destructive examination, the existence of cracks were confirmed and the cause of crack was investigated. Remedy was also suggested to mitigate the present circumstances. Stress corrosion cracking (SCC) was observed for Alloy 600TT tubes. It was recommended that the sludge content should be lowered, deleterious elements and MRI should be maintained continuously. For Ulchin 4, SCC was confirmed for Alloy 600MA tubes. It is necessary to lower and remove the sludge in the near term. In the longer term, replacement of steam generator was suggested

  7. Draining down of a nuclear steam generating system

    International Nuclear Information System (INIS)

    Jawor, J.C.

    1987-01-01

    The method is described of draining down contained reactor-coolant water from the inverted vertical U-tubes of a vertical-type steam generator in which the upper, inverted U-shaped ends of the tubes are closed and the lower ends thereof are open. The steam generator is part of a nuclear powered steam generating system wherein the reactor coolant water is normally circulated from and back into the reactor via a loop comprising the steam generator and inlet and outlet conduits connected to the lower end of the steam generator. The method comprises continuously introducing a gas which is inert to the system and which is under pressure above atmospheric pressure into at least one of the downwardly facing open ends of each of the U-tubes from below the tube sheet in which the open ends of the U-tubes are mounted adjacent the lower end of the steam generator, while permitting the water to flow out from the open ends of the U-tubes

  8. Highly Flexible and Efficient Solar Steam Generation Device.

    Science.gov (United States)

    Chen, Chaoji; Li, Yiju; Song, Jianwei; Yang, Zhi; Kuang, Yudi; Hitz, Emily; Jia, Chao; Gong, Amy; Jiang, Feng; Zhu, J Y; Yang, Bao; Xie, Jia; Hu, Liangbing

    2017-08-01

    Solar steam generation with subsequent steam recondensation has been regarded as one of the most promising techniques to utilize the abundant solar energy and sea water or other unpurified water through water purification, desalination, and distillation. Although tremendous efforts have been dedicated to developing high-efficiency solar steam generation devices, challenges remain in terms of the relatively low efficiency, complicated fabrications, high cost, and inability to scale up. Here, inspired by the water transpiration behavior of trees, the use of carbon nanotube (CNT)-modified flexible wood membrane (F-Wood/CNTs) is demonstrated as a flexible, portable, recyclable, and efficient solar steam generation device for low-cost and scalable solar steam generation applications. Benefitting from the unique structural merits of the F-Wood/CNTs membrane-a black CNT-coated hair-like surface with excellent light absorbability, wood matrix with low thermal conductivity, hierarchical micro- and nanochannels for water pumping and escaping, solar steam generation device based on the F-Wood/CNTs membrane demonstrates a high efficiency of 81% at 10 kW cm -2 , representing one of the highest values ever-reported. The nature-inspired design concept in this study is straightforward and easily scalable, representing one of the most promising solutions for renewable and portable solar energy generation and other related phase-change applications. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  9. 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:

  10. Low chemical concentrating steam generating cycle

    Science.gov (United States)

    Mangus, James D.

    1983-01-01

    A steam cycle for a nuclear power plant having two optional modes of operation. A once-through mode of operation uses direct feed of coolant water to an evaporator avoiding excessive chemical concentration buildup. A recirculation mode of operation uses a recirculation loop to direct a portion of flow from the evaporator back through the evaporator to effectively increase evaporator flow.

  11. A fault detection and diagnosis in a PWR steam generator

    International Nuclear Information System (INIS)

    Park, Seung Yub

    1991-01-01

    The purpose of this study is to develop a fault detection and diagnosis scheme that can monitor process fault and instrument fault of a steam generator. The suggested scheme consists of a Kalman filter and two bias estimators. Method of detecting process and instrument fault in a steam generator uses the mean test on the residual sequence of Kalman filter, designed for the unfailed system, to make a fault decision. Once a fault is detected, two bias estimators are driven to estimate the fault and to discriminate process fault and instrument fault. In case of process fault, the fault diagnosis of outlet temperature, feed-water heater and main steam control valve is considered. In instrument fault, the fault diagnosis of steam generator's three instruments is considered. Computer simulation tests show that on-line prompt fault detection and diagnosis can be performed very successfully.(Author)

  12. Steam generator deposit control program assessment at Comanche Peak

    International Nuclear Information System (INIS)

    Stevens, J.; Fellers, B.; Orbon, S.

    2002-01-01

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

  13. Primary separator replacement for Bruce Unit 8 steam generators

    International Nuclear Information System (INIS)

    Roy, S.B.; Mewdell, C.G.; Schneider, W.G.

    2000-01-01

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

  14. Sodium and steam leak simulation studies for fluidized bed steam generators

    International Nuclear Information System (INIS)

    Keeton, A.R.; Vaux, W.G.; Lee, P.K.; Witkowski, R.E.

    1976-01-01

    An experimental program is described which was conducted to study the effects of sodium or steam leaking into an operating fluidized bed of metal or ceramic particles at 680 to 800 0 K. This effort was part of the early development studies for a fluidized-bed steam generator concept using helium as the fluidizing gas. Test results indicated that steam and small sodium leaks had no effect on the quality of fluidization, heat transfer coefficient, temperature distribution, or fluidizing gas pressure drop across the bed. Large sodium leaks, however, immediately upset the operation of the fluidized bed. Both steam and sodium leaks were detected positively and rapidly at an early stage of a leak by instruments specifically selected to accomplish this

  15. Steam generator and condenser design of WWER-1000 type of nuclear power plant

    International Nuclear Information System (INIS)

    Zare Shahneh, Abolghasem.

    1995-03-01

    Design process of steam generator and condenser at Russian nuclear power plant type WWER-1000 is identified. The four chapter of the books are organized as nuclear power plant, types of steam generators specially horizontal steam generator, process of steam generator design and the description of condenser and its process design

  16. The SNR-300 steam generator small leak detection system

    International Nuclear Information System (INIS)

    Dumm, K.

    1984-01-01

    Small leak detection in the SNR-300 steam generator moduls is achieved by hydrogen meters. Development and design of the Nickel membrane - ion getter pump combination are described and sensitivity requests derived. Results of calibration tests by water/steam injections in a sodium loop are presented. The arrangement and interconnection of signals in SNR-300 are given and possibilities for inservice calibrations are discussed, supported by long time operation tests in the KNK-reactor plant. (author)

  17. Apparatus and methods of reheating gas turbine cooling steam and high pressure steam turbine exhaust in a combined cycle power generating system

    Science.gov (United States)

    Tomlinson, Leroy Omar; Smith, Raub Warfield

    2002-01-01

    In a combined cycle system having a multi-pressure heat recovery steam generator, a gas turbine and steam turbine, steam for cooling gas turbine components is supplied from the intermediate pressure section of the heat recovery steam generator supplemented by a portion of the steam exhausting from the HP section of the steam turbine, steam from the gas turbine cooling cycle and the exhaust from the HP section of the steam turbine are combined for flow through a reheat section of the HRSG. The reheated steam is supplied to the IP section inlet of the steam turbine. Thus, where gas turbine cooling steam temperature is lower than optimum, a net improvement in performance is achieved by flowing the cooling steam exhausting from the gas turbine and the exhaust steam from the high pressure section of the steam turbine in series through the reheater of the HRSG for applying steam at optimum temperature to the IP section of the steam turbine.

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

    Energy Technology Data Exchange (ETDEWEB)

    Allen, R.P.; Clark, R.L.; Reece, W.D.

    1984-08-01

    The Steam Generator Group Project utilizes a retired-from-service pressurized-water-reactor steam generator as a test bed and source of specimens for research. An important preparatory step to primary side research activities was reduction of the radiation field in the steam generator channel head. This task report describes the channel head decontamination activities. Though not a programmatic research objective it was judged beneficial to explore the use of dilute reagent chemical decontamination techniques. These techniques presented potential for reduced personnel exposure and reduced secondary radwaste generation, over currently used abrasive blasting techniques. Two techniques with extensive laboratory research and vendors prepared to offer commercial application were tested, one on either side of the channel head. As indicated in the report, both techniques accomplished similar decontamination objectives. Neither technique damaged the generator channel head or tubing materials, as applied. This report provides details of the decontamination operations. Application system and operating conditions are described.

  19. Hydrogen-based power generation from bioethanol steam reforming

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  20. 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.

  1. 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.

  2. Steam generator assessment for sustainable power plant operation

    International Nuclear Information System (INIS)

    Drexler, Andreas; Fandrich, Joerg; Ramminger, Ute; Montaner-Garcia, Violeta

    2012-09-01

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

  3. Method of controlling steam temperature of a fluid heating separation type steam generator

    International Nuclear Information System (INIS)

    Iwashita, Tsuyoshi; Monta, Kazuo.

    1975-01-01

    Object: To keep constant the stability and normal deviation in the entire control system by connecting an element of variable gain substantially in proportion to a preset load in series with the ordinary PID type control system. Structure: Changes in steam temperature at an evaporator outlet due to changes in sodium flow rate are detected by a thermocouple. The resultant detection signal is compared with a preset value of the steam generator output temperature, and a portion proportional to the difference between them is added as an operating signal, the operating signal also being used as a sodium flow rate control signal coupled to a sodium flow rate control means. In this method of control of vapor temperature of a fluid heating separation type steam generator, a control gain variable means is connected in series with a temperature control system to obtain control substantially proportional to the preset load. (Kamimura, M.)

  4. Automation of steam generator services at public service electric & gas

    Energy Technology Data Exchange (ETDEWEB)

    Cruickshank, H.; Wray, J.; Scull, D. [Public Service Electric & Gas, Hancock`s Bridge, NJ (United States)

    1995-03-01

    Public Service Electric & Gas takes an aggressive approach to pursuing new exposure reduction techniques. Evaluation of historic outage exposure shows that over the last eight refueling outages, primary steam generator work has averaged sixty-six (66) person-rem, or, approximately tewenty-five percent (25%) of the general outage exposure at Salem Station. This maintenance evolution represents the largest percentage of exposure for any single activity. Because of this, primary steam generator work represents an excellent opportunity for the development of significant exposure reduction techniques. A study of primary steam generator maintenance activities demonstrated that seventy-five percent (75%) of radiation exposure was due to work activities of the primary steam generator platform, and that development of automated methods for performing these activities was worth pursuing. Existing robotics systems were examined and it was found that a new approach would have to be developed. This resulted in a joint research and development project between Westinghouse and Public Service Electric & Gas to develop an automated system of accomplishing the Health Physics functions on the primary steam generator platform. R.O.M.M.R.S. (Remotely Operated Managed Maintenance Robotics System) was the result of this venture.

  5. 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.).

  6. Implications of Steam Generator Fouling on the Degradation of Material and Thermal Performance

    Science.gov (United States)

    Turner, Carl W.

    Fouling of steam generators has a significant negative impact on the material and thermal performance the steam generators of pressurized water reactors. Corrosion products that originate from various components in the steam cycle of a nuclear power plant get pumped forward with the feed water to steam generators where they deposit on the tube bundle, tube support structure and the tube sheet. Heavy accumulation of deposit within the steam generator has led to some serious operational problems, including loss of thermal performance, under deposit corrosion, steam generator level oscillations, flow accelerated corrosion of carbon steel tube support plates and the failure of steam generator tubes due to high cycle fatigue.

  7. Evaluation of a dryer in a steam generator

    International Nuclear Information System (INIS)

    Xue Yunkui; Liu Shixun; Guandao, Xie; Chen Junliang

    1998-01-01

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

  8. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

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

  9. Thermo-hydraulic stability study of a steam generator

    International Nuclear Information System (INIS)

    Magni, M C; Marcel, C P; Delmastro, D F

    2012-01-01

    In this work a mathematical model developed to investigate the thermalhydraulic stability of a helically coiled steam generator is presented. Such a steam generator is prone to experiment density wave oscillations. The model is therefore used to analyze the stability of the CAREM-25 reactor steam generators. The model is linear, numerically non-diffusive and nodal. In addition, it is able to represent non-uniform heat transfer fluxes between the primary and secondary coolant circuits. By using this model the marginal stability condition is found by varying the inlet friction coefficient for different conditions. The results are then compared with those obtained with a different model for which a simple uniform heat flux profiled is assumed. It is found that with such simplification the density waves instability mechanism is overestimated in a wide range of operating powers. For very low powers, in the contrary, the so-called uniform model underestimates the stabilizing inlet friction and therefore it gives non-conservative results. With the use of the more realistic non-uniform power profile model, it was possible to determine that, for a CAREM-25 steam generator, the most stable conditions is found at 60MW when the reactor operates at nominal pressure. Moreover, it is found that at high power levels the stability performance is dominated by the two-phase friction component while at low power levels the friction component originated in the over heated steam region prevail (author)

  10. Water leak detection in steam generator of SUPER PHENIX

    International Nuclear Information System (INIS)

    Brunet, M.; Garnaud, P.; Ghaleb, D.; Kong, N.

    1988-01-01

    With the intent of detecting water leaks inside steam generators, we developed a third system, called acoustic detector, to complement hydrogen detectors and rupture disks (burst disks). The role of the acoustic system is to enable rapid intervention in the event of a leak growing rapidly which could rupture neighbouring tubes. In such a case, the detectable flow rate of the leak varies from a few tens of g/s to a few hundred g/s. At the SUPER PHENIX, three teams work in [20-100 kHz] and CEA/STA* [50-300 kHz]. The simulation of water leaks in the steam generator by the argon injections performed to date at 50% of the rated power has shown promising results. An anomaly in the evolution of the background noise at more than 50% loading of one of the two instrumented steam generators would make difficult any extrapolation to full power behaviour. (author)

  11. The steam generators components inspection programme of PISC III

    International Nuclear Information System (INIS)

    Birac, C.; Herkenrath, H.

    1988-07-01

    As a result of an enquiry conducted in 1986 and 1987 the extention of the PISC methodology has been decided for the Steam Generators Components Testing techniques evaluation. Besides calibration tubes and training tubes to be circulated amongst the participants, three phases are organized: capability tests on loose tubes, capability tests on a transportable mock-up, reliability tests on fixed mock-ups one of which has to contain interesting portions of the Surry Steam Generator. It is hoped to assess the performance and to identify clear trends for what concerns defects detection location and characterization in steam generators tubing as it has been obtained by PISC II in pressure vessel steel structures

  12. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  13. ABB Combustion Engineering approach to steam generator sludge lancing

    Energy Technology Data Exchange (ETDEWEB)

    Estes, G.T. [ABB CE, Chattanooga, TN (United States)

    1994-12-31

    During the normal operation of a recirculating steam generator, insoluble corrosion products, commonly called {open_quotes}sludge,{close_quotes} are formed. Sludge accumulates on the tube sheet, heat transfer tubes, support plates, and other surfaces. It can be expected that corrosion products will continue to build up even with rigid secondary water chemistry controls. This corrosion product buildup directly affects thermal efficiency, resulting in megawatt and pressure loss. Secondary-side corrosion causes an increase in tubes that require plugging or sleeving. More importantly, corrosion product buildup has caused forced outages and steam-generator replacements. One plant has permanently shut down due to steam-generator secondary-side corrosion. This report describes a process for sludge removal.

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

    International Nuclear Information System (INIS)

    1975-07-01

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

  15. Functional performance of the helical coil steam generator, Consolidated Nuclear Steam Generator (CNSG) IV system. Executive summary report

    International Nuclear Information System (INIS)

    Watson, G.B.

    1975-10-01

    The objective of this project was to study the functional performance of the CNSG - IV helical steam generator to demonstrate that the generator meets steady-state and transient thermal-hydraulic performance specifications and that secondary flow instability will not be a problem. Economic success of the CNSG concepts depends to a great extent on minimizing the size of the steam generator and the reactor vessel for ship installation. Also, for marine application the system must meet stringent specifications for operating stability, transient response, and control. The full-size two-tube experimental unit differed from the CNSG only in the number of tubes and the mode of primary flow. In general, the functional performance test demonstrated that the helical steam generator concept will exceed the specified superheat of 35F at 100% load. The experimental measured superheat at comparable operating conditions was 95F. Testing also revealed that available computer codes accurately predict trends and overall performance characteristics

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

    International Nuclear Information System (INIS)

    Sudardjo; Histori; Triyadi, Ari

    1998-01-01

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

  17. Overview of steam generator tube degradation and integrity issues

    International Nuclear Information System (INIS)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

    Kumar, Kundan; Nathani, D.K.; Kayal, J.N.; Rupani, B.B.

    2006-01-01

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

  19. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

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

    International Nuclear Information System (INIS)

    Chetal, S.C.; Vaidyanathan

    1997-01-01

    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

  1. The impact of maintenance strategy on steam generator reliability

    International Nuclear Information System (INIS)

    Cizelj, L.; Mavko, B.

    1992-01-01

    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) [sl

  2. 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

  3. Chemistry and corrosion on steam generators in PWRs

    International Nuclear Information System (INIS)

    Berge, J.P.; Nordmann, F.

    1989-01-01

    After a review of the objectives of primary coolant chemistry, the reasons are given for the 'decaying lithium' specification and for its development to decrease dose rates while avoiding increasing the risks of primary side cracking of steam generator pipes. For conditioning secondary coolant, the choice of volatile conditioning (ammoniac or morpholine) and its characteristics are specified. The different types of corrosion of steam generators are discussed, particularly cracking under stress corrosion on the primary side and intergranular attack of the pipes on the secondary side; the associated remedies and consequences are also discussed. 8 figs., 3 tabs., 5 refs

  4. CANDU 6 steam generator thermalhydraulic modeling and simulation

    International Nuclear Information System (INIS)

    Zahedi, P.; Borairi, M.

    2009-01-01

    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

  5. Electropolishing process development for PWR steam generator channel heads

    International Nuclear Information System (INIS)

    Asay, R.H.; Graves, P.; Guastaferro, C.T.; Spalaris, C.N.

    1991-04-01

    A broad range of process parameters was established to smoothen the surface of 309 L weld clad overlay, prototypic of surfaces common is channel heads of replacement PWR [pressurized water reactor] steam generators. Mechanical and electropolishing steps were studied to explore process boundaries, which result in acceptable degree of surface smoothness, without compromising metallurgical properties. Recommended processes and acceptance criteria established in this work, can be applied to electropolish steam generator channel heads. Smooth surfaces are less likely to retain radioactive species, and potentially develop lower radiation fields when these components are placed into service. 7 refs., 11 figs., 12 tabs

  6. Chemical cleaning of PWR steam generators: application at Nogent 1

    International Nuclear Information System (INIS)

    Fiquet, J.M.; Veysset, J.P.; Esteban, L.; Saurin, P.

    1990-01-01

    EDF has developed and patented a chemical cleaning process for PWR steam generators, based on the use of a mixture of organic acids in order to: - dissolve iron oxides and copper with a single solution; - clean dented crevices. Qualification tests have permitted to demonstrate effectiveness of the solution and its inocuousness related to steam generator materials. The process, the license of which belongs to SOMAFER R.A. 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 [fr

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

    International Nuclear Information System (INIS)

    1980-01-01

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

  8. Leak location in a double wall tube steam generator. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Greene, D.A.; McMurtrie, K.R.; Gaubatz, D.C.

    1980-01-01

    An experimental and analytical study was made of an acoustic location system for use with double wall tube steam generators. A conceptual design of an acoustic leak location system was developed. This system has the potential for detecting the third fluid gas escaping from the annulus between the double walled tubes into the sodium. The acoustic system uses an array of accelerometers mounted onto the outside of the steam generator vessel. Electronic circuitry selects a number of accelerometers, forms them into an array which focuses within the vessel. The volume of the vessel is sequentially scanned for anomalous acoustic energy.

  9. Slurry steam generator program and baseline eddy current examination

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

    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. 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.)

  12. A balanced strategy in managing steam generator thermal performance

    International Nuclear Information System (INIS)

    Hu, M. H.; Nelson, P. R.

    2009-01-01

    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

  13. Tube tightness survey during Phenix steam generator operation

    International Nuclear Information System (INIS)

    Cambillard, E.

    1976-01-01

    Phenix steam generators are once-through vessels with single-wall heat-exchange tubes. This design means that any leakage of water into the sodium must be detected as quickly as possible so that the installation can be shut down before extensive damage occurs. The detection of water leaks in Phenix steam generators is based on measurement of the concentration in the sodium, of hydrogen produced by the sodium-water reaction. Since the various modules--evaporators, superheaters, and reheaters--have no free sodium surfaces, detection of hydrogen in argon is not used in Phenix steam generators. The measurement systems employ a probe made of nickel tubes 0.3 mm thick. Hydrogen in the sodium diffuses into a chamber kept under vacuum by an ion pump. The hydrogen pressure in the chamber is measured by a quadrupole mass spectrometer. The nine measurement systems (three per steam generator) are calibrated by injecting hydrogen into the sodium of the secondary circuits. The data-processing computer calculates the hydrogen concentration in the sodium from the spectrometer signals and the probe temperatures, which are not regulated in Phenix; it generates instructions that enable the operator to act if a leak appears. So far, no leaks have been detected. These systems also make it possible to determine rates of hydrogen diffusion caused by corrosion of the steel walls on the water side

  14. 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)

    Cambillard, E.; Lacroix, A.; Langlois, J.; Viala, J.

    1975-01-01

    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)

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

    International Nuclear Information System (INIS)

    Pappx, L.

    1994-01-01

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

  16. Leak suppression at steam generator man-, hand-, and eyeholes

    International Nuclear Information System (INIS)

    Sylvain, C.; Sutz, P.; Gemma, A.

    1988-01-01

    Plant unavailability associated with primary and secondary holes is approximately the same as that caused by steam generator tube defects, i.e., 0.5%. Problems encountered with steam generator man-, hand-, and eyeholes during plant operation have led Electricite de France (EdF) and Framatome to improve hole seal design and to develop robots for closing and cleaning them. The data base available in France in this field on some 150 steam generators in 900- and 1300-MW(electric) pressurized water reactors (the equivalent of 300 reactor-yr of operation) has been the base of the developments described in this paper. Incidents occurring in operation primarily concern had-and inspection holes located on the steam generator's secondary side. They include four kinds: (1) leakage detected in operation, requiring forced outages, (2) leakage detected during plant restart after a scheduled shutdown and resulting in a restart delay, (3) pitting of seal mating surfaces, not inducing any leakage but jeopardizing subsequent compliance and requiring difficult and costly repairs, and (4) seizing of screws or bolts. New primary and secondary hole stud tightening and maintenance machines help to improve the efficiency of the in-service closing operations. They provide savings of up to 80% on labor, duration of operations, and exposure

  17. Leak detection in LMFBR steam generators during operation

    International Nuclear Information System (INIS)

    Dumm, K.

    1978-01-01

    This paper deals with the following four main aspects: requirement on the leak detection of the SNR-300 steam generators; the hydrogen detector of SNR-300; remarks on the disadvantage of impurity detectors; and the first approach to acoustic leak detection systems

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

    International Nuclear Information System (INIS)

    Oliveira, Magno Jose de; Amaral, Marcos Antonio do; Minelli, Edson; Ferreira, William Alves

    2009-01-01

    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)

  19. Unique systems for inspection and repair of VVER steam generators

    International Nuclear Information System (INIS)

    Kovalyk, Adrian; Pilat, Peter; Jablonicky, Pavol

    2014-01-01

    During over 30 years, VUJE Trnava has developed a series of remote control manipulators for in-service inspection and maintenance of steam generator collectors. The manipulators have been widely used on WWER type reactors in Slovakia and abroad. Some of the manipulators for non-destructive testing are shown.

  20. Experimental fretting-wear studies of steam generator materials

    International Nuclear Information System (INIS)

    Fisher, N.J.; Chow, A.B.; Weckwerth, M.K.

    1994-01-01

    Flow-induced vibration of steam generator tubes results in fretting-wear damage due to impacting and rubbing of the tubes against their supports. This damage can be predicted by computing tube response to flow-induced excitation forces using analytical techniques, and then relating this response to resultant wear damage using experimentally-derived wear coefficients. Fretting-wear of steam generator materials has been studied experimentally at Chalk River Laboratories for two decades. Tests are conducted in machines that simulate steam generator environmental conditions and tube-to-support dynamic interactions. Different tube and support materials, tube-to-support clearances and tube support geometries have been studied. As well, the effect of environmental conditions, such as temperature, oxygen content, pH and chemistry control additive, have been investigated. Early studies showed that damage was related to contact force as long as other parameters, such as geometry and motion were held constant. Later studies have shown that damage is related to a parameter called work-rate, which combines both contact force and sliding distance. Results of short- and long-term fretting-wear tests for CANDU steam generator materials at realistic environmental conditions are presented. These results demonstrate that work-rate is appropriate correlating parameter for impact-sliding interaction

  1. Steam generators regulatory practices and issues in Spain

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  2. Numerical analysis of thermal hydraulics in secondary side steam generator

    International Nuclear Information System (INIS)

    Yang Zhilin

    1997-01-01

    The author presents the analysis on the possibility for COBRA-TF to model thermal-hydraulics of SG (Steam Generator) secondary side, numerically analyzes the thermal-hydraulic processes of heater section of SG secondary side of Qinshan Nuclear Power Plant, the results are compared with those from ALBERTINE-2

  3. Work plan for steam generator leak detector tests in CCTL

    International Nuclear Information System (INIS)

    1973-01-01

    This work plan outlines a test program on devices for the detection of water to sodium leaks in an LMFBR steam generator, to be carried out in the Core Component Test Loop (CCTL) at the Argonne National Laboratory. The objective of these tests is to develop the data on the performance characteristics of leak detectors which will be applicable to the design of reliable leak detection systems for large scale LMFBR plants, with particular attention to the needs and requirements of the Demonstration Plants. In doing this, water will be injected into sodium in the CCTL in controlled quantities and at rates which it is considered necessary to detect in LMFBR steam generators. The experiments will be planned to establish the performance and limitations of the leak detection devices, and in order to furnish data applicable to the design of the leak detection systems of the LMFBR Steam Generators, will include an analysis to determine the scaling factors needed to translate the results of the CCTL tests to meaningful predictions on leak detection system behavior in large steam generators

  4. Steam generation by combustion of processed waste fats

    Energy Technology Data Exchange (ETDEWEB)

    Pudel, F.; Lengenfeld, P. [OEHMI Forschung und Ingenieurtechnik GmbH, Magdeburg (Germany)

    1993-12-31

    The use of specially processed waste fats as a fuel oil substitute offers, at attractive costs, an environmentally friendly alternative to conventional disposal like refuse incineration or deposition. For that purpose the processed fat is mixed with EL fuel oil and burned in a standard steam generation plant equipped with special accessories. The measured emission values of the combustion processes are very low.

  5. Full bundle mechanical cleaning of pressurized water reactor steam generators

    International Nuclear Information System (INIS)

    Reinhardt, M.; Lewis, G.

    1998-01-01

    An improved system for cleaning the upper bundle region of pressurized water reactor steam generators is now available. This system employs high pressure water jets to remove sludge and scale from tube and support plate surfaces and from the broach holes throughout the upper bundle region of steam generators. Benefits of sludge removal can include mitigation of the corrosion process, improvement of plant performance and increased operational stability. The cleaning equipment is installed through existing handholes in the tubesheet region and accesses the upper bundle through flow slots in the tube support plates. The system includes a Bulk Cleaning Head which directs high pressure water jets into the tube bundle, a Vertical Deployment System which delivers and positions the head at the desired locations in the upper bundle region, and a high pressure water pumping and sludge processing system. This paper describes the system, several key components, overall system operation, safety issues, and field use chronology. Specific performance issues that are discussed include the cleaning coverage achieved, the time required to clean a steam generator, and cleaning effectiveness. The Upper Bundle Hydraulic Cleaning System described in this paper is available for use in several types of steam generators

  6. Mechanical cleaning of steam generators in PWR nuclear power plant

    International Nuclear Information System (INIS)

    Ding Xunshen; Li Sujia

    1994-01-01

    The mechanical cleaning of steam generators and its equipment currently employed internationally and in Daya Bay Nuclear Power Plant are described. The site mechanical cleaning practices for nuclear power plant are also introduced. A number of remarks on mechanical cleaning, chemical cleaning, TV inspection of cleanness and withdrawal of foreign materials are given

  7. Mushrooms as Efficient Solar Steam-Generation Devices.

    Science.gov (United States)

    Xu, Ning; Hu, Xiaozhen; Xu, Weichao; Li, Xiuqiang; Zhou, Lin; Zhu, Shining; Zhu, Jia

    2017-07-01

    Solar steam generation is emerging as a promising technology, for its potential in harvesting solar energy for various applications such as desalination and sterilization. Recent studies have reported a variety of artificial structures that are designed and fabricated to improve energy conversion efficiencies by enhancing solar absorption, heat localization, water supply, and vapor transportation. Mushrooms, as a kind of living organism, are surprisingly found to be efficient solar steam-generation devices for the first time. Natural and carbonized mushrooms can achieve ≈62% and ≈78% conversion efficiencies under 1 sun illumination, respectively. It is found that this capability of high solar steam generation is attributed to the unique natural structure of mushroom, umbrella-shaped black pileus, porous context, and fibrous stipe with a small cross section. These features not only provide efficient light absorption, water supply, and vapor escape, but also suppress three components of heat losses at the same time. These findings not only reveal the hidden talent of mushrooms as low-cost materials for solar steam generation, but also provide inspiration for the future development of high-performance solar thermal conversion devices. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  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. Hydrogen generation utilizing integrated CO2 removal with steam reforming

    Science.gov (United States)

    Duraiswamy, Kandaswamy; Chellappa, Anand S

    2013-07-23

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

  10. Steam generator performance improvements for integral small modular reactors

    OpenAIRE

    Muhammad Ilyas; Fatih Aydogan

    2017-01-01

    Background: Steam generator (SG) is one of the significant components in the nuclear steam supply system. A variety of SGs have been designed and used in nuclear reactor systems. Every SG has advantages and disadvantages. A brief account of some of the existing SG designs is presented in this study. A high surface to volume ratio of a SG is required in small modular reactors to occupy the least space. In this paper, performance improvement for SGs of integral small modular reactor is proposed...

  11. Environmental codes of practice for steam electric power generation

    International Nuclear Information System (INIS)

    1985-03-01

    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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  13. 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.

  14. 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.

  15. Downhole steam generator having a downhole oxidant compressor

    Science.gov (United States)

    Fox, Ronald L.

    1983-01-01

    Apparatus and method for generation of steam in a borehole for penetration into an earth formation wherein a downhole oxidant compressor is used to compress relatively low pressure (atmospheric) oxidant, such as air, to a relatively high pressure prior to mixing with fuel for combustion. The multi-stage compressor receives motive power through a shaft driven by a gas turbine powered by the hot expanding combustion gases. The main flow of compressed oxidant passes through a velocity increasing nozzle formed by a reduced central section of the compressor housing. An oxidant bypass feedpipe leading to peripheral oxidant injection nozzles of the combustion chamber are also provided. The downhole compressor allows effective steam generation in deep wells without need for high pressure surface compressors. Feedback preheater means are provided for preheating fuel in a preheat chamber. Preheating of the water occurs in both a water feed line running from aboveground and in a countercurrent water flow channel surrounding the combustor assembly. The countercurrent water flow channels advantageously serve to cool the combustion chamber wall. The water is injected through slotted inlets along the combustion chamber wall to provide an unstable boundary layer and stripping of the water from the wall for efficient steam generation. Pressure responsive doors are provided at the steam outlet for closing and sealing the combustion chamber from entry of reservoir fluids in the event of a flameout.

  16. Conceptual design study of Cu bonded steam generator

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Konomura, Mamoru

    2004-05-01

    In phase II of feasibility study of commercialized fast reactor cycle systems of JNC, we make a concept of a sodium cooled reactor without secondary sodium circuits. And a sodium cooled reactor with Cu bonded steam generators is one of promising concept. As the result of FY 2001 study, the construction cost of reactor cooling system with rectangular tube Cu bonded steam generators is 0.71 to 1.23 times as much as that of an ordinary sodium cooled reactor with secondary sodium circuits. In the FY 2003 study, plastic and creep analysis to evaluate life distortion are carried out and inelastic strains and creep fatigue damage are checked for full code compliance. The NNC's crack growth experiments show that there are few possibility to penetrate a crack from the steam tube side to the sodium tube side at the operating temperature. But penetration is observed in a four point bend test at the room temperature, because the notch opens widely in the bend test. In the FY 2004 study, a gas pressurized crack growth experiment is planed to confirm that there is no crack penetration in the condition of operating steam generators. (author)

  17. Qualification of Alloy 800 for sodium heated steam generators

    International Nuclear Information System (INIS)

    Duke, J.M.; Sessions, C.E.; Ray, W.E.

    1976-01-01

    A reference specification of Alloy 800 for use in LMFBR steam generators is defined considering waterside corrosion, weldability and mechanical properties. Additional mechanical test data are being generated to support ASME Code acceptance. Candidate weld filler metals were assessed for use in subsequent weld process development. Progress on resolving technical concerns related to the role of tertiary creep identified the impact of test conditions and gamma prime strengthening in determining the creep behavior of Alloy 800

  18. Steam generator secondary side chemical cleaning at Gentilly-2

    International Nuclear Information System (INIS)

    Plante, S.

    2006-01-01

    After more than 20 years of operation, the secondary side of the four steam generators at Gentilly-2 were chemically cleaned during the 2005 annual outage. The FRAMATOME ANP high temperature cleaning process used to remove magnetite loading involved stepwise injection of solvent with PHT temperature in the range 160 o C to 175 o C. The heat required to maintain the PHT temperature was provided by the operation of the main PHT pumps and the reactor core residual heat. The temperature control was accomplished by the shutdown cooling system heat exchangers. A total of 1280 kg of magnetite was removed from the four steam generators. A copper-cleaning step was applied after the iron step. The PHT has been cooled down and the steam generators drained to temporary tanks and dried in preparation of the copper step. The process has been applied at room temperature, two boilers at a time. The solvent removed a total of 116 kg of copper. During the iron step, steam flow to the feedwater tank chemically contaminate the Balance Of Plant (BOP) systems. The isolation of this path should have been part of the G2 procedures. Around 700 m3 of water had to be drained to interim storage tanks for subsequent resin treatment before disposal. Visual inspection of BO1 tubesheet and first support plate showed clean surfaces without measurable sludge pile. Upper support plates visual inspection of BO4 revealed that broach holes blockage reported in 2000 is still present in peripheral area. Following the plant restart, the medium range level measurement instability observed since several years for BO3 was no more present. As anticipated, it also has been observed that the medium and wide range level measurements have shifted down as a result of downcomer flow increase after the cleaning. The cleaning objectives were achieved regarding the fouling reduction on the steam generators secondary side but broach holes blockage of the upper support plate is still present in periphery. (author)

  19. Tests and analysis on steam generator tube failure propagation

    International Nuclear Information System (INIS)

    Tanabe, Hiromi

    1990-01-01

    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

  20. Recent technology for nuclear steam turbine-generator units

    International Nuclear Information System (INIS)

    Moriya, Shin-ichi; Kuwashima, Hidesumi; Ueno, Takeshi; Ooi, Masao

    1988-01-01

    As the next nuclear power plants subsequent to the present 1,100 MWe plants, the technical development of ABWRs was completed, and the plan for constructing the actual plants is advanced. As for the steam turbine and generator facilities of 1,350 MWe output applied to these plants, the TC6F-52 type steam turbines using 52 in long blades, moisture separation heaters, butterfly type intermediate valves, feed heater drain pumping-up system and other new technologies for increasing the capacity and improving the thermal efficiency were adopted. In this paper, the outline of the main technologies of those and the state of examination when those are applied to the actual plants are described. As to the technical fields of the steam turbine system for ABWRs, the improvement of the total technologies of the plants was promoted, aiming at the good economical efficiency, reliability and thermal efficiency of the whole facilities, not only the main turbines. The basic specification of the steam turbine facilities for 50 Hz ABWR plants and the main new technologies applied to the turbines are shown. The development of 52 in long last stage blades, the development of the analysis program for the coupled vibration of the large rotor system, the development of moisture separation heaters, the turbine control system, condensate and feed water system, and the generators are described. (Kako, I.)

  1. Mechanical design of the hot steam headers of the THTR-300 steam generators

    International Nuclear Information System (INIS)

    Blumer, U.; Stumpf, M.

    1988-01-01

    The high pressure steam headers of the THTR steam generators have been subject to special attention during the design phase due to the following reasons: these components are the pressure retaining parts with the heaviest wall thickness in the region of the steam generators; they therefore are sensitive to thermal transient conditions; they are operated in the elevated temperature regime, where creep effects cannot be neglected; there is almost no service experience from fossil steam generators with this type of material (Alloy 800). Safety consideration therefore have been rather extensive and have focussed on two main areas which will be treated in this paper: 1. Analytical investigations on the cyclic material behaviour under all specified operating conditions, taking into account the non-elastic response of the material. 2. Limitation of the consequences of a header rupture by installation of heavy whip restraints. Elastic-plastic-creep analyses: The analyses were performed in different stages and are explained in the corresponding order: Evaluation of the critical location on the header and establishment of a simplified model of a nozzle region for further analysis. Preliminary thermal analyses of all specified transient conditions on simplified procedures, in order to establish a severity ranking of the conditions. Establishment of representative loading blocks. Evaluation of the material properties for thermal and structural, especially non-elastic behaviour. Detailed thermal analyses. Detailed structural analyses of the non-elastic cyclic response. Extrapolation for all cycles and assessment of the results by design codes. Discussion of the results. Header whip restraint design: In addition to the above analysis efforts, heavy whip restraints were provided to assure limitation of the effects of a header failure. This pager shows the measures that were taken to restrain the movement in case of longitudinal and transverse breaks: The anti-whip designs are

  2. Characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation

    Energy Technology Data Exchange (ETDEWEB)

    Pagan, Sandra [Ontario Power Generation, 889 Brock Road, Pickering, Ontario (Canada); Duan Xinjian [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario (Canada)], E-mail: duanx@aecl.ca; Kozluk, Michael J. [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario (Canada); Mills, Brian; Goszczynski, Guylaine [Kinectrics Inc., 800 Kipling Avenue, Toronto, Ontario (Canada)

    2009-03-15

    The Canadian Nuclear Standard CSA N285.4 requires the periodic metallurgical examination of removed ex-service steam generator tubes. This paper describes the practices used for the characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation (OPG). It shows that there is no degradation of mechanical properties of Monel 400 tubes after 7-18 effective full power years (EFPY) of operation and Incoloy 800 tubes after more than 10 EFPY of operation.

  3. Glas generator for the steam gasification of coal with nuclear generated heat

    International Nuclear Information System (INIS)

    Buchner, G.

    1980-01-01

    The use of heat from a High Temperature Reactor (HTR) for the steam gasification of coal saves coal, which otherwise is burnt to generate the necessary reaction heat. The gas generator for this process, a horizontal pressure vessel, contains a fluidized bed of coal and steam. An ''immersion-heater'' type of heat exchanger introduces the nuclear generated heat to the process. Some special design problems of this gasifier are presented. Reference is made to the present state of development of the steam gasification process with heat from high temperature reactors. (author)

  4. 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.

  5. Electric power generating plant having direct coupled steam and compressed air cycles

    Science.gov (United States)

    Drost, Monte K.

    1982-01-01

    An electric power generating plant is provided with a Compressed Air Energy Storage (CAES) system which is directly coupled to the steam cycle of the generating plant. The CAES system is charged by the steam boiler during off peak hours, and drives a separate generator during peak load hours. The steam boiler load is thereby levelized throughout an operating day.

  6. Electric power generating plant having direct-coupled steam and compressed-air cycles

    Science.gov (United States)

    Drost, M.K.

    1981-01-07

    An electric power generating plant is provided with a Compressed Air Energy Storage (CAES) system which is directly coupled to the steam cycle of the generating plant. The CAES system is charged by the steam boiler during off peak hours, and drives a separate generator during peak load hours. The steam boiler load is thereby levelized throughout an operating day.

  7. 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)

    Park, Chi-Yong; Ryu, Ki-Wahn; Rhee, Huinam

    2013-01-01

    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

  8. 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)

  9. Disposal and handling of nuclear steam generator chemical cleaning wastes

    International Nuclear Information System (INIS)

    Larrick, A.P.; Schneidmiller, D.

    1978-01-01

    A large number of pressurized water nuclear reactor electrical generating plants have experienced a corrosion-related problem with their steam generators known as denting. Denting is a mechanical deformation of the steam generator tubes that occurs at the tube support plates. Corrosion of the tube support plates occurs within the annuli through which the tubes pass and the resulting corrosion oxides, which are larger in volume than the original metal, compress and deform the tubes. In some cases, the induced stresses have been severe enough to cause tube and/or support cracking. The problem was so severe at the Turkey Point and Surrey plants that the tubing is being replaced. For less severe cases, chemical cleaning of the oxides, and other materials which deposit in the annuli from the water, is being considered. A Department of Energy-sponsored program was conducted by Consolidated Edison Co. of New York which identified several suitable cleaning solvents and led to in-plant chemical cleaning pilot demonstrations in the Indian Point Unit 1 steam generators. Current programs to improve the technology are being conducted by the Electric Power Research Institute, and the three PWR NSSS vendors with the assistance of numerous consultants, vendors, and laboratories. These programs are expected to result in more effective, less corrosive solvents. However, after a chemical cleaning is conducted, a large problem still remains- that of disposing of the spent wastes. The paper summarizes some of the methods currently available for handling and disposal of the wastes

  10. The steam generating heavy water reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

    A review is presented on the evolution of the SGHWR concept by the United Kingdom Atomic Energy Authority and the production of early commercial designs, together with later development by the Design and Construction Companies. This is followed by a description of the current commercial design. Possible future developments are suggested. The many advantageous features of the concept are mentioned with a view to supporting optimism for the future of the system. Headings include the following: safety criteria and risk assessment; emergency core cooling system design and development; protective systems; reactor coolant system; reactivity control; off-load refuelling; pressure containment; 'fence' header coolant circuit design; feed water injection; continuous spray cooling; low pressure cooling systems for residual heat removal during refuelling; high pressure cooling system for guaranteed feed water supply; auxiliary systems; structural materials; calandria and neutron shields; fuel element development; alternative loop circuit design; future developments (use of hydraulic diodes to provide a substantial reverse flow resistance by the generation of a vortex; multi-drum and multi-pump schemes; refuelling alternatives; coolant circuit inversion; use of superheat channels). (U.K.)

  11. Vertical steam generators for VVER: myths and reality

    International Nuclear Information System (INIS)

    Trunov, N.

    2009-01-01

    The paper covers two basic concepts of steam generators, horizontal and vertical. It gives references to the analysis of 40-year operating experience and a short description of recent and future designs of horizontal SGs. The main design features and efficiency of the two SG concepts are compared including a comparative calculation analysis of heat efficiency. A short review of alternative designs of vertical steam generators (SG) for VVER of Russian design is presented with an analysis of the possible benefits of implementing vertical SGs in VVER reactor plants. The paper also covers an analysis of the effect of SG type on the layout of reactor plant building, taking into account the transportation and replacement technologies. The concept of VVER-1200A advanced reactor plant taken as an example corroborates lack of arguments in favor of vertical SGs implementation. (author)

  12. Leak detection and location in MONJU steam generators

    International Nuclear Information System (INIS)

    Saito, T.; Kosugi, T.

    1978-01-01

    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)

  13. Acoustic tool for leak inspections in steam generator divider plates

    International Nuclear Information System (INIS)

    Kowalewski, J.; Maak, P.; Colbert, M.

    1997-01-01

    Cross flow leakage through divider plates in the primary head of steam generators, may be a contributor to RIHT rise which can have a negative effect on the operation of a plant. A method to provide quick and reliable inspection of divider plate leakage can be very useful in helping make timely and effective maintenance decisions. A novel acoustic tool for performing inspections in drained steam generators during shutdowns has been developed by OH Technologies and successfully demonstrated in a field application. The technique, referred as ALIS (acoustic leak inspection system), performs a scan of the divider plate face and presents a graphic image of the leakage paths and an estimate of the total leakage area. (author)

  14. Methods for preventing steam generator failure or degradation

    International Nuclear Information System (INIS)

    Green, S.J.

    1986-01-01

    PWR steam generators have suffered from a variety of degradation phenomena. This paper identifies the corrosion-related defects and their probable causes and suggests approaches to correct and prevent corrosive activity. In the attempt to solve the degradation problems, research programs have concentrated on modifying materials, stresses, and the chemical environment in both new and operating steam generators. The following corrosion-related defects have been studied: tube wastage, denting, primary side (ID) intergranular stress corrosion cracking (IGSCC), OD-initiated intergranular attack (IGA), pitting, and corrosion fatigue. Plants affected by wastage have greatly reduced their problem by adopting an all volatile treatment (AVT). In the case of denting, a less aggressive chemical environment is recommended. Primary side IGSCC responds to temperature reduction, stress relief, and material improvements, while flushing and boric acid addition minimizes OD-initiated IGA. It has further been shown that pitting can be minimized by sludge lancing and by reducing impurity ingress. (author)

  15. Homogenized behaviour of the steam generator perforated plates

    International Nuclear Information System (INIS)

    Voldoire, F.

    1993-06-01

    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

  16. Detection and localisation of leaks in steam generators

    International Nuclear Information System (INIS)

    Westenbrugge, J.K. van; Visbach, D.M.

    1978-01-01

    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)

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

    International Nuclear Information System (INIS)

    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

  18. 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.

  19. Corrosion in steam generators of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Hyspecka, L.; Tvrdy, M.

    1988-01-01

    Problems are discussed of heat exchange tubes of Westinghouse type vertical steam generators exhibiting corrosion damage such as point corrosion, planar corrosion, tube denting, corrosion stress cracking, crevice corrosion, fretting corrosion and intergranular corrosion. Attention is also paid to problems of WWER-440 type horizontal steam generators, where the level fluctuation area is critical; noncompact porous deposits of the corrosion products give rise to crevice effects and cause significant concentration of chloride ions and other additions. This problem can be partly resolved by a modification of the collector design at the level variation area. An additional measure is the production of steel 08Kh18N10T with a very low level of harmful elements and inclusions. (Z.M.). 3 figs., 11 refs

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

    International Nuclear Information System (INIS)

    Pigeon, M.; David, B.

    1983-06-01

    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 [fr

  1. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  2. Probe for detection of denting in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gerardin, J.P.; Germain, J.L.; Nio, J.C.

    1994-07-01

    In certain types of PWR steam generator, oxide deposits can lead to embedding, and subsequently to deformation of a tube (the phenomenon of ''denting''). Such embedding changes the vibratory behavior of the tubes and can result in fatigue cracking. This type of cracking can also be worsened in the event of improper assembly of the anti-vibration spacer bars supporting the U-bends. To prevent such incidents and provide for effective preventive condition-directed maintenance of its PWR steam generators, EDF has undertaken the study and development of a probe to detect this type of phenomenon. The studies began in 1990 and led to the building of an initial prototype probe. The principle behind the probe consists in inducing vibration in the U-bend and determining the main resonance modes of the tube. Measurements of frequency and amplitude and calculation of damping enable characterization of the mechanical behavior of the U-bend. The most important parameter is damping, for which the value must be sufficiently high to ensure that the tube is not subjected to major vibratory amplitudes during operation. Numerous tests have been performed with the first prototype version of the probe, on a mock-up in the test area and on one of the demounted steam generators on the Dampierre site. These different tests have enabled validation of the operating principle, fine-tuning the process, pinpointing certain mechanical problems in the probe design, and obtaining the first indications as to the real vibratory behavior of U-bends on a steam generator. On the basis of these preliminary tests, the specifications were drawn up for an industrial version of the probe. Following a call for bids and the choice of a manufacturer, work began on fabrication of a new probe model in 1993. This version was delivered at the end of 1993 and testing began in 1994. (authors). 5 figs., 2 tabs

  3. Linear stability model for the once-through steam generator

    International Nuclear Information System (INIS)

    Kim, Y. S.; Kim, O. K.; Wi, M. H.; A, J. H.

    2001-01-01

    A linear stability model was developed to evaluate the stability of the once-through steam generator. Using the perturbation and Laplace transformation the governing equations were transformed to perturbed linear equations whose solutions are used to calculate pertubed pressure drop relations. An applicable algorithm for numercial analysis was adopted and draft version of a computer code named SGSTAR was developed. Using the SGSTAR a test calculation on the density wave instability has been conducted

  4. Development of stabilizers for steam generator tube repair

    International Nuclear Information System (INIS)

    Au-Yang, M.K.

    1987-01-01

    Fluidelastic stability, turbulence-induced and vortex-induced vibration analysis of different types of stabilizers for repairing steam generator tubes are presented. The performances of the different designs are compared with that of a common basis - the virgin tube. It was found that in addition to permitting the stabilized tubes to remain in operation, the sleeve has the additional merit of being the best performer of all the designs. In addition, it is adaptable to remote installation. (orig.)

  5. The SCC testing of nuclear steam generator tubing materials

    Science.gov (United States)

    Doherty, P. E.; Sarver, J. M.; Miglin, B. P.

    1996-05-01

    The integrity of heat-exchanger tubes in a nuclear reaction system is crucial for the safe operation of a power plant. In order to study the corrosion behavior of certain alloys, constant extension rate (CERT) tests were performed on alloy 690 and alloy 800 nuclear steam generator tubing specimens. In this article, the CERT test results (such as maximum stress achieved and crack morphology) are correlated to tubing microstructure, chemistry, and manufacturing processes.

  6. 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.

  7. 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.

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

    International Nuclear Information System (INIS)

    Shoemaker, P.; Cislo, D.

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Shin, C. W.; No, H. C.

    2015-01-01

    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

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

    International Nuclear Information System (INIS)

    Prock, J.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. An improved steam generator model for the SASSYS code

    International Nuclear Information System (INIS)

    Pizzica, P.A.

    1989-01-01

    A new steam generator model has been developed for the SASSYS computer code, which analyzes accident conditions in a liquid metal cooled fast reactor. It has been incorporated into the new SASSYS balance-of-plant model but it can also function on a stand-alone basis. The steam generator can be used in a once-through mode, or a variant of the model can be used as a separate evaporator and a superheater with recirculation loop. The new model provides for an exact steady-state solution as well as the transient calculation. There was a need for a faster and more flexible model than the old steam generator model. The new model provides for more detail with its multi-mode treatment as opposed to the previous model's one node per region approach. Numerical instability problems which were the result of cell-centered spatial differencing, fully explicit time differencing, and the moving boundary treatment of the boiling crisis point in the boiling region have been reduced. This leads to an increase in speed as larger time steps can now be taken. The new model is an improvement in many respects. 2 refs., 3 figs

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  14. Development of KALANS to remove sludge of steam generator

    International Nuclear Information System (INIS)

    Jeong, W. T.; Yang, J. S.; You, G. W.; Sung, G. B.; Kim, B. Y.

    1999-01-01

    CECIL -4 lancing system by Foster-Miller has been used for the secondary side landing of Westinghouse type steam generators (YGN No. 1,2 and Kori No. 1,2,3,4) in Korean nuclear power plants. Shared use of only one lancing system for the 6 nuclear power plants required frequent transportation of the system from Nori to Youngkwang which may cause malfunction of sophisticated control equipment. Sometimes, the overlap in preventive maintenance schedule of nuclear power plants prevented the plant maintenance personnel from doing sufficient lancing work. Purchasing a new lancing system may be a candidate for this problem, but the price tag is surprisingly high. Furthermore, prompt maintenance was another serious problem because the manufacturer is far away in a foreign country. Last year, the steam generator of Kori No. 1 was changed to Δ60. The difference in the tube arrangement of the new steam generator made it impossible to use the CECIL system. Sufficient technological background of our development team and the conclusion that the development project is economically feasible made us to commence the design and fabrication of 'KEPCO Advanced LANcing System - KALANS.' This paper presents the state of the art on lancing system and the developed concept and design of KALANS

  15. Syngas Generation from Organic Waste with Plasma Steam Reforming

    Science.gov (United States)

    Diaz, G.; Leal-Quiros, E.; Smith, R. A.; Elliott, J.; Unruh, D.

    2014-05-01

    A plasma steam reforming system to process waste is in the process of being set up at the University of California, Merced. The proposed concept will use two different plasma regimes, i.e. glow discharge and arc torches to process a percentage of the total liquid waste stream generated at the campus together with shredded local organic solid waste. One of the main advantages of the plasma technology to be utilized is that it uses graphite electrodes that can be fed to the reactor to achieve continuous operation, thus, electrode or nozzle life is not a concern. The waste to energy conversion process consists of two stages, one where a mixture of steam and hydrogen is generated from the liquid in a glow-discharge cell, and a second stage where the mixture of exhaust gases coming out of the first device are mixed with solid waste in a reactor operating in steam reforming mode interacting with a plasma torch to generate high-quality syngas. In this paper, the results of a thermodynamic model developed for the two stages are shown. The syngas composition obtained indicates that the fraction of CO2 present decreases with increasing temperature and the molar fractions of hydrogen and carbon monoxide become dominant. The fraction of water vapour present in the product gases coming out of the second stage needs to be condensed before the syngas can be utilized in a prime mover.

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

    International Nuclear Information System (INIS)

    Olesovsky

    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)

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

    International Nuclear Information System (INIS)

    Papp, L.

    1995-01-01

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

  18. 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.

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

    International Nuclear Information System (INIS)

    Hart, R.S.

    2006-01-01

    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

  20. GVTRAN-PC-a steam generator transient simulator

    International Nuclear Information System (INIS)

    Nakata, H.

    1991-02-01

    Since an accuracy and inexpensive analysis capability is one of the desirable requirement in the reactor licensing procedure and, also, in instances when a severe accident sequence and its degree of severity must be estimated, the present work tries partially to fulfill that present need by developing a fast and acceptably accurate simulator. The present report presents the methodology utilized to develop GVTRAN-PC program, the steam generator simulation program for the microcomputer environment, which possess a capability to reproduce the experimental data with accuracies comparable to those of the mainframe simulators. The methodology is based on the mass and energy conservation in the control volumes which represents both the primary and the secondary fluid in the U-tube steam generator. The quasi-static momentum conservation in the secondary fluid control volumes determines in a semi-iterative scheme the liquid level in the feedwater chamber. The implementation of the moving boundary technique has allowed the tracking of the boundary of bulk boiling region with the utilization of a reduced number of control volumes in the tube region. GVTRAN-PC program has been tested against typical PWR pump trip transient experimental data and the calculation results showed good agreement in most representative parameters, viz. the feedchamber water-level and the steam dome pressure. (author)

  1. Energy balance and flow in steam generator part with sodium-water reaction

    International Nuclear Information System (INIS)

    Matal, O.

    1980-01-01

    Relations were derived for the calculation of heat liberated during the sodium water reaction in a tube failure in different parts of a steam generator. The results are graphically shown in i-T diagrams. Heat removal is described from the reaction zone to water and steam in undisturbed tubes and to the steam generator metal structure. (author)

  2. 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.

  3. Evolution of management activities and performance of the Point Lepreau Steam Generators

    International Nuclear Information System (INIS)

    Slade, J.; Keating, J.; Gendron, T.

    2007-01-01

    The Point Lepreau steam generators have been in service since 1983 when the plant was commissioned. During the first thirteen years of operation, Point Lepreau steam generator maintenance issues led to 3-4% unplanned plant incapability Steam generator fouling, corrosion, and the introduction of foreign materials during maintenance led to six tube leaks, two unplanned outages, two lengthy extended outages, and degraded thermal performance during this period. In recognition of the link between steam generator maintenance activities and plant performance, improvements to steam generator management activities have been continuously implemented since 1987. This paper reviews the evolution of steam generator management activities at Point Lepreau and the resulting improved trends in performance. Plant incapability from unplanned steam generator maintenance has been close to zero since 1996. The positive trends have provided a strong basis for the management strategies developed for post-refurbishment operation. (author)

  4. Test run of 50 MW steam generator for prototype fast breeder reactor of PNC

    International Nuclear Information System (INIS)

    Tanaka, Kenji

    1974-01-01

    A sodium heated steam generator is one of the most important machinery in a fast reactor plant, influencing the aspects of economy and reliability. In the Power Reactor and Nuclear Fuel Development Corporation (PNC), the 50 MW steam generator and associated testing facility have been developed, which lead to the demonstration of the steam generator and the establishment of plant control technology for the fast breeder reactor rackled by PNC. The steam generator is separate helical-coil type with an evaporator, a superheater and a reheater. Taking two to three years in construction, in June 1974, the steam generator with the testing facility was successfully operated at rated power of 50 MW continuously for 72 hours. The steam generator planned for the prototype FBR ''Monju'' is first outlined, together with its devlopment program. The 50 MW steam generator and the testing facility are then described, including construction and operation; finally, the results of test and future test plans are explained. (Mori, K.)

  5. Thermoelectric generation coupling methanol steam reforming characteristic in microreactor

    International Nuclear Information System (INIS)

    Wang, Feng; Cao, Yiding; Wang, Guoqiang

    2015-01-01

    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

  6. Full range nuclear power plant steam generator level control system

    International Nuclear Information System (INIS)

    Geets, J.M.

    1988-01-01

    In a method of controlling feedwater level in a steam generator of a pressurized water reactor nuclear power plant, the steam generator having secondary loop feedwater lines. The method is described including the steps of outputting a first signal from a first three-mode controller to control feedwater flow through the secondary loop feedwater lines at reactor power levels below a predetermined percentage, and outputting a second signal from a second three-mode controller to control feedwater flow through the secondary loop feedwater lines at reactor power levels above the predetermined percentage. The improvement in combination therewith comprises: receiving the first and second signals at a transition controller means; generating a first feedwater temperature-dependent function; operating on the first input signal with the first temperature dependent function to produce a third signal; generating a second feedwater temperature-dependent function; operating on the second input signal with the second temperature-dependent function to produce a fourth signal; and adding the third and fourth signals resulting from the operating steps to control the feedwater flow through the secondary loop feedwater lines throughout an entire range of reactor load

  7. Optimization of perforated distribution plate in steam generator PGV-1000MKO using CFD

    International Nuclear Information System (INIS)

    Volkov, V.Yu.; Golibrodo, L.A.; Krutikov, A.A.; Kudryavtsev, O.V.; Lakhov, D.A.; Nadinskij, Yu.N.; Nikolaeva, A.V.; Skibin, A.P.; Sotskov, V.V.

    2015-01-01

    Design of steam distribution perforated sheet SDPS for PGV-1000MKO was optimized using CFD methods, and guidelines for choosing the design of SDPS perforation were provided. Computational studies for optimized design of SDPS were performed for several characteristic variants of the mass water level in the steam generator and the distribution of the steam load. Recommendations for commissioning tests procedure for the SG were given on the basis of the research results. Processes in the PGV-1000MKO version with the collector equipped with ten steam nozzles were modeled. A comparison of technical parameters of the design with a single steam nozzle and with ten steam nozzles was performed [ru

  8. Nuclear reactor fuel element sub-assemblies

    International Nuclear Information System (INIS)

    Ashton, M.W.

    1975-01-01

    Reference is made to fuel element sub-assemblies for use in a Na cooled fast reactor. Such sub-assemblies may comprises a hexagonal bundle of slender fuel elements enclosed in a tubular sleeve, often referred to as a 'wrapper'. The fuel elements are spaced apart by helical wire wraps forming fins and which also space the wrapper from the bundle. The wire wraps make contact with the sheaths of adjacent elements and with the wrapper, so that each fuel element is well supported against thermal bowing, rattling and vibration, whilst allowing adequate coolant flow passages through the bundle. It has been found, however, that the outer fuel elements of the bundle are subject to over-cooling in this arrangement; this problem can, however, be largely overcome by reducing the flow passage between the bundle and the wrapper. In the arrangement described a wire filler is employed, extending along each outer coolant flow passage, and constructed in wave form. Fillers of such form have been found to reduce over-cooling considerably and they avoid the need for varied height wraps on the fuel elements. The fuel elements also have improved lateral support by contact with the fillers. (U.K.)

  9. Leak detection of steam or water into sodium in steam generators of liquid-metal fast breeder reactors

    International Nuclear Information System (INIS)

    Hans, R.; Dumm, K.

    1977-01-01

    The leakage of water or steam into sodium in LMFBR steam generators, including a study of how leaks are detected and located as well as the potential damage that could be caused by such leaks, is surveyed. The most interesting steam generator designs evolving in those countries that develop and construct LMFBRs are presented. The relevant protection measures are described. Fault conditions are defined and descriptions given of possible sequences of events leading to abnormal conditions in a steam generator. Taking into account theory, the potential of the hydrogen and oxygen detection systems is discussed. Different hydrogen and oxygen detection systems are fully described. In so far as interesting technical solutions are concerned, previously developed devices have also been taken into account. The way oxygen detection supplements hydrogen detection is described by listing the available oxygen measuring devices and the relevant theory. Only a few sonic and accelerometer measurements have been made on complete steam generator units so there is little system data available. Descriptions, however, have been included to give the state of the art achieved for the sensors and the achieved sensitivities or band widths. The potential of this monitoring method is made evident by adding the technical data of the sensors. Furthermore, the available systems for monitoring medium and large leakages are described. Finally, recommendations are made concerning steam generator development and the application of hydrogen and oxygen detection systems, as well as acoustic measuring methods for small-leakage detection

  10. Steam generator replacement at Doel 3 NPP (Belgium)

    International Nuclear Information System (INIS)

    Danhier, B.

    1993-01-01

    The reasons are presented that led to the conclusion that the most cost-effective strategy for the Doel 3 unit was the immediate replacement of the SG. Discussed are the advantages and drawbacks of the replacement techniques, the so-called 2, 3 and 4 cuts methods. The advantages are emphasized of intensive use of computer aided engineering in this kind of backfitting. The methodology applied to combine a power uprating of 10% over the nominal power with the steam generator replacement is presented. (author) 1 fig

  11. Steam generator leak detection at Bruce A Unit 1

    International Nuclear Information System (INIS)

    Maynard, K.J.; McInnes, D.E.; Singh, V.P.

    1997-01-01

    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)

  12. Influence of sodium water reaction on MONJU steam generator

    International Nuclear Information System (INIS)

    Takahashi, T.; Ohmori, Y.; Hoshi, Y.

    1984-01-01

    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)

  13. LMFBR steam generator leak detection development in the United States

    International Nuclear Information System (INIS)

    Magee, P.M.; Gerrels, E.E.; Greene, D.A.; McKee, J.

    1978-01-01

    Leak detection for Liquid Metal Fast Breeder Reactor steam generators is an important economic factor in the shutdown, repair and restart of a plant. Development of leak detection systems in the U.S. has concentrated on four areas: (1) chemical (H 2 and O 2 ) leak detection meters; (2) acoustic leak detection/location techniques; (3) investigation of leak behavior (enlargement, damage effects, plugging and unplugging); and (4) data management for plant operations. This paper discusses the status, design aspects, and applications of leak detection technology for LMFBR plants. (author)

  14. Modelling local chemistry in PWR steam generator crevices

    International Nuclear Information System (INIS)

    The accumulation of impurities in local regions of PWR Steam Generators (SG) has resulted in the accelerated corrosion of SG materials. The chemical conditions in crevices and sludge piles is dependent on thermal hydraulic and mass transfer processes as well as the physical chemistry of the concentrated solution itself. This paper discusses the different modelling approaches which can be used to describe the concentration process and the local chemistry in these regions. The limitations of each approach and the applicability of model results to field conditions are discussed in the paper. EPRI's program in this area, including past accomplishments and the models used in the MULTEQ code are described in the paper. (author)

  15. Method for repairing a steam turbine or generator rotor

    International Nuclear Information System (INIS)

    Clark, R.E.; Amos, D.R.

    1987-01-01

    A method is described for repairing low alloy steel steam turbine or generator rotors, the method comprising: a. machining mating attachments on a replacement end and a remaining portion of the original rotor; b. mating the replacement end and the original rotor; c. welding the replacement end to the original rotor by narrow-gap gas metal arc or submerged arc welding up to a depth of 1/2-2 inches from the rotor surface; d. gas tungsten arc welding the remaining 1/2-2 inches; e. boring out the mating attachment and at least the inside 1/4 inch of the welding; and f. inspecting the bore

  16. 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...

  17. Polymeric dispersants for control of steam generator fouling

    Energy Technology Data Exchange (ETDEWEB)

    Balakrishnan, P.V.; Klimas, S.J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Lepine, L. [Institut de Recherche d' Hydro-Quebec, Varennes, Quebec (Canada); Turner, C.W

    1999-05-01

    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-temperaturesettling 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

  18. An improved steam generator model for the SASSYS code

    International Nuclear Information System (INIS)

    Pizzica, P.A.

    1989-01-01

    A new steam generator model has been developed for the SASSYS computer code, which analyzes accident conditions in a liquid-metal-cooled fast reactor. It has been incorporated into the new SASSYS balance-of-plant model, but it can also function as a stand-alone model. The model provides a full solution of the steady-state condition before the transient calculation begins for given sodium and water flow rates, inlet and outlet sodium temperatures, and inlet enthalpy and region lengths on the water side

  19. Criteria for maintenance and repair - LMFBR steam generators

    International Nuclear Information System (INIS)

    Essebaggers, J.

    1975-01-01

    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)

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

    International Nuclear Information System (INIS)

    Tellier, N.; Zilliox, C.

    1984-10-01

    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

  1. Heat exchanger, particularly liquid sodium heated steam generator

    International Nuclear Information System (INIS)

    Robin, Marcel; Tillequin, Jean.

    1977-01-01

    This invention relates to a liquid sodium heated steam generator the characteristic of which is an annular distribution chamber fed by two independent and diametrically opposed manifolds on a common horizontal axis, issuing respectively into two adjacent compartments made in the chambers on both sides of a vertical transversal partition containing the axis of the casing and extending perpendicularly to the manifolds, each compartment being itself divided into a number of adjacent sectors marked by folded metal sheets fixed to the distributor and shaped so as to present in pairs and with the chamber opposite the manifold issuing into a compartment two independent ducts for distributing the sodium flow [fr

  2. WWER steam generator tube structural and leakage integrity

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  3. Quadratic controller syntheses for the steam generator water level

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

  4. Studying the process following small steam leaks into sodium using the counterflow steam generator model

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Sergeev, G.P.; Volchkov, L.G.; Sednev, A.P.; Makarov, V.M.; Tonov, B.I.

    1982-01-01

    Peculiarities of hydrodynamic conditions arising in a defective tube of the counterflow steam generator following water leaks into the sodium have been studied using suitable models made of 12Kh18N10T and 10Kh2M steels. Data on the rate of the steam leaks spontaneous evolution and corrosive effect on the opposite wall of the tube are obtained as a function of the leak position and sizes. On the basis of the analysis of the experimental results a conclusion is drawn that two stages are distinguished during the leak spontaneous evolution. Changes in the leak sizes occur gradually and remain small during the first stage. The leak increases quickly and exceeds the initial level more than by two orders during the second stage. The results of metallographlic analysis of the tube steels carried out after the experiments confirm the earlier obtianed data on the presence of inter. granular interaction in 12Kh18N10T steel following the effect of of the water-to-sodium reaction products, and on a preferentially frontal type of 10Kh2M steel corrosion

  5. Program for the development of design data: LMFBR steam generator materials

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-01-01

    Progress in fiscal year 1976 is reported for two tasks in the LMFBR steam generator materials development program. The primary objective of Task 18 is characterization of materials, mainly 2-1/4 Cr-1 Mo and austenitic stainless steels, to assure that satisfactory materials compatibility is achieved with LMFBR steam generator environments. Studies described include: the kinetics and magnitude of decarburization of 2-1/4 Cr-1 Mo in high temperature liquid sodium; decarburization effects on mechanical properties; stress corrosion susceptibility of 2-1/4 Cr-1 Mo in water/steam environments of LMFBR steam generators; and stainless steel mechanical properties within a carburizing sodium environment at high temperatures. In addition, two activities are in progress that directly support LMFBR steam generator design and fabrication. Under Subtask E, General Electric heads the working committee on steam generator materials for the Nuclear Systems Materials Handbook. Subtask F, an experimental effort, is investigating thermal degradation effects produced by particulate deposition (in sodium) on heat transfer surfaces. The Steam Generator Materials Qualification Task (Task 10-G) was initiated in support of the development of materials and processes used in Clinch River Breeder Reactor steam generators. The topics discussed are: LMFBR steam generator tubing; LMFBR steam generator tubesheet forgings (characterization of VAR/ESR melted 2-1/4 Cr-1 Mo); friction, wear and self-welding; nondestructive examination development; water chemistry studies; and transition joint development.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-30

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

  7. Fuqing nuclear power of nuclear steam turbine generating unit No.1 at the implementation and feedback

    International Nuclear Information System (INIS)

    Cao Yuhua; Xiao Bo; He Liu; Huang Min

    2014-01-01

    The article introduces the Fuqing nuclear power of nuclear steam turbine generating unit no.l purpose, range of experience, experiment preparation, implementation, feedback and response. Turn of nuclear steam turbo-generator set flush, using the main reactor coolant pump and regulator of the heat generated by the electric heating element and the total heat capacity in secondary circuit of reactor coolant system (steam generator secondary side) of saturated steam turbine rushed to 1500 RPM, Fuqing nuclear power of nuclear steam turbine generating unit no.1 implementation of the performance of the inspection of steam turbine and its auxiliary system, through the test problems found in the clean up in time, the nuclear steam sweep turn smooth realization has accumulated experience. At the same time, Fuqing nuclear power of nuclear steam turbine generating unit no.1 at turn is half speed steam turbine generator non-nuclear turn at the first, with its smooth realization of other nuclear power steam turbine generator set in the field of non-nuclear turn play a reference role. (authors)

  8. Vibration and wear characteristics of steam generator tubes

    International Nuclear Information System (INIS)

    Choi, Young Hwan

    2003-06-01

    This study investigates the fluid elastic instability characteristics of Steam Generator (SG) U-tubes with defect and the safety assessment of the potential for fretting-wear damages on Steam Generator (SG) U-tubes caused by foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions for determining the fluid elastic instability or fretting-wear parameters such as damping ratio, added mass and flow velocity are obtained from three-dimensional SG flow calculation using the ATHOS3 code. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, the wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the internal pressure on the vibration and fretting-wear characteristics of the tube

  9. Flow distribution in the inlet plenum of steam generator

    International Nuclear Information System (INIS)

    Khadamakar, H.P.; Patwardhan, A.W.; Padmakumar, G.; Vaidyanathan, G.

    2011-01-01

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

  10. Process technology development of Ni electroplating in steam generator tube

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-15

    Alloy 600 having superior resistance to corrosion is used as a steam generator tubing in Nuclear Power Plants. In spite of its high corrosion resistance, there are many tubes which have experienced the corrosion problems such as SCC, pitting under high temperature and high pressure environments of NPP, leading to a menace to the safety of NPPs as well as economical loss. A commonly applied approach to rehabilitation has been to repair the damaged areas of the tubes via the insertion of tubular sleeves which are either welded or mechanically bonded at their extremities to the host tube. Such intrusive sleeves have weak points, such as the crevices, the tube deformation and an introduction of stress onto the host tube which then usually requires stress relief to improve the in-service life. However a lot of problems including these during and after repairing can be solved by Ni electroplating having excellent corrosion resistance to such as SCC. This work is related to optimum process development for Ni electrodeposition inside damaged steam generator tubing for repairing and the damage prevention. The optimum electroplating process for planar specimens was developed and the electrodeposition was performed successfully inside tube specimens by using the modified and improved anode probe. The Ni-electrodeposit plated on the inner surfaces of the tube specimens was confirmed to show excellent SCC resistance. A multiple electrodeposition facility for simultaneous electroplating inside three tubes at the same time was built and proved to work properly.

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

    International Nuclear Information System (INIS)

    Hart, R.S.

    2007-01-01

    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

  12. An effect of downcomer feedwater fraction on steam generator performance with an axial flow economizer

    International Nuclear Information System (INIS)

    Jung, Byung Ryul; Park, Hu Shin; Chung, Duk Muk; Baik, Se Jin

    2000-01-01

    The effects of feedwater flow fraction introduced into the downcomer region have been evaluated in terms of steam generator performance based on the same steam generator thermal output for the Korea Standard Nuclear Power Plant (KSNP) steam generator. The KSNP steam generator design has an integral axial flow economizer which is designed such that most of the feedwater is introduced through the economizer region and only a portion of feedwater through the downcomer region. The feedwater flow introduced into the downcomer region is not normally controlled during the power operation. However, the actual feedwater fraction into the downcomer region may differ from the design flow depending on the as-built system and component characteristics. Investigated in this paper were the downcomer feedwater flow effects on the steam pressure, circulation ratio, internal void fraction and velocity distribution in the tube bundle region at the steady state operation using SAFE and ATHOS3 codes. The results show that the steam pressure increases and the resultant total feedwater flow increases with reducing the downcomer feedwater flow fraction for the same steam generator thermal output. The slight off-design condition of downcomer feedwater flow fraction renders no significant effect on the steam generator performance such as circulation ratios, steam qualities, void fractions and internal velocity distributions. The evaluation shows that the slight off-design downcomer feedwater flow fraction deviation up to ± 5% is acceptable for the steam generator performance

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  14. Development and Transient Analysis of a Helical-coil Steam Generator for High Temperature Reactors

    International Nuclear Information System (INIS)

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

    2011-01-01

    A high temperature gas-cooled reactor (HTGR) is under development by the Next Generation Nuclear Plant (NGNP) Project at the Idaho National Laboratory (INL). Its design emphasizes electrical power production which may potentially be coupled with process heat for hydrogen production and other industrial applications. NGNP is considering a helical-coil steam generator for the primary heat transport loop heat exchanger based on its increased heat transfer and compactness when compared to other steam generators. The safety and reliability of the helical-coil steam generator is currently under evaluation as part of the development of NGNP. Transients, such as loss of coolant accidents (LOCA), are of interest in evaluating the safety of steam generators. In this study, a complete steam generator inlet pipe break (double ended pipe break) LOCA was simulated by an exponential loss of primary side pressure. For this analysis, a model of the helical-coil steam generator was developed using RELAP5-3D, an INL inhouse systems analysis code. The steam generator model behaved normally during the transient simulating the complete steam generator inlet pipe break LOCA. Further analysis is required to comprehensively evaluate the safety and reliability of the helical-coil steam generator design in the NGNP setting.

  15. Crack Growth Prediction of the Steam Turbine Generator Shaft

    International Nuclear Information System (INIS)

    Jiang Dongxiang; Liu Chao

    2011-01-01

    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.

  16. Seismic analysis of steam generator and parameter sensitivity studies

    International Nuclear Information System (INIS)

    Qian Hao; Xu Dinggen; Yang Ren'an; Liang Xingyun

    2013-01-01

    Background: The steam generator (SG) serves as the primary means for removing the heat generated within the reactor core and is part of the reactor coolant system (RCS) pressure boundary. Purpose: Seismic analysis in required for SG, whose seismic category is Cat. I. Methods: The analysis model of SG is created with moisture separator assembly and tube bundle assembly herein. The seismic analysis is performed with RCS pipe and Reactor Pressure Vessel (RPV). Results: The seismic stress results of SG are obtained. In addition, parameter sensitivities of seismic analysis results are studied, such as the effect of another SG, support, anti-vibration bars (AVBs), and so on. Our results show that seismic results are sensitive to support and AVBs setting. Conclusions: The guidance and comments on these parameters are summarized for equipment design and analysis, which should be focused on in future new type NPP SG's research and design. (authors)

  17. Preventive and corrective actions for tube degradation and new steam generator design concept

    International Nuclear Information System (INIS)

    Tsuge, A.; Hirano, H.; Sato, M.; Takamatsu, H.

    2004-01-01

    This paper describes the updated comprehensive overview on, (1) Tube degradation experiences through twenty three years operation of PWR Steam Generators in Japan. (2) Corrective and preventive techniques for tube repair operations, non-destructive examinations, and up-graded water chemistry control. (3) Strategy on the option of Steam Generator replacement. (4) Up-graded design features of Mitsubishi Steam Generator based on the long term operating experiences. (author)

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

    International Nuclear Information System (INIS)

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

    1978-02-01

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

  19. Nuclear reactor fuel element sub-assemblies

    International Nuclear Information System (INIS)

    Dodd, J.A.; Ashton, M.W.; Aubertin, J.C.

    1975-01-01

    Reference is made to fuel element sub-assemblies for use in a Na cooled fast reactor. Such sub-assemblies may comprise a hexagonal bundle of slender fuel elements enclosed in a tubular sleeve, often referred to as a 'wrapper'. The fuel elements are spaced apart by helical wire wraps forming fins and which also space the wrapper from the bundle. The wire wraps make contact with the sheaths of adjacent elements and with the wrapper, so that each fuel element is well supported against thermal, bowing, rattling and vibration, whilst allowing adequate coolant flow passages through the bundle. The arrangement of the fuel elements usually provides two groups of passageways for coolant flow through the bundle -an inner group each bounded by four fuel elements, and an outer group each bounded by two outer fuel elements and the wrapper. It has been found, however, that this arrangement results in over-cooling of the outer fuel elements. An arrangement is described that overcomes this disadvantage to a considerable extent. The fuel elements are arranged in three groups - an inner group, and intermediate group and an outer group. The fins on the outer group elements are helically wound at half the pitch of the other two groups and alternate outer group elements are wound opposite handed to abutting outer row elements; elements of the outer and intermediate groups are orientated and end located to enable them to nest together. By utilising half pitch helical fins on the outer elements a greater number of lateral support points is obtained for the outer and intermediate groups of elements and the increased number of turns of fins on the outer elements assist in reducing overcooling of the outer elements. (U.K.)

  20. Recent developments in plugging of steam generator tubes

    International Nuclear Information System (INIS)

    Buhay, S.; Abucay, R.C.

    1995-01-01

    Mechanical Plugging capability has been developed for Bruce Nuclear Generating Station (BNGS) steam generator (SG) tubes and Darlington Nuclear Generating Station (DNGS) SG tubes and tubesheet holes. The plug concept was a modified ABB/Combustion Engineering Inconel 690 plug with a nickel band, rolled into the tube or tubesheet hole from the primary side of the tubesheet. The qualification program included analytical justification of the plug body and experimental testing to verify the leak tightness of the rolled joint under conditions which meet or exceed all service or design requirements. Tools and procedures were developed and tested for manual and remote/robotic installation and removal of the mechanical plugs. Additionally, tools and procedures were developed to plug tubes/tubesheet holes at DNGS in the event the steam generator is recalled to service to act as a heat sink. A crew of Ontario Hydro personnel were trained and qualified for the installation of mechanical plugs for permanent and recall applications. During the DNGS Unit 4 spring 1995 outage, 6 tubes were plugged and the 'Recall Plugging Capability' was deployed and ready for use during a primary side SG tube removal. The mechanical plugs were installed manually with a typical 3 minute/plug in-bowl duration time with an average radiation dose of 12.5 mrem per plug. This compares favourably with manual plug welding during the same outage in the same SG bowl at approximately 15-30 minutes/plug in-bowl duration with an average radiation dose of 117 mrem/plug. (author)

  1. Safety of LMFBR steam generators - IAEA specialists' meeting in the Hague

    International Nuclear Information System (INIS)

    Matal, O.

    1984-01-01

    The meeting was held on November 9-11, 1983 in five sessions. The first session discussed problems of the integrity and reliability of steam generators in different countries. The second session heard papers on experimental facilities and the technology of leakage simulation. The third session discussed detectors of steam generator leaks. The fourth session was devoted to the mathematical modelling of processes in steam generators owing to big leaks. The fifth session dealt with the extrapolation of experimental results to the steam generator. (E.S.)

  2. Development of an acoustic steam generator leak detection system using delay-and-sum beamformer

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka

    2009-01-01

    A new acoustic steam generator leak detection system using delay-and-sum beamformer is proposed. The major advantage of the delay-and-sum beamformer is it could provide information of acoustic source direction. An acoustic source of a sodium-water reaction is supposed to be localized while the background noise of the steam generator operation is uniformly distributed in the steam generator tube region. Therefore the delay-and-sum beamformer could distinguish the acoustic source of the sodium-water reaction from steam generator background noise. In this paper, results from numerical analyses are provided to show fundamental feasibility of the new method. (author)

  3. A study on the steam generator data base and the evaluation of chemical environment

    International Nuclear Information System (INIS)

    Yang, Kyung Rin; Yoo, Je Hyoo; Lee, Eun He; Hong, Kwang Pum

    1990-01-01

    In order to make steam generator data base, the basic plant information and water quality control data on the steam generators of the PWR nuclear power plant operating in the world have been collected by EPRI. In this project, the basic information and water quality control data of the domestic PWR nuclear power plants were collected to make steam generator data base on the basic of the EPRI format table, and the computerization of them was performed. Also, the technical evaluation of chemical environments on steam generator of the Kori 2 plant chemists. Workers and researchers working at the research institute and universities and so on. Especially, it is able to be used as a basic plant information in order to develop an artificial intellegence development system in the field on the technical development of the chemical environment. The scope and content of the project are following. The data base on the basic information data in domestic PWR plant. The steam generator data base on water quality control data. The evaluation on the chemical environment in the steam generators of the Kori 2 plant. From previous data, it is concluded as follows. The basic plant information on the domestic PWR power plant were computerized. The steam generator data base were made on the basis of EPRI format table. The chemical environment of the internal steam generators could be estimated from the analytical evaluation of water quality control data of the steam generator blowdown. (author)

  4. Flow-induced vibration in LMFBR steam generators: a state-of-the-art review

    International Nuclear Information System (INIS)

    Shin, Y.S.; Wambsganss, M.W.

    1975-05-01

    This state-of-the-art review identifies and discusses existing methods of flow-induced vibration analysis applicable to steam generators, their limitations, and base-technology needs. Also included are discussions of five different LMFBR steam-generator configurations and important design considerations, failure experiences, possible flow-induced excitation mechanisms, vibration testing, and available methods of vibration analysis. The objectives are to aid LMFBR steam-generator designers in making the best possible evaluation of potential vibration in steam-generator internals, and to provide the basis for development of design guidelines to avoid detrimental flow-induced vibration

  5. The temperature control and water quality regulation for steam generator secondary side hydrostatic test

    International Nuclear Information System (INIS)

    Xiao Bo; Liu Dongyong

    2014-01-01

    The secondary side hydrostatic test for the steam generator of M310 unit is to verify the pressure tightness of steam generator secondary side tube sheet and related systems. As for the importance of the steam generator, the water temperature and water quality of hydrostatic test has strict requirements. The discussion on the water temperature control and water quality regulation for the secondary loop hydrostatic test of Fuqing Unit 1 contribute greatly to the guiding work for the preparation of the steam generator pressure test for M310 unit. (authors)

  6. SWAAM-code development and verification and application to steam generator designs

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  7. North Anna Power Station - Unit 1: Overview of steam generator replacement project activities

    International Nuclear Information System (INIS)

    Gettler, M.W.; Bayer, R.K.; Lippard, D.W.

    1993-01-01

    The original steam generators at Virginia Electric and Power Company's (Virginia Power) North Anna Power Station (NAPS) Unit 1 have experienced corrosion-related degradation that require periodic inspection and plugging of steam generator tubes to ensure their continued safe and reliable operation. Despite improvements in secondary water chemistry, continued tube degradation in the steam generators necessitated the removal from service of approximately 20.3 percent of the tubes by plugging, (18.6, 17.3, and 25.1 for steam generators A, B, and C, respectively). Additionally, the unit power was limited to 95 % during, its last cycle of operation. Projections of industry and Virginia Power experience indicated the possibility of mid-cycle inspections and reductions in unit power. Therefore, economic considerations led to the decision to repair the steam generators (i.e., replace the steam generator lower assemblies). Three new Model 51F Steam Generator lower assembly units were ordered from Westinghouse. Virginia Power contracted Bechtel Power Corporation to provide the engineering and construction support to repair the Unit 1 steam generators. On January 4, 1993, after an extended coastdown period, North Anna Unit 1 was brought off-line and the 110 day (breaker-to-breaker) Steam Generator Replacement Project (SGRP) outage began. As of this paper, the outage is still in progress

  8. Composite electric generator equipped with steam generator for heating reactor coolant

    Energy Technology Data Exchange (ETDEWEB)

    Watabe, Masaharu; Soman, Yoshindo; Kawanishi, Kohei; Ota, Masato

    1997-08-12

    The present invention concerns a composite electric generator having coolants, as a heating source, of a PWR type reactor or a thermonuclear reactor. An electric generator driving gas turbine is disposed, and a superheater using a high temperature exhaust gas of the gas turbine as a heating source is disposed, and main steams are superheated by the superheater to elevate the temperature at the inlet of the turbine. This can increase the electric generation capacity as well as increase the electric generation efficiency. In addition, since the humidity in the vicinity of the exit of the steam turbine is reduced, occurrence of loss and erosion can be suppressed. When cooling water of the thermonuclear reactor is used, the electric power generated by the electric generator driven by the gas turbine can be used upon start of the thermonuclear reactor, and it is not necessary to dispose a large scaled special power source in the vicinity, which is efficient. (N.H.)

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

    International Nuclear Information System (INIS)

    Millett, P.

    1997-01-01

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

  10. Experiences with industrial solar process steam generation in Jordan

    Science.gov (United States)

    Krüger, Dirk; Berger, Michael; Mokhtar, Marwan; Willwerth, Lisa; Zahler, Christian; Al-Najami, Mahmoud; Hennecke, Klaus

    2017-06-01

    At the Jordanian pharmaceuticals manufacturing company RAM Pharma a solar process heat supply has been constructed by Industrial Solar GmbH in March 2015 and operated since then (Figure 1). The collector field consists of 394 m² of linear Fresnel collectors supplying saturated steam to the steam network at RAM Pharma at about 6 bar gauge. In the frame of the SolSteam project funded by the German Federal Ministry for Economic Affairs and Energy (BMWi) the installation has been modified introducing an alternative way to separate water and steam by a cyclone. This paper describes the results of experiments with the cyclone and compares the operation with a steam drum. The steam production of the solar plant as well as the fuel demand of the steam boiler are continuously monitored and results are presented in this paper.

  11. Preliminary design: duplex tube low-pressure saturated steam generator for large LMFBR plant. Final report

    International Nuclear Information System (INIS)

    Dawson, B.E.

    1979-10-01

    A preliminary design was completed for a steam generator which is applicable to a four loop LMFBR plant of 1000 MWe gross output employing dry and saturated steam conditions. Two steam generators of 364 MWt thermal capacity would be in each loop. The steam generator is a straight, duplex tube design with a shell bellows for thermal expansion and a single tubesheet with an integral leak detection system. The report outlines a proposed series of development tasks to provide additional design data for the duplex tube and to demonstrate the steam generator design with test modules. The design can be adapted readily as either the evaporator or the superheater for a superheated steam cycle with temperatures up to at least 430 0 C (800 0 F), which is below the creep range for the 2 1/4 Cr-1 Mo material

  12. Optimization of costs for the DOEL 3 steam generator replacement

    International Nuclear Information System (INIS)

    Leblois, C.

    1994-01-01

    Several aspects of steam generator replacement economics are discussed on the basis of the recent replacement carried out in the Doel 3 unit. The choice between repair of replacement policies, as well as the selection of the intervention date were based on a comparison of costs in which various possible scenarios were examined. The contractual approach for the different works to be performed was also an important point, as well as the project organization in which CAD played an important role. This organization allowed to optimize the outage duration and to realize numerous interventions in the reactor building in parallel with the replacement itself. A last aspect of the optimization of costs is the possibility to uprate the plant power. In the case of Doel 3, the plant restarted with a nominal power increased by 10%, of which 5,7% were possible by the increase of the SG heat transfer area. (Author) 6 refs

  13. Developments in steam generator leak detection at Ontario Hydro

    International Nuclear Information System (INIS)

    Maynard, K.J.; Singh, V.P.

    1998-01-01

    A method for locating small tube leaks in steam generators has been developed and implemented at Ontario Hydro. The technique utilizes both helium leak detection and moisture leak detection. The combination of these two methods allows tube leaks to be detected in any part of the tube bundle, including those submerged below water near the tubesheet. The estimated detection limits for the helium and moisture leak detection systems are 0.001 kg/hr and 0.05 kg/hr respectively, expressed as leak rates measured at typical boiler operating conditions. This technology is best utilized in situations where the leak rate under operating conditions is smaller than the practical limit for fluorescein dye techniques (∼2 kg/hour). Other novel techniques have been utilized to increase the reliability and speed of the boiler leak search process. These include the use of argon carrier gas to stabilize the buoyant helium gas in the boiler secondary. (author)

  14. Duplex-tube sodium-indication steam generator

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.

    1984-01-01

    The steam generator with duplex tubes and sodium indication is connected to the main sodium input and output via the inlet and outlet chambers and has indication spaces connected to the interspaces of the duplex tubes. The first indication space is linked with the auxiliary inlet pipe to the inlet chamber and the second indication space is connected with the auxiliary pipe to the outlet chamber. Mounted to the auxiliary inlet pipe is at least one closure, i.e., a valve or electromagnetic stop. Mounted on the auxiliary outlet pipe is an indication sensor, e.g., a sodium flow sensor. At least one indication space is provided with an alarm sensor, e.g., a thermocouple, a pressure gauge and one sensor to monitor the hydrogen content of sodium. (J.P.)

  15. Innovated feed water distributing system of VVER steam generators

    International Nuclear Information System (INIS)

    Matal, O.; Sousek, P.; Simo, T.; Lehota, M.; Lipka, J.; Slugen, V.

    2000-01-01

    Defects in feed water distributing system due to corrosion-erosion effects have been observed at many VVER 440 steam generators (SG). Therefore analysis of defects origin and consequently design development and testing of a new feed water distributing system were performed. System tests in-situ supported by calculations and comparison of measured and calculated data were focused on demonstration of long term reliable operation, definition of water flow and water chemical characteristics at the SG secondary side and their measurements and study of dynamic characteristics needed for the innovated feed water distributing system seismic features approval. The innovated feed water distributing system was installed in the SGs of two VVER units already. (author)

  16. Steam generator cleaning campaigns at Bruce A: 1993-1996

    International Nuclear Information System (INIS)

    Puzzuoli, F.V.; Leinonen, P.J.; Lowe, G.A.

    1997-01-01

    Boiler chemical cleaning (BOCC) and high-pressure water lancing operations were performed during the Bruce A 1993 Unit 3, 1994 Unit 3, 1995 Unit 1 and 1996 Unit 3 outages to remove secondary side deposits. High-pressure water lancing focused on three boiler areas: tube support plates, to remove broached hole deposits, hot leg U-bend supports to dislodge deposits contributing to boiler tube stress corrosion cracking and tube sheets with the aim of removing accumulated sludge piles and post BOCC insoluble residues. The chemical cleaning processes applied were modified versions of the one developed by the Electric Power Research Institute/Steam Generator Owners Group. During these BOCC operations, corrosion for several key boiler materials was monitored and was well below the specified allowances

  17. Burnout power in once-through tubular steam generators

    International Nuclear Information System (INIS)

    Campolunghi, F.; Cumo, M.; Ferrari, G.; Leo, R.; Vaccaro, G.

    1974-01-01

    The present experimental research provides experimental data in a design field where existing burnout correlations are usually extrapolated outside their validity range. Low surface heat fluxes and very long geometries typical of L.M.F.B.R. sodium heated steam generators require full scale water tests. 218 burnout qualities at pressures ranging from 60 to 160 kg/cm 2 and specific mass flowrates from 110 to 260 g/cm 2 s were measured and tested against the most known burnout correlations. Physically it appears that burnout quality is substantially independent from the flowrate, while a sharp subdivision with pressure suggests that for 60 less than p less than 90 kg/cm 2 and 110 less than p less than 160 kg/cm 2 two different burnout mechanisms may set up. (U.S.)

  18. Two-phase dynamics of gas-heated steam generators

    International Nuclear Information System (INIS)

    Schittke, H.J.

    1977-01-01

    The dynamic behavior of a once-through steam generator plant operating in the secondary loop of a gas-cooled high-temperature reactor is considered. The mathematical model used for the description of the thermohydraulics of the problem comprises not only the dynamic behavior of the primary heating gas flow and the tube wall temperatures but especially the effects of pressure dynamics in the secondary fluid and the relevant two-phase flow phenomena: using an additional momentum balance equation for the dynamics of the slip velocity it is shown that the analytical computation of the slip velocity it is shown that the analytical computation of slip and two-phase pressure drop effects from the model equations is possible without the use of external correlations. Based on this mathematical model a generally applicable computer model is used to simulate the dynamic response of a given system

  19. Nickel electroplating as a remedy to steam generator tubing PWSCC

    International Nuclear Information System (INIS)

    Michaut, B.; Steltzlen, F.; Sala, B.; Laire, Ch.; Stubbe, J.

    1993-01-01

    Nickel plating appears to be a versatile process, as the application field, even if always used against PWSCC, is different from plant-to-plant. Its usage has been from a purely preventive action on tubes without defects, to a corrective action on through-wall cracked and leaking tubes. As a background for the large scale on-site operations of Doel 2 in 1990 (345 tubes) and Tihange 2 in 1992 (600 tubes), studies on four points are outlined, i.e. corrosion tests, stress measurements, sulfamate bath quality control, and in-service inspection. In conclusion, it appears that the nickel plating technique, following a case-by-case study, can often be a convenient remedy against Alloy 600 stress corrosion problems. New applications, in locations other than the steam generator field are under consideration

  20. Characterization of corrosion deposits on components of candu steam generators

    International Nuclear Information System (INIS)

    Dinu, A.; Mincu, M.; Stanciulescu, M.; Diniasi, D.

    2015-01-01

    X-ray diffraction at small angles of incidence was used to investigate oxidised samples obtained from Incoloy 800, steel 516 and Inconel 600, materials used to manufacture different components of CANDU steam generators. The samples were exposed for 157 days in demineralised water with LiOH (pH=10.5), at 3100C and 5MPa. In the case of 516 steel it was observed that the magnetite is formed at interface metal-oxide, while the external layer is formed preponderant from hematite particles. The external layer present on the surface of Inconel 600 is formed by nickel hydroxide, while the spinelic compounds such as (Cr, Fe)3O4 are distributed in the inner layer. Because the oxide formed on the surface of Incoloy 800 sample is very thin, there is not major difference regarding the composition in the depth profile. The X-rays results were sustained by the SEM results. (authors)

  1. Non linear identification applied to PWR steam generators

    International Nuclear Information System (INIS)

    Poncet, B.

    1982-11-01

    For the precise industrial purpose of PWR nuclear power plant steam generator water level control, a natural method is developed where classical techniques seem not to be efficient enough. From this essentially non-linear practical problem, an input-output identification of dynamic systems is proposed. Through Homodynamic Systems, characterized by a regularity property which can be found in most industrial processes with balance set, state form realizations are built, which resolve the exact joining of local dynamic behaviors, in both discrete and continuous time cases, avoiding any load parameter. Specifically non-linear modelling analytical means, which have no influence on local joined behaviors, are also pointed out. Non-linear autoregressive realizations allow us to perform indirect adaptive control under constraint of an admissible given dynamic family [fr

  2. In situ corrosion monitoring of steam generators. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kendig, M W; Isaacs, H S

    1978-06-01

    An ac electrochemical technique which meets the basic requirements for an in situ localized corrosion monitor within the secondary coolant of PWR steam generators has been investigated. The technique uses two electrodes to measure the electrochemical impedance of a surface in an occluded region with high heat flux. The impedance is related to the kinetics of corrosion. Marked decreases indicate the onset of a high corrosion rate. Experiments have demonstrated the ability of the technique to determine the onset of corrosion under conditions of high solution resistance and solution agitation due to local boiling. Experiments have shown the technique operates similarly in pressurized 300/sup 0/C water, 1,400 ppM in Na/sub 2/SO/sub 4/.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-15

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

  4. Check of condition of steam generators, volume compensators and turbine condensers in nuclear power plants

    International Nuclear Information System (INIS)

    Matal, O.; Klinga, J.; Holy, F.; Sobotka, J.

    1989-01-01

    A negative pressure leak detector is described designed for leak testing of tubes in steam generators and steam turbine condensers. The principle, operation and use are described of inflatable bags and an inflatable platform. The bags are designed for insulating and sealing spaces in nuclear reactor components while the inflatable platform is used in pressurizer inspections and repairs. Their properties, and other facilities for detecting leaks in steam generator tubes are briefly described. (M.D.). 3 figs

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

    International Nuclear Information System (INIS)

    2007-12-01

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

  6. Importance of deposit information in the design and execution of steam generator chemical cleaning

    International Nuclear Information System (INIS)

    Flores, O.; Remark, J.

    1997-01-01

    During the planning stages of the chemical cleaning of the San Onofre Nuclear Generating Station (SONGS) units 2 and 3 steam generators, it was determined that an understanding of the steam generator deposit loading and composition was essential to the design and success of the project. It was also determined that qualification testing, preferably with actual deposits from the SONGS steam generators, was also essential. SONGS units 2 and 3 have Combustion Engineering (CE)-designed pressurized water reactors. Each unit has two CE model 3410 steam generators. Each steam generator has 9350 alloy 600 tubes with 1.9-cm (3/4 in.) outside diameter. Unit 2 began commercial operation in 1983, and unit 3, in 1984. The purpose of this technical paper is to explain the effort and methodology for deposit composition, characterization, and quantification. In addition, the deposit qualification testing and design of the cleaning are discussed

  7. Fast reactor steam generators with sodium on the tube side. Design and operational parameters

    International Nuclear Information System (INIS)

    1994-01-01

    A comparison of design and operational characteristics as well as analysis of experience gained during the long terms operation of the Micro Module Inverse Steam Generator and Module Inverse Steam Generator at BOR 60 reactor are main aims of this technical report. 20 refs, 47 figs, 14 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  10. Effect of tube plugging in the thermalhydraulic performance of 'U' tube steam generators

    International Nuclear Information System (INIS)

    Braga, C.V.M.; Carajilescov, P.

    1981-05-01

    The thermalhydraulic performance of Angra II steam generator has been simulated using the model developed by Braga, C.V.M., 'Thermohydraulic model for steam generator of PWR power plants', in steady state, with plugging up to 40% of total number of tubes. (E.G.) [pt

  11. Reliability study: steam generation and distribution system, Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    Baker, F.E.; Davis, E.L.; Dent, J.T.; Walters, D.E.; West, R.M.

    1982-10-01

    A reliability study for determining the ability of the Steam Generation and Distribution System to provide reliable and adequate service through the year 2000 has been completed. This study includes an evaluation of the X-600 Steam Plant and the steam distribution system. The Steam Generation and Distribution System is in good overall condition, but to maintain this condition, the reliability study team made twelve recommendations. Eight of the recommendations are for repair or replacement of existing equipment and have a total estimated cost of $540,000. The other four recommendations are for additional testing, new procedure implementation, or continued investigations

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

    International Nuclear Information System (INIS)

    Kassen, W.R.

    1987-01-01

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

  13. Specific features of emergency processes associated with water leacs into sodium in a reverse steam generator

    International Nuclear Information System (INIS)

    Sroelov, V.S.; Nikol'skij, R.V.; Chernobrovkin, Yu.V.; Privalov, Yu.V.; Bocharin, P.P.; Shtynda, Yu.E.

    1986-01-01

    Experimental and theoretical data characterizing the development of emergency processes arising in the course of water leaks into sodium in a reverse steam generator (sodium in tubes, water in intertube space) are considered. The results of calculations performed for BOR-60 reactor steam generator at initial leaks of 0.01 and 0.55 g/s are presented. It is shown that in the reverse steam generator the development of accident occurs much slower than in steam generators of traditional design. At same stage of accident sodium is displaced from the damaged tube and as a result the destruction of tube material discontinues. The conclusion is drawn that by the development of emergency protection systems for reverse steam generator the requirements for sensitivity and fast response of leak detectors could be reduced

  14. Specialists' meeting on maintenance and repair of LMFBR steam generators. Summary report

    International Nuclear Information System (INIS)

    2002-01-01

    The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topic areas were discussed by participants: National review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; Research and Development work on maintenance and repair; Experience on steam generator maintenance and repair. During the meeting papers were presented by the participants on behalf of their countries and organizations. A final discussion session was held and summaries, general conclusions and recommendations were approved by consensus

  15. Wear behavior of steam generator tubes in nuclear power plant operating condition

    International Nuclear Information System (INIS)

    Kim, In-Sup; Hong, Jin-Ki; Kim, Hyung-Nam; Jang, Ki-Sang

    2003-01-01

    Reciprocating sliding wear tests were performed on steam generator tubes materials at steam generator operating temperature. The material surfaces react with oxygen to form oxides. The oxide properties such as formation rate and mechanical properties are varied with the test temperature and alloy composition. So, it is important to investigate the wear properties of each steam generator tube materials in steam generator operating condition. The tests results indicated that the wear coefficient in work rate model of alloy 690 was faster than that of alloy 800. From the scanning electron microscopy observation, the wear scars were similar each other and worn surfaces were covered with oxide layers. It seemed that the oxide layers were formed by wear debris sintering or cold welding and these layer properties affected the wear rate of steam generator tube materials. (author)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  17. Locating hot and cold-legs in a nuclear powered steam generation system

    Science.gov (United States)

    Ekeroth, Douglas E.; Corletti, Michael M.

    1993-01-01

    A nuclear reactor steam generator includes a reactor vessel for heating water and a steam generator with a pump casing at the lowest point on the steam generator. A cold-leg pipe extends horizontally between the steam generator and the reactor vessel to return water from the steam generator to the reactor vessel. The bottom of the cold-leg pipe is at a first height above the bottom of the reactor vessel. A hot-leg pipe with one end connected to the steam generator and a second end connected to the reactor vessel has a first pipe region extending downwardly from the steam generator to a location between the steam generator and the reactor vessel at which a bottom of the hot-leg pipe is at a second height above the bottom of the reactor vessel. A second region extends from that location in a horizontal direction at the second height to the point at which the hot-leg pipe connects to the reactor vessel. A pump is attached to the casing at a location below the first and second heights and returns water from the steam generator to the reactor vessel over the cold-leg. The first height is greater than the second height and the bottom of the steam generator is at a height above the bottom of the reactor vessel that is greater than the first and second heights. A residual heat recovery pump is below the hot-leg and has an inlet line from the hot-leg that slopes down continuously to the pump inlet.

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

    Science.gov (United States)

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

    2016-01-01

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

  19. 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)

  20. Improved eddy-current inspection for steam generator tubing

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  1. Short history of steam generators in the USSR

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The first power stations appeared in Russia in the late 1880s. Early pioneers in generator design are mentioned. Lenin considered power production essential for rapid industrialization. In the early 1920s power stations were designed to make use of local fuels: peat, brown coal, and anthracite culm. The high-pressure, once-through boiler technology was introduced in the 1930s. At the same time cogeneration was a widely used technology, and efforts were being made to increase boiler capacity. In 1939, in line with prewar policies of dispersing Soviet industry to protect it from enemy attack, boiler capacity was limited to 25 tons/hr. Almost all of the multi-drum boilers were destroyed as a result of WWII. A novel method of salvaging the boilers by welding 2 or 3 units together to make a single unit was implemented after the war. Research organizations are mentioned along with their specific contributions. Modern steam generators use boiler turbines and supercritical once-through boilers. It was only in the late 1950s that economic planners discovered that oil and gas in power stations was cost effective. In 1954 a 5-MW graphite-water reactor became the world's first nuclear power plant. For the next 20 years, two types of nuclear reactors began production: pressurized water-cooled, water-moderated reactors in the 200-400 MW range; and channel-type graphite-moderated, water-cooled reactors in the 100-200 MW range

  2. Determination of moisture content in steams and variation in moisture content with operating boiler level by analyzing sodium content in steam generator water and steam condensate of a nuclear power plant using ion chromatographic technique

    International Nuclear Information System (INIS)

    Pal, P.K.; Bohra, R.C.

    2015-01-01

    Dry steam with moisture content less than <1% is the stringent requirements in a steam generator for good health of the turbine. In order to confirm the same, determination of sodium is done in steam generator water and steam condensate using Flame photometer in ppm level and ion chromatograph in ppb level. Depending on the carry over of sodium in steam along with the water droplet (moisture), the moisture content in steam was calculated and was found to be < 1% which is requirements of the system. The paper described the salient features of a PHWR, principle of Ion Chromatography, chemistry parameters of Steam Generators and calculation of moisture content in steam on the basis of sodium analysis. (author)

  3. Thermodynamic analysis of heat recovery steam generator in combined cycle power plant

    OpenAIRE

    Ravi Kumar Naradasu; Rama Krishna Konijeti; Sita Rama Raju Venkata Alluru

    2007-01-01

    Combined cycle power plants play an important role in the present energy sector. The main challenge in designing a combined cycle power plant is proper utilization of gas turbine exhaust heat in the steam cycle in order to achieve optimum steam turbine output. Most of the combined cycle developers focused on the gas turbine output and neglected the role of the heat recovery steam generator which strongly affects the overall performance of the combined cycle power plant. The present paper is a...

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

    Science.gov (United States)

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

    1994-05-03

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

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

    Science.gov (United States)

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

    1994-01-01

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

  6. TRAC analysis of steam-generator overfill transients for TMI-1

    International Nuclear Information System (INIS)

    Bassett, B.

    1983-01-01

    A reactor safety issue concerning the overfilling of once-through steam generators leading to combined primary/secondary blowdown has been raised recently. A series of six calculations, performed with the LWR best-estimate code, TRAC-PD2, on a Babcock and Wilcox Plant (TMI-1), was performed to investigate this safety issue. The base calculation assumed runaway main feedwater to one steam generator causing it to overfill and to break the main steam line. Four additional calculations build onto the base case with combinations of a pump-seal failure, a steam-generator tube rupture, and the pilot-operated relief valve not reseating. A sixth calculation involved only the rupture of a single steam-generator tube. The results of these analyses indicate that for the transients investigated, the emergency cooling system provided an adequate make-up coolant flow to mitigate the accidents

  7. Thermal hydraulic characteristics of a double-walled tube advanced nuclear steam generator

    International Nuclear Information System (INIS)

    Cho, S.M.; Seltzer, A.H.

    1989-01-01

    In this paper the thermal hydraulic characteristics of double-walled tube steam generator designed for sodium-cooled nuclear reactors are presented. The double-walled tube construction, along with double-barrier welds for tube-to-tubesheet joints, virtually eliminates the probability of heat transfer tube failure. Considerations are given to the use of the internal core tube, helical vane swirl generator, external protector tube, and variably perforated flow baffles to improve thermal and hydraulic performance of the steam generator. These thermal hydraulic design features with a particular reference to a 432 MW PRISM steam generator are discussed

  8. Ontario Hydro's operating experience with steam generators with specifics on Bruce A and Bruce B problems

    International Nuclear Information System (INIS)

    Eatock, J.W.; Patterson, R.W.; Dyck, R.W.

    1991-01-01

    The performance of the steam generators in Ontario Hydro nuclear power stations is reviewed. This performance has generally been outstanding compared to world averages, with very low tube failure and plugging rates. Steam generator problems have made only minor contributions to Ontario Hydro nuclear station incapability factors. The mechanisms responsible for the the observed tube degradation and failures are described. The majority of the leaks have been due fatigue in the U-bend of the Bruce 'A' steam generators. There have been very few failures attributed to corrosion of the three tube materials used in Ontario Hydro steam generators. Recent performance has been deteriorating primarily due to deposit accumulation in the steam generators. Plugging of the broached holes in the upper support plates at Bruce 'A' has caused some derating of two units. Increases have been observed in the primary heat transport system reactor inlet temperature of several units. These increases may be attributed to steam generator tube surface fouling. In addition, several units have accumulated deep, hard sludge piles on the tube sheet, although little damage been observed. Recently some fretting of tubes has been observed at BNGSB in the U-bend support region. Remedial measures are being taken to address the current problems. Solutions are being evaluated to reduce the generation of corrosion products in the feedtrain and their subsequent transport to the steam generators. (author)

  9. U.S. Advanced Materials Development Program for steam generators

    International Nuclear Information System (INIS)

    Patriarca, P.; Harkness, S.D.; Duke, J.M.

    1975-01-01

    The selection of construction materials for LMFBR steam generators is reviewed, presenting the advantages and limitations of 2 1 / 2 Cr-1 Mo steel selected for the Clinch River Breeder Reactor Plant. These limitations indicate that further development of high-strength ferritic steels containing 9 to 12 percent Cr and the high-nickel Alloy 800 could lead to superior materials, and programs to develop these materials have been started. Combustion Engineering has surveyed the experience with the high-strength ferritic steels and prepared ingots of 26 selected compositions. Charpy V-notch tests and metallography have been used to characterize these alloys, and optimum welding rod compositions for these alloys are under development. Westinghouse-Tampa is undertaking a program to gain code acceptance of Alloy 800. A program has been set up to provide the information required for design, justification, and fabrication of reliable components. Progress has been made on characterization, the role of tertiary creep in failure, and the development of welding processes. (U.S.)

  10. Prognostics for Steam Generator Tube Rupture using Markov Chain model

    International Nuclear Information System (INIS)

    Kim, Gibeom; Heo, Gyunyoung; Kim, Hyeonmin

    2016-01-01

    This paper will describe the prognostics method for evaluating and forecasting the ageing effect and demonstrate the procedure of prognostics for the Steam Generator Tube Rupture (SGTR) accident. Authors will propose the data-driven method so called MCMC (Markov Chain Monte Carlo) which is preferred to the physical-model method in terms of flexibility and availability. Degradation data is represented as growth of burst probability over time. Markov chain model is performed based on transition probability of state. And the state must be discrete variable. Therefore, burst probability that is continuous variable have to be changed into discrete variable to apply Markov chain model to the degradation data. The Markov chain model which is one of prognostics methods was described and the pilot demonstration for a SGTR accident was performed as a case study. The Markov chain model is strong since it is possible to be performed without physical models as long as enough data are available. However, in the case of the discrete Markov chain used in this study, there must be loss of information while the given data is discretized and assigned to the finite number of states. In this process, original information might not be reflected on prediction sufficiently. This should be noted as the limitation of discrete models. Now we will be studying on other prognostics methods such as GPM (General Path Model) which is also data-driven method as well as the particle filer which belongs to physical-model method and conducting comparison analysis

  11. Repair technique for steam generator tubes using electroforming

    International Nuclear Information System (INIS)

    Kim, Seung Ho; Jeong, Hyun Kyu; Seo, Moon Hong

    2001-07-01

    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

  12. Steam generator management at Ontario Hydro Nuclear Stations

    Energy Technology Data Exchange (ETDEWEB)

    Nickerson, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Maruska, C.C. [Ontario Hydro, Toronto, Ontario (Canada)

    1998-07-01

    Managing ageing steam generators involves costly decisions for the utility, both in terms of the cost of the maintenance activities andin terms of having the unit shutdown and consequent power loss while performing these activities. The benefits of these activities are seldom guaranteed and are sometimes very intangible. For nuclear utilities the most pertinent questions that arise are have we identified all the problem(s), can we predict the risk due to these problems? Can we implement corrective and preventive activities to manage the problem and what is the optimum timing of implementation? Is the money spent worthwhile, i.e. has it given us a return in production and safety? Can we avoid surprises? How can we tangibly measure success? This paper touches briefly on all the questions mentioned above but it mainly addresses the last question: 'how can we tangibly measure success?' by using several success indicators proposed by EPRI and by applying them to actual Ontario Hydro experience. The appropriateness of these success indicators as the means to assess the success of these programs, to feed back the results, and to enhance or revise the programs will be discussed. (author)

  13. Corrosion fracture of bolts of nuclear power steam generators

    International Nuclear Information System (INIS)

    Hrivnak, I.

    1990-01-01

    Bolts connecting collector bodies with lids were the first components of steam generators at Czechoslovak WWER-440 units on which corrosion damage was observed in 1982 to 1983. Corrosion cracks developed particularly in the cylindrical parts of the bolts. This was due to intergranular corrosion caused by the unsuitable chemical composition of the steel used, by secondary water level fluctuations, by the surrounding environment of the bolts being unamenable to deaeration, as well as by inappropriate tightening of the bolts which gave rise to additional deformation stress. Steps were taken to eliminate all these drawbacks, and owing to this the corrosion cracking of the bolts was prevented for longer than 6 years. Cracks were observed again in 1989; they occurred then not only in the cylindrical parts but also in the thread parts of the bolts. The corrosion was again of intergranular nature. As yet, the cause of the corrosion cracking of the bolts is not unambiguously known. It is largely assumed that the material used, viz. the high-strength KhN35VT nickel alloy, is exceedingly sensitive to the working procedure and to stress. (Z.M.)

  14. Quantification of the safety status of PWR steam generators

    International Nuclear Information System (INIS)

    Bartonicek, J.; Jonas, O.; Schoeckle, F.

    1996-01-01

    The operational safety of components is affected during the period of operation (e.g., by loads and environmental effects). Additionally, changes in codes and standards can affect the safety margin of a component. In order to determine a quantitative status of the safety, the existing status of the component and the actual knowledge have to be regarded and evaluated. Differences (deficits) with respect to actual demands must be balanced by use of redundant measure to guarantee safety. The procedure is demonstrated using the steam generators of GKN 1 (German nuclear power plant--in operation since 1975) as an example. The existing status of the component regarding design and construction, material, load history (results of monitoring), the results of nondestructive testing and operational experience is determined and related to the actual regulatory and code compliance and to the state of the art in technics. Non-specified loads were measured in operation. Stress and fatigue analysis computations are performed on the base of realistic load patterns (results from monitoring). Thus the safety status can be quantified. Finally the GKN-concept of measures to guarantee the reliability and the safety of the component in future operation is discussed

  15. Sodium-water reaction studies for MONJU steam generators

    International Nuclear Information System (INIS)

    Hori, M.; Sato, M.; Nei, H.; Harasaki, T.; Hishida, M.; Saito, T.

    1975-01-01

    The R and D results of the PNC's sodium-water reaction project are reviewed. The purposes of the project with the specific object for each test rig and computer code are given. The test items which should be investigated for the safety evaluation of the MONJU steam generators are discussed, and the status of the PNC's work on each item is described. The results on the small-leak wastage measurement are shown and the improved experimental equations to predict the wastage rate from the leak rate and the sodium temperature are given. The preliminary results on the wastage of tube bundle in the intermediate leak range are shown. The depth and the area of the wastage and also the wastage rate for each tube are shown graphically. The measured peak value of the initial pressure spike for the large leak is shown. The scatter of the data and its causes are discussed. The bubble growth rate estimated from the void probe measurement is presented. The results of the simulation experiment on the pressure wave propagation to the secondary circuit are given, comparing them with the prediction by the one-dimensional computer codes SWAC-5K and SWAC-5H. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  17. Characterization of crevice deposits in Ringhals 3 steam generators tubes

    International Nuclear Information System (INIS)

    Lancha, A.M.; Gomez-Briceno, D.; Garcia-Mazario, M.; Maffiotte, C.; Mcllree, A.

    1998-01-01

    With the aim of characterizing the TSP crevices and tube deposits two tubes, including tube/SP intersections, from a retired Ringhals 3 steam generator have been destructively examined at CIEMAT. The characterization has been performed physically (porosity and microhardness mainly) and chemically (elemental composition and distribution, compounds identification and oxide layers analyses). In addition, a thorough study of the morphology, thickness, coloration, and distribution of the deposits, in both axial and radial directions, in several locations around the tube has been carried out. The results show a heterogeneous distribution of the corrosion products inside the crevice area, relative to thickness, composition and porosity. On the other hand a gap, with values from 4 μm to 140 μm, has been detected between the tube and the deposits in many locations around the tube. This gap might correspond to the shrinkage of the tube during the cooldown. Therefore, during the shutdowns, in presence of air and stagnant water, the existing species in the deposits close to this gap may become oxidized. This fact could be particularly important in the case of copper because it is well known that copper oxides can increase the electrochemical potential up to values in which the SCC susceptibility of alloy 600 appears. (authors)

  18. System for installing a steam generator nozzle dam

    International Nuclear Information System (INIS)

    McDonald, F.X.; Weisel, E.M.; Schukei, G.E.

    1991-01-01

    This patent describes a system for installing a nozzle dam in a nuclear steam generator having a head including a head internal surface, a manway penetrating the head, and a nozzle penetrating the head. It comprises a manipulator adapted to be passed through the manway and having one end adapted to be attached remotely to the head internal surface and a free end including a clamp member; nozzle dam segments, each segment sized to pass through the manway and having means thereon for engaging at least one other segment, the segments when fully engaged to each other forming a dam subassembly sized to pass into and seat against the nozzle; and means for controlling the manipulator while the one end is attached to the head internal surface, such that the clamp member grasps and supports one of the dam segments within the head until the subassembly is formed within the head, and then translates the dam subassembly within the head until the dam subassembly seats within the nozzle

  19. Strain measurement on a compact nuclear reactor steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Scaldaferri, Denis Henrique Bianchi; Gomes, Paulo de Tarso Vida; Mansur, Tanius Rodrigues, E-mail: dhbs@cdtn.b, E-mail: gomespt@cdtn.b, E-mail: tanius@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Pozzo, Renato del, E-mail: delpozzo@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil); Mola, Jairo [Unitecnica Engenharia, Sao Paulo, SP (Brazil)

    2011-07-01

    This work presents the strain measurement procedures applied to a compact nuclear reactor steam generator, during a hydrostatic test, using strain gage technology. The test was divided in two steps: primary side test and secondary side test. In the primary side test twelve points for strain measurement using rectangular rosettes, three points (two external and one internal) for temperature measurement using special strain gages and one point for pressure measurement using a pressure transducer were monitored. In the secondary side test 18 points for strain measurement using rectangular rosettes, four points (two external and two internal) for temperature measurement using special strain gages and one point for pressure measurement using a pressure transducer were monitored. The measurement points on both internal and external pressurizer walls were established from pre-calculated stress distribution by means of numerical approach (finite elements modeling). Strain values using a quarter Wheatstone bridge circuit were obtained. Stress values, from experimental strain were determined, and to numerical calculation results were compared. (author)

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

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Jarmila Degmová

    2012-01-01

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

  2. Laboratory study of corrosion of steam generator tubes: Preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Sala, B.; Organista, M. [Centre Technique Framatome, Le Creusot (France). Dept. Chimie-Corrosion; Henry, K.; Erre, R. [CNRS-CRMD, Orleans (France); Gelpi, A. [Framatome, Paris la Defense (France). Dept. Materiaux et Technologies; Cattant, F.; Dupin, M. [EDF, Avoine (France)

    1995-12-31

    The secondary side intergranular attack (IGA) and intergranular stress corrosion cracking (IGSCC) of steam generator tubes often occurs in crevices where impurities are concentrated, due to local elevated temperatures and restricted water flow. From the analysis of tubes pulled from plants, it is believed that alumino-silicates deposits and/or organic species may play a role in the development of IGA in near neutral environments. New observations suggest that similar environments and similar processes are operative inside the corroded grain boundaries. A former study using autoclave tests was mainly devoted to the formation of alumino-silicate deposits similar to those observed in plants. The present work pursued the study of local environments responsible for IGA/SC. It confirms former results on the catalytic decomposition of organic species into acetates and presents more details on the mechanism of formation of alumino-silicate deposits on alloy 600, particularly on the role of iron and, to a lesser extent, nickel cations. It was showed that, under the alumino-silicate deposits and in the presence of some organic species, a non-protective chromium rich layer may grow instead of the usual protective spinel oxide. The mechanism responsible for the formation of this layer is believed to involve interaction between iron and, to a lesser extent, nickel with silica and/or possible interaction between chromium and acetates. Preliminary capsule tests indicate that these conditions may induce the initiation of IGA.

  3. Functionalized Graphene Enables Highly Efficient Solar Thermal Steam Generation.

    Science.gov (United States)

    Yang, Junlong; Pang, Yunsong; Huang, Weixin; Shaw, Scott K; Schiffbauer, Jarrod; Pillers, Michelle Anne; Mu, Xin; Luo, Shirui; Zhang, Teng; Huang, Yajiang; Li, Guangxian; Ptasinska, Sylwia; Lieberman, Marya; Luo, Tengfei

    2017-06-27

    The ability to efficiently utilize solar thermal energy to enable liquid-to-vapor phase transition has great technological implications for a wide variety of applications, such as water treatment and chemical fractionation. Here, we demonstrate that functionalizing graphene using hydrophilic groups can greatly enhance the solar thermal steam generation efficiency. Our results show that specially functionalized graphene can improve the overall solar-to-vapor efficiency from 38% to 48% at one sun conditions compared to chemically reduced graphene oxide. Our experiments show that such an improvement is a surface effect mainly attributed to the more hydrophilic feature of functionalized graphene, which influences the water meniscus profile at the vapor-liquid interface due to capillary effect. This will lead to thinner water films close to the three-phase contact line, where the water surface temperature is higher since the resistance of thinner water film is smaller, leading to more efficient evaporation. This strategy of functionalizing graphene to make it more hydrophilic can be potentially integrated with the existing macroscopic heat isolation strategies to further improve the overall solar-to-vapor conversion efficiency.

  4. Application of modeling to local chemistry in PWR steam generators

    International Nuclear Information System (INIS)

    Fauchon, C.; Millett, P.J.; Ollar, P.

    1998-01-01

    Localized corrosion of the SG tubes and other components is due to the presence of an aggressive environment in local crevices and occluded regions. In crevices and on vertical and horizontal tube surfaces, corrosion products and particulate matter can accumulate in the form of porous deposits. The SG water contains impurities at extremely low levels (ppb). Low levels of non-volatile impurities, however, can be efficiently concentrated in crevices and sludge piles by a thermal hydraulic mechanism. The temperature gradient across the SG tube coupled with local flow starvation, produces local boiling in the sludge and crevices. Since mass transfer processes are inhibited in these geometries, the residual liquid becomes enriched in many of the species present in the SG water. The resulting concentrated solutions have been shown to be aggressive and can corrode the SG materials. This corrosion may occur under various conditions which result in different types of attack such as pitting, stress corrosion cracking, wastage and denting. A major goal of EPRI's research program has been the development of models of the concentration process and the resulting chemistry. An improved understanding should eventually allow utilities to reduce or eliminate the corrosion by the appropriate manipulation of the steam generator water chemistry and or crevice conditions. The application of these models to experimental data obtained for prototypical SG tube support crevices is described in this paper. The models adequately describe the key features of the experimental data allowing extrapolations to be made to plant conditions. (author)

  5. Lessons learned from tubes pulled from French steam generators

    International Nuclear Information System (INIS)

    Berge, Ph.; Boursier, J.M.; Dallery, D.; De Keroulas, F.; Rouillon, Y.

    1998-01-01

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

  6. Steam generator management at Ontario Hydro Nuclear Stations

    International Nuclear Information System (INIS)

    Nickerson, J.; Maruska, C.C.

    1998-01-01

    Managing ageing steam generators involves costly decisions for the utility, both in terms of the cost of the maintenance activities and in terms of having the unit shutdown and consequent power loss while performing these activities. The benefits of these activities are seldom guaranteed and are sometimes very intangible. For nuclear utilities the most pertinent questions that arise are have we identified all the problem(s), can we predict the risk due to these problems? Can we implement corrective and preventive activities to manage the problem and what is the optimum timing of implementation? Is the money spent worthwhile, i.e. has it given us a return in production and safety? Can we avoid surprises? How can we tangibly measure success? This paper touches briefly on all the questions mentioned above but it mainly addresses the last question: 'how can we tangibly measure success?' by using several success indicators proposed by EPRI and by applying them to actual Ontario Hydro experience. The appropriateness of these success indicators as the means to assess the success of these programs, to feed back the results, and to enhance or revise the programs will be discussed. (author)

  7. Trip report: United States LMFBR Steam Generator Team. IAEA symposium, Bensberg, Germany, October 14--17, 1974

    International Nuclear Information System (INIS)

    1974-01-01

    Information is presented concerning steam generator design characteristics for the AFR reactor, SNR reactor, PHENIX reactor, SUPER PHENIX reactor, MONJU reactor, and BN-350 reactor; steam generator development programs for West Germany, France, Japan, U. K., and the U. S. S. R.; and the fabrication and inspection of steam generator components. Steam generator performance and maintenance requirements for operating LMFBR reactors are reviewed. (U.S.)

  8. PWR steam generator chemical cleaning. Phase I: solvent and process development. Volume II

    International Nuclear Information System (INIS)

    Larrick, A.P.; Paasch, R.A.; Hall, T.M.; Schneidmiller, D.

    1979-01-01

    A program to demonstrate chemical cleaning methods for removing magnetite corrosion products from the annuli between steam generator tubes and the tube support plates in vertical U-tube steam generators is described. These corrosion products have caused steam generator tube ''denting'' and in some cases have caused tube failures and support plate cracking in several PWR generating plants. Laboratory studies were performed to develop a chemical cleaning solvent and application process for demonstration cleaning of the Indian Point Unit 2 steam generators. The chemical cleaning solvent and application process were successfully pilot-tested by cleaning the secondary side of one of the Indian Point Unit 1 steam generators. Although the Indian Point Unit 1 steam generators do not have a tube denting problem, the pilot test provided for testing of the solvent and process using much of the same equipment and facilities that would be used for the Indian Point Unit 2 demonstration cleaning. The chemical solvent selected for the pilot test was an inhibited 3% citric acid-3% ascorbic acid solution. The application process, injection into the steam generator through the boiler blowdown system and agitation by nitrogen sparging, was tested in a nuclear environment and with corrosion products formed during years of steam generator operation at power. The test demonstrated that the magnetite corrosion products in simulated tube-to-tube support plate annuli can be removed by chemical cleaning; that corrosion resulting from the cleaning is not excessive; and that steam generator cleaning can be accomplished with acceptable levels of radiation exposure to personnel

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Hoch, G.

    1999-01-01

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

  12. Macroporous Double-Network Hydrogel for High-Efficiency Solar Steam Generation Under 1 sun Illumination.

    Science.gov (United States)

    Yin, Xiangyu; Zhang, Yue; Guo, Qiuquan; Cai, Xiaobing; Xiao, Junfeng; Ding, Zhifeng; Yang, Jun

    2018-04-04

    Solar steam generation is one of the most promising solar-energy-harvesting technologies to address the issue of water shortage. Despite intensive efforts to develop high-efficiency solar steam generation devices, challenges remain in terms of the relatively low solar thermal efficiency, complicated fabrications, high cost, and difficulty in scaling up. Herein, a double-network hydrogel with a porous structure (p-PEGDA-PANi) is demonstrated for the first time as a flexible, recyclable, and efficient photothermal platform for low-cost and scalable solar steam generation. As a novel photothermal platform, the p-PEGDA-PANi involves all necessary properties of efficient broadband solar absorption, exceptional hydrophilicity, low heat conductivity, and porous structure for high-efficiency solar steam generation. As a result, the hydrogel-based solar steam generator exhibits a maximum solar thermal efficiency of 91.5% with an evaporation rate of 1.40 kg m -2 h -1 under 1 sun illumination, which is comparable to state-of-the-art solar steam generation devices. Furthermore, the good durability and environmental stability of the p-PEGDA-PANi hydrogel enables a convenient recycling and reusing process toward real-life applications. The present research not only provides a novel photothermal platform for solar energy harvest but also opens a new avenue for the application of the hydrogel materials in solar steam generation.

  13. Studying the processes of sodium-water interaction in the BOR-60 reactor micromodule steam generator

    International Nuclear Information System (INIS)

    Tsykanov, V.A.; Antipin, G.K.; Borisov, V.V.

    1981-01-01

    Main results of experimental studies of emergency regimes of micromodule steam generator (MSG) at small and big leaks of water into sodium, realized using the 30 MW MSG, operating in the BOR-o0 reactor, are considered. The aims of the study are as follows: the modelling of macroleak in ''Nadja'' steam generator for the BN-350 reactor; testing the conceptions of alarm signalling and MSG protection; testing under real conditions of new perspective systems of leak detection; gaining the experimence and development of the ways to eliminate the consequences of accident caused by big water leak into sodium; accumulation of knowledge on restoration of MSG operating ability after accident; experimental test of calculational techniques for big leak accidents to use them in future for calculational studies of similar situations at other reactors equipped with sodium-water steam generators; refinement of characteristics of hydrodynamic and thermal effects interaction zone for big leak in real circuit during the plant operation. A series of experiments with the imitation of water leak into sodium by means of argon and steam supply through injection devices, located before the steam superheater module of one of the sections and between evaporator module of the same section, is conducted. The range of steam flow rate is 0.02-0.45 g/s. Duration of steam supply is 100-400 s. A conclusion is made that the results obtained can be used for steam generator of the BN-350 reactor [ru

  14. Study on power generation performance of sea water evaporation steam

    Science.gov (United States)

    Zhao, Chenguang; Ma, Jinghuan; Liu, Ying; Shen, Dongfang

    2018-02-01

    Seawater with softening pretreatment was used as the medium for experiment evaporation, and the influence of influent water, influent water temperature and flue gas flow on the steam flow and evaporation rate is discussed in this paper. The results show that with the increase of water inflow, the evaporation rate decreases at the beginning and then increases. Steam flow increases with the increasing water inflow; the increase of influent temperature has great influence on the evaporation rate and steam flow, however, the evaporation rate hardly changes when the influent temperature reaches a certain value; Steam flow and evaporation rate increases with increasing flue gas flow, and increases slightly with increasing before reaching 90m3/h. The surface of the heating tube was analyzed by means of SEM and X-ray diffraction. These discussions and results has some guiding significance in designing evaporation equipment using softening pretreatment.

  15. Experimental and theoretical investigations on safety of the SNR - straight-tube design steam generator with sodium-water reactions

    International Nuclear Information System (INIS)

    Dumm, K.; Sauermann, F.; Schnitker, W.; Welter, A.

    A number of large sodium-water reaction tests has been performed in a steam generator model in order to verify the layout criteria of the SNR straight-tube design steam generators under accident conditions. The experimental setup is described. The test results and their applicability to the SNR steam generators are given and discussed. (U.S.)

  16. MHTGR steam generator on-line heat balance, instrumentation and function

    International Nuclear Information System (INIS)

    Klapka, R.E.; Howard, W.W.; Etzel, K.T.; Basol, M.; Karim, N.U.

    1991-09-01

    Instrumentation is used to measure the Modular High Temperature Gas-Cooled Reactor (MHTGR) steam generator dissimilar metal weld temperature during start-up testing. Additional instrumentation is used to determine an on-line heat balance which is maintained during the 40 year module life. In the process of calibrating the on-line heat balance, the helium flow is adjusted to yield the optimum boiling level in the steam generator relative to the dissimilar metal weld. After calibration is complete the weld temperature measurement is non longer required. The reduced boiling level range results in less restrictive steam generator design constraints

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

    International Nuclear Information System (INIS)

    Ferreira, W.J.

    1982-01-01

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

  18. Optimization of the steam generator project of a gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Sakai, Massao

    1978-01-01

    The present work is concerned with the modeling of the primary and secondary circuits of a gas cooled nuclear reactor in order to obtain the relation between the parameters of the two cycles and the steam generator performance. The procedure allows the optimization of the steam generator, through the maximization of the plant net power, and the application of the optimal control theory of dynamic systems. The heat balances for the primary and secondary circuits are carried out simultaneously with the optimized - design parameters of the steam generator, obtained using an iterative technique. (author)

  19. Measuring steam generator performance using non-intrusive downcomer flow measurements

    International Nuclear Information System (INIS)

    Taylor, C.E.; McGregor, J.E.; Kittmer, C.A.

    1998-01-01

    Nuclear plant reliability depends directly on steam generator performance. Downcomer flow is a good monitor of steam generator performance. It provides information critical to the efficient and safe operation of steam generators as determined by the recirculation ratio and water inventory. In addition, reduced downcomer flow may indicate steam generator crudding or inadequate chemical cleaning. This paper describes recent advances in the application of ultrasonic technology to measure flow velocity in the downcomer annulus during operation. This technique is non-intrusive since the measurements are taken with ultrasonic transducers mounted on the outer shell of the steam generator. New transducers and improved installation techniques have resulted in increased transducer reliability. Through on-site testing, it was determined that some CANDU steam generators are experiencing carry-under (steam from the separators is carried into the downcomer). To measure the downcomer flow under these conditions, a different ultrasonic technique was required. A new technique became available in 1995 and was successfully adapted for high-temperature application. This transflection method was attached to a Bruce A steam generator in January of 1996. Whereas previous installations provided data for two to three months, this installation was still operating when the reactors were shut down in 1997. Options for movable measuring systems and simpler surface preparation have also been examined. This research has determined several obstacles and some possibilities for the use of magnets in temporarily holding the transducers at a given location. This would allow for measurements to be taken in a larger number of locations using the same flow measurement system. In addition, the need for minor welding on the surface of the steam generator shell would be eliminated. (author)

  20. Results of the secondary side chemical cleaning of the steam generators

    International Nuclear Information System (INIS)

    Doma, A.; Patek, G.

    2001-01-01

    A significant amount of deposit has developed on the secondary side of the heat transfer tubes of the steam generators (SG) of the Paks Nuclear Power Plant units in course of the years. More than 99.5% of the deposit is made up of magnetite (Fe 3 O 4 ) generated in the secondary circuit of the power plant. Those deposits lead to the decrease of the heat transfer. Even more important is its role from the point of view of operational reliability of the steam generators, leak tightness between the primary and secondary sides. The first series of cleaning took place following 8-9 years of operation of the units. Following the first cleaning cycle the transport of the corrosion products into the steam generators did not change, and thus obviously new cleaning was required. Periodical cleaning of the steam generators shall be assured. (R.P.)

  1. Steam generator design for solar towers using solar salt as heat transfer fluid

    Science.gov (United States)

    González-Gómez, Pedro Ángel; Petrakopoulou, Fontina; Briongos, Javier Villa; Santana, Domingo

    2017-06-01

    Since the operation of a concentrating solar power plant depends on the intermittent character of solar energy, the steam generator is subject to daily start-ups, stops and load variations. Faster start-up and load changes increase the plant flexibility and the daily energy production. However, it involves high thermal stresses on thick-walled components. Continuous operational conditions may eventually lead to a material failure. For these reasons, it is important to evaluate the transient behavior of the proposed designs in order to assure the reliability. The aim of this work is to analyze different steam generator designs for solar power tower plants using molten salt as heat transfer fluid. A conceptual steam generator design is proposed and associated heat transfer areas and steam drum size are calculated. Then, dynamic models for the main parts of the steam generator are developed to represent its transient performance. A temperature change rate that ensures safe hot start-up conditions is studied for the molten salt. The thermal stress evolution on the steam drum is calculated as key component of the steam generator.

  2. Optimization of steam generator replacement with virtual reality modeling

    International Nuclear Information System (INIS)

    Kim, Jeong H.; Suh, Kune Y.

    2008-01-01

    Nuclear power plants (NPPs) have to be carefully examined and maintained up to the point of replacing major components during the overhaul period for continued operation. Most understandably the cost of maintenance and upgrading will tend to increase with the NPP power. There is thus an escalating need for developing an optimized process management method to reduce the cost involved. Albeit the steam generators (SGs) may not directly affect the expected lifespan of NPP, thousands of tubes with diameter on the order of 3 cm in the SG operating at 320degC and 16 MPa may well tend to be called Achilles' heel of the pressurized water reactors (PWRs). For instance, the SGs of Kori Nuclear Unit 1 (KNU 1) were replaced in October 1998 after 20 years of service on account of aging and potential threat to operational safety. In the same year the SG tubes of Ulchin Nuclear Units 1 and 2 were ruptured to result in leakage of the primary coolant to the secondary side. As a result their SGs are planned to be replaced in a few years. There is, however, a limit to improving the replacement process by trial and error in practice on account of the size of NPP with the ensuing complexity in process management. This paper proposes an optimization method for the SG replacement process based on the KNU 1 experience in 1998. The whole process was simulated accounting for interactions of each part in virtual reality utilizing the computer aided design solution CATIA, and the digital process management solution DELMIA. (author)

  3. Steam generator maintenance and life management at Embalse Nuclear Station

    International Nuclear Information System (INIS)

    Sainz, R.; Diaz, G.; Sveruga, H.; Ramakrishnan, T.K.; Azeez, S.

    2004-01-01

    The Embalse Nuclear Station has four steam generators (SGs) with inverted vertical U tubes manufactured by Babcock and Wilcox Canada (B and W). These are main components, both from the operative point of view as the heat transfer from the Primary Heat Transport System (PHTS) to the Secondary System, and from the point of view of safety, as they are the part of the PHTS and its radioactive inventory pressure barrier. In addition, they are one of the most important cost-related elements for potential life extensions. Maintenance and inspections are carried out in order to maintain a high availability of the SGs, as they have had a positive impact on the operational availability of the plant, and to reduce the tube failure probabilities, thus minimizing the amount of radioactive effluents and taking care of the condition of the main components in order to enable the plant life management and the planning of the plant life extension. The most relevant maintenance activities performed have been the inspections performed on 100% of the tubes every 3 years. the mechanical cleaning of the inside of the tubes, the sludge removal from the secondary side tubesheet, the divider plate replacement, and the inspection of internals of the secondary side.Thanks to the latter and to the eddy current inspections, the degradation in the U-bend supports was detected early and every effort is being made to repair them shortly. Besides, a life management program has been started covering the entire plant starting with this important component. The Embalse Nuclear Station's SGs show a low percentage of plugged tubes compared to other stations in similar conditions, but they must be monitored continually and systematically if a life extension is intended. (author)

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

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.H.; Subash, N.; Wright, M.D.

    2001-10-01

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

  5. Overview of steam generator tube-inspection technology

    International Nuclear Information System (INIS)

    Obrutsky, L.; Renaud, J.; Lakhan, R.

    2014-01-01

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

  6. Investigation of separation and hydrodynamic characteristics of steam generators used at the NPPs running on PWR-1000 reactors

    International Nuclear Information System (INIS)

    Ageev, A.G.; Korolkov, B.M.; Nigmatulin, B.I.; Vasileva, R.V.; Nekrasov, A.V.; Titiv, V.F.; Tarankov, G.A.

    1997-01-01

    The tests were accomplished at the steam generator of unit 5 of the Novovoronezh nuclear power plant. The outbursts of the steam-water mixture from the gap between the steam generator housing and the submerged perforated screen rim at the side of the inlet coolant manifold were investigated. Tests of the steam generator with a modified steam separation system were carried out on the Balakovo nuclear power plant. The gilled separator of the steam generator was replaced with a steam collecting perforated screen, while the gap between the steam generator housing and the heat exchange bundle rim was closed with additional perforated screens at the side of the inlet manifold. This new solution of moisture separation is better. (M.D.)

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

    International Nuclear Information System (INIS)

    Petrusek, R.A.

    1991-01-01

    This patent describes an improved method of draining down contained reactor-coolant water from the inverted vertical U-tubes of at least one vertical-type steam generator in which the upper inverted U-shaped ends of the tubes are closed and the lower ends thereof are open, the steam generator having a channel head at its lower end including a vertical dividing wall defining a primary water inlet side and a primary water outlet side of the generator, the steam generator having chemical volume control system means and residual heat removal system means, and the steam generator being part of a nuclear-powered steam generating system wherein the reactor-coolant water is normally circulated from and back into the reactor via a loop comprising the steam generator and inlet and outlet conduits connected to the lower end of the steam generator, and the reactor being in communication with pressurizer means and comprising the steps of introducing a gas which is inert to the system and which is under pressure above atmospheric pressure into at least one of the downwardly facing open ends of each of the U-tubes from below the tubesheet in which the open ends of the U-tubes are mounted adjacent the lower end of the steam generator while permitting the water to flow out from the open ends of the U-tubes, the improvement in combination therewith for substantially increasing the effectiveness and efficiency of such water removal from the tubes. It includes determining the parameters effecting a first average volumetric rate of removal for a predetermined period of time, infra, of the reactor-coolant water from the inverted vertical U-tubes, the specific unit for the first average volumetric rate expressing properties identical with the properties expressed in a second average volumetric rate maintained in a later mentioned step

  8. Detection of steam leaks into sodium in fast reactor steam generators by acoustic techniques - An overview of Indian programme

    International Nuclear Information System (INIS)

    Prabhakar, R.; Vyjayanthi, R.K.; Kale, R.D.

    1990-01-01

    Realising the potential of acoustic leak detection technique, an experimental programme was initiated a few years back at Indira Gandhi Centre for Atomic Research (IGCAR) to develop this technique. The first phase of this programme consists of experiments to measure background noise characteristics on the steam generator modules of the 40 MW (thermal) Fast Breeder Test Reactor (FBTR) at Kalpakkam and experiments to establish leak noise characteristics with the help of a leak simulation set up. By subjecting the measured data from these experiments to signal analysis techniques, a criterion for acoustic leak detection for FBTR steam generator will be evolved. Second phase of this programme will be devoted to developing an acoustic leak detection system suitable for installation in the 500 MWe Prototype Fast Breeder Reactor (PFBR). This paper discusses the first phase of the experimental programme, results obtained from measurements carried out on FBTR steam generators and results obtained from leak simulation experiments. Acoustic leak detection system being considered for PFBR is also briefly described. 4 refs, 8 figs, 1 tab

  9. Linking gas turbine with lignite fired steam generators: alternatives, limitations, benefits and costs

    Energy Technology Data Exchange (ETDEWEB)

    Stamatelopoulos, G.N.; Leithner, R.; Karakas, E.; Papageorgiou, N. [Technische Universitaet Braunschweig, Braunschweig (Germany). Inst. fuer Waerme und Breunnstofftechnik

    1996-09-01

    In this paper the possibility of linking high efficiency gas turbines with lignite fired steam generators is examined. The possible combinations of gas and steam turbine and the experience gained from their practical application are presented. The technical limitations set by the existing boilers and steam turbines are discussed. For the efficiency and power output increase of the power plant Kardia in Northern Greece the following alternatives are suggested: supplementary firing of oil or natural gas together with lignite; topping of every steam generator with small gas turbines and installation of a gas turbine with turbine with heat recovery steam generator and water/steam side link with every steam generator. The possibility of using the exhaust gas from the gas turbine as the only oxygen carrier is also discussed. These alternatives are calculated with a process simulation program and are compared to each other with emphasis on the technical and the economical aspects. A sensitivity analysis concerning the natural gas price was carried out in order to evaluate its influence on the costs and the affectivity of the investment. 16 refs., 7 figs., 4 tabs.

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

    International Nuclear Information System (INIS)

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

    1981-10-01

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

  11. Steam Generator tube plugging analysis of natural circulation conditions for NPP Krsko

    International Nuclear Information System (INIS)

    Bajs, T.; Mirkovic, D.

    1989-01-01

    Pump trip for NPP Krsko was analysed by deterministic approach. Analyses for 0% and 10% tube plugging were performed using computer code RELAP4/MOD6. The influence of steam generator tube plugging on natural circulation conditions is discussed. (author)

  12. Steam-generator dose rates on Westinghouse pressurized-water reactors. Interim report

    International Nuclear Information System (INIS)

    Vernon, C.W.

    1982-06-01

    Steam generator channel head radiation field data from nineteen Westinghouse PWR nuclear power plants were evaluated to define trends and determine factors which can influence these trends. The steam generator channel head data are the largest consistent set of radiation field data available at this time. Nine plants with measurements reported for a period of at least two effective full power years form the basis for the majority of the evaluations. All of the reported dose rates presented together yield no obvious trend. Although some individual steam generators show what may be the beginning of a trend, of a decreasing rate of radiation field buildup, not enough data is available to identify a trend. Comparisons were performed on dose rates buildup on the basis of several parameters. However, none of these parameters could be identified as a primary influence on steam generator channel head radiaton fields. 14 figures, 2 tables

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

    International Nuclear Information System (INIS)

    Matal, O.; Simo, T.; Kucak, L.; Urban, F.

    1997-01-01

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

  14. Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

    International Nuclear Information System (INIS)

    1996-10-01

    This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    2009-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Esch Markus

    2014-01-01

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

  19. Sodium-Water Reaction approach and mastering for ASTRID Steam Generator design

    International Nuclear Information System (INIS)

    Saez, Manuel; Allou, Alexandre; Beauchamp, François; Bertrand, Carole; Rodriguez, Gilles; Menou, Sylvain; Prele, Gérard

    2013-01-01

    Conclusions: • Modular Steam Generator concept selected for ASTRID: → Brings flexibility for the expertise of failed modules after their removal; → Intrinsically limit the mechanical consequences of a postulated large Sodium-Water Reaction. • Sodium-Water-Air Reaction studies include both prevention and mitigation aspects, with dedicated tools to be developed through R&D. • Regarding Safety analysis, the possibility to move from the scenario of instantaneous failure of the whole Steam Generator tube bundle toward a scenario with sequenced failure needs to be investigated. • The Steam Generator is one of the key components in the Sodium-cooled Fast Reactor system for it provides an interface between sodium and water. The design objective for the Steam Generator is related to the improvement of mastering of Sodium-Water Reaction. • Potential Sodium-Water Reactions can be eliminated by adopting a Gas based Power Conversion System

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  1. PWR steam generator chemical cleaning. Phase I: Final report, Volume I

    International Nuclear Information System (INIS)

    1978-07-01

    Two chemical cleaning solvent systems and two application methods were developed to remove the sludge in nuclear steam generators and to remove the corrosion products in the annuli between the steam generator tubes and the support plates. Laboratory testing plus subsequent pilot testing has demonstrated that, in a reasonable length of time, both solvents are capable of dissolving significant amounts of sludge, and of dissolving tightly packed magnetite in tube/support plate crevices. Further, tests have demonstrated that surface losses of the materials of construction in steam generators can be controlled to acceptable limits for the duration of the required cleaning period. Areas requiring further study and test have been identified, and a preliminary procedure for chemical cleaning nuclear steam generators has been chosen subject to quantification based on additional tests prior to actual in-plant demonstration

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  4. Steam generator tube rupture in an experimental facility scaled from a pressurized water reactor

    International Nuclear Information System (INIS)

    Loomis, G.G.

    1984-09-01

    Results from an experimental investigation of steam generator tube rupture in the Semiscale Mod-2B system are presented. From the experimental results, the characteristic system response signature for a wide range of number of tubes ruptured has been described. The tube rupture was assumed to occur during normal full power operation (15.6 MPa system pressure, 37 0 K core differential temperature). In addition, recovery scenarios involving operator actions were examined. The recovery scenarios included use of pressurizer auxiliary spray and internal heaters, steam generator feed and steam, primary feed and bleed, and main cooling pump operation. Recovery scenarios suggested by typical US pressurized water reactor emergency operating procedures were followed

  5. Analysis of density wave instability in counter-flow steam generators using STEAMFREQ-X

    International Nuclear Information System (INIS)

    Chan, K.C.; Yadigaroglu, G.

    1986-01-01

    The STEAMFREQ-X computer model was developed to provide a more comprehensive modeling of the different phenomena that are important to stability analysis of counter-flow steam generators. It uses a frequency-domain analysis and considers heat-flux/flow coupling between the primary and secondary fluids in space and time. Predictions by STEAMFREQ-X were compared with data from both a multi-channel liquid-sodium heated steam generator and a set of single pipe test data. Predicted outlet steam qualities at instability thresholds were within 15% of experimental data for all test points. (orig.)

  6. Mechanized inspection of steam generator components during manufacture

    International Nuclear Information System (INIS)

    Otte, H.-J.; Leupoldt, K.; Meister, W.

    2009-01-01

    Steam Generator (SG) parts are intensively inspected by UT in the course of the manufacturing process. These inspections - mostly performed manually using different codes - are time consuming and call for a sophisticated documentation, figuring part of the life time documentation package. In order to reduce time and costs mechanized inspection equipment is introduced, combining short inspection times, avoiding influence of the human factor and providing proper electronic storage of all inspection results prepared for comparison with data generated during in-service inspection. Since 2001 Cegelec delivered various UT systems for gas turbine disks and rotor ends called SIRO-MAN. Within only a few years the majority of important providers of such components successfully switched from manual inspection to mechanized inspection following the requirements of manufacturers like ALSTOM, GE and Siemens. The SIRO-MAN is now adapted to the needs of mechanized inspection of SG components. The inspection is performed on the products during rotation around the vertical axis. The multi - probe assemblies are manoeuvred on the products by a manipulator system backed by a NC control unit. Acoustic coupling of UT probes to the product surface is performed with oil or water in a closed circuit. UT and - if requested ET - data along with position information of the probe assembly provided by the control unit are acquired, processed and evaluated by an UT / ET electronic system delivered by either Olympus or ZETEC. As performed already on rotor ends a sequence of inspections using different parameter settings can be programmed with simple means (Teach In) so that such inspection sequence can be executed without operating personnel. Probe assemblies allow for individual operation of probes out of the probe assembly according to the individual needs. Conventional UT and phased array applications or combination of both techniques can be provided. The UT / ET electronic equipment offers

  7. Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

    International Nuclear Information System (INIS)

    1978-10-01

    The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  9. The analysis of thermal hydraulics in heater of steam generator secondary side

    International Nuclear Information System (INIS)

    Yang Zhilin; Li Shixing; Xu Ming

    1996-01-01

    COBRA-TF code is a multi-dimensional multi-field computer program, used for analysis of the thermal hydraulics in reactor system. According to the characteristics of steam generator in a nuclear power plant, the validation of using of COBRA-TF code to analyze the thermal hydraulics of the heater of the secondary side of steam generator has been made. The results are compared with those from ALBERTINE-2

  10. Dryout in sodium-heated helically-coiled steam generator tubes

    International Nuclear Information System (INIS)

    Tomita, Y.; Kosugi, T.; Kubota, J.; Nakajima, K.; Tsuchiya, T.

    1984-01-01

    Experimental research on the dryout phenomenon in sodium heated, helically coiled steam generator tubes was carried out. The fluctuation of the tube wall temperature caused by dryout was measured with thermocouples installed in the center of the tube wall. Empirical correlations of dryout quality were developed as functions of critical heat flux, water mass velocity and saturation pressure. These correlations confirmed that the design criterion of the MONJU steam generator was reasonable. (author)

  11. Visual inspection equipment for the secondary side of steam generators. Final report

    International Nuclear Information System (INIS)

    Hinton, M.S.; Chagnon, C.W.

    1981-05-01

    The development of miniature visual examination equipment capable of inspecting regions inside the tube bundles of nuclear steam generators to determine such factors as gap condition, tube outside surface condition, and material buildup is described. Tube-to-support plate interface regions at the tenth-support plate level in a once-through steam generator (Oconee 1) were visually examined. Other areas in the peripheral region were examined for conditions associated with available eddy-current and profilometer data

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    OpenAIRE

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  15. Water experiment on phased array acoustic leak detection system for sodium-heated steam generator

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Yoshiuji, Takahiro

    2015-01-01

    Highlights: • An acoustic leak detection system for sodium heated steam generator is proposed. • The new system can separate leak source from steam generator background noise. • Performance of the new system has been confirmed in water experiments. - Abstract: A phased array acoustic leak detection system for sodium heated steam generator has been proposed. The major advantage of the new system is it could provide information of acoustic source direction. An acoustic source of a sodium–water reaction is supposed to be localized while the background noise of the steam generator operation is uniformly distributed in the steam generator tube region. Therefore the new system could separate the target leak source from steam generator background noise. In the previous study, the methodology was proposed and basic performance was confirmed by numerical analysis. However, in the numerical analysis, acoustic transportation through the SG tube bundle was not modeled. In the present study, performance the proposed system has been confirmed in water experiments with mockup tube bundles

  16. Scale Thickness Measurement of Steam Generator Tubing Using Eddy Current Signal of Bobbin Coil

    International Nuclear Information System (INIS)

    Kim, Chang Soo; Um, Ki Soo; Kim, Jae Dong

    2012-01-01

    Steam generator is one of the major components of nuclear power plant and steam generator tubes are the pressure boundary between primary and secondary side, which makes them critical for nuclear safety. As the operating time of nuclear power plant increases, not only damage mechanisms but also scaled deposits on steam generator tubes are known to be problematic causing tube support flow hole blockage and heat fouling. The ability to assess the extent and location of scaled deposits on tubes became essential for management and maintenance of steam generator and eddy current bobbin data can be utilized to measure thickness of scale on tubes. In this paper, tube reference standards with various thickness of scaled deposit has been set up to provide information about the overall deposit condition of steam generator tubes, providing essential tool for steam generator management and maintenance to predict and prevent future damages. Also, methodology to automatically measure scale thickness on tubes has been developed and applied to field data to estimate overall scale amount.

  17. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

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

    International Nuclear Information System (INIS)

    Melikhov, V.; Melikhov, O.; Parfenov, Y.; Nerovnov, A.

    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 non soluble impurities and determine the efficiency and reliability of the steam generator operation. The 3D code for simulation of the three-dimensional steam-water flows in the steam generator could be effective tool for design and optimization of the horizontal steam generator. The results of the code calculations are determined mainly by the set of the correlations describing interaction of the steam-water mixture with the inner constructions of the SG and interfacial friction. The results obtained by 3D code STEG with the usage of the different interfacial friction correlations are presented and discussed in the paper. These results are compared with the experimental ones obtained at the experimental test facility PGV-1500 constructed for investigation of the processes in the horizontal steam generator

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

    International Nuclear Information System (INIS)

    Montero, Gisela

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  1. The impact of NPP Krsko steam generator tube plugging on minimum DNBR at nominal conditions

    International Nuclear Information System (INIS)

    Lajtman, S.

    1996-01-01

    Typically, steam generator tube plugging (SGTP) both decreases the reactor coolant system (RCS) flow rate and the heat transfer surface area of the steam generator. At a constant thermal power and vessel outlet temperature, as tube plugging increases, the vessel average temperature, vessel inlet temperature and steam generator secondary side steam pressure decrease. This paper presents the analysis of impact of SGTP on Minimum Departure from Nucleate Boiling Ratio (MDNBR) at NPP Krsko (NEK), using the Improved Thermal Design Procedure (ITDP), WRB-1 correlation, and COBRA-III-C computer code. No credit was given to high plugging percentage region power reduction resulting from turbine volumetric flow limitations. MDNBR is found to be decreasing with increasing plugging, but not under the limiting values. (author)

  2. Calculation model for 16N transit time in the secondary side of steam generators

    International Nuclear Information System (INIS)

    Liu Songyu; Xu Jijun; Xu Ming

    1998-01-01

    The 16 N transit time is essential to determine the leak-rate of steam generator tubes leaks with 16 N monitoring system, which is a new technique. A model was developed for calculation 16 N transit time in the secondary side of steam generators. According to the flow characters of secondary side fluid, the transit times divide into four sectors from tube sheet to the sensor on steam line. The model assumes that 16 N is moving as vapor phase in the secondary-side. So the model for vapor velocity distribution in tube bundle is presented in detail. The 16 N transit time calculation results of this model compare with these of EDF on steam generator of Qinshan NPP

  3. Hydrogen-oxygen steam generator applications for increasing the efficiency, maneuverability and reliability of power production

    Science.gov (United States)

    Schastlivtsev, A. I.; Borzenko, V. I.

    2017-11-01

    The comparative feasibility study of the energy storage technologies showed good applicability of hydrogen-oxygen steam generators (HOSG) based energy storage systems with large-scale hydrogen production. The developed scheme solutions for the use of HOSGs for thermal power (TPP) and nuclear power plants (NPP), and the feasibility analysis that have been carried out have shown that their use makes it possible to increase the maneuverability of steam turbines and provide backup power supply in the event of failure of the main steam generating equipment. The main design solutions for the integration of hydrogen-oxygen steam generators into the main power equipment of TPPs and NPPs, as well as their optimal operation modes, are considered.

  4. Steam Generator Analysis Tools and Modeling of Degradation Mechanisms

    International Nuclear Information System (INIS)

    Yetisir, M.; Pietralik, J.; Tapping, R.L.

    2004-01-01

    The degradation of steam generators (SGs) has a significant effect on nuclear heat transport system effectiveness and the lifetime and overall efficiency of a nuclear power plant. Hence, quantification of the effects of degradation mechanisms is an integral part of a SG degradation management strategy. Numerical analysis tools such as THIRST, a 3-dimensional (3D) thermal hydraulics code for recirculating SGs; SLUDGE, a 3D sludge prediction code; CHECWORKS a flow-accelerated corrosion prediction code for nuclear piping, PIPO-FE, a SG tube vibration code; and VIBIC and H3DMAP, 3D non-linear finite-element codes to predict SG tube fretting wear can be used to assess the impacts of various maintenance activities on SG thermal performance. These tools are also found to be invaluable at the design stage to influence the design by determining margins or by helping the designers minimize or avoid known degradation mechanisms. In this paper, the aforementioned numerical tools and their application to degradation mechanisms in CANDU recirculating SGs are described. In addition, the following degradation mechanisms are identified and their effect on SG thermal efficiency and lifetime are quantified: primary-side fouling, secondary-side fouling, fretting wear, and flow-accelerated corrosion (FAC). Primary-side tube inner diameter fouling has been a major contributor to SG thermal degradation. Using the results of thermalhydraulic analysis and field data, fouling margins are calculated. Individual effects of primary- and secondary-side fouling are separated through analyses, which allow station operators to decide what type of maintenance activity to perform and when to perform the maintenance activity. Prediction of the fretting-wear rate of tubes allows designers to decide on the number and locations of support plates and U-bend supports. The prediction of FAC rates for SG internals allows designers to select proper materials, and allows operators to adjust the SG maintenance

  5. Experience of steam generator replacement works in Kyushu Electric Power Co., Inc

    International Nuclear Information System (INIS)

    Dangami, Mamoru; Oshinden, Kimihisa

    1995-01-01

    Genkai No. 1 plant is the two-loop plant of 559 MW output, which started the operation in October, 1975. Grain boundary corrosion was found in the heating tubes of steam generators in the 5th periodic inspection in 1981. By repairing using plugs or sleeves, this damage does not become a problem regarding the safety or the performance, but in view of the reduction of radiation exposure and the improvement of the rate of operation and economy, it was decided to replace the steam generators. The replacement works were carried out during the period of 15th periodic inspection from May to November, 1994. The method of the replacement works was almost same as the preceding works in Takahama No. 2 plant of Kansai Electric Power Co., Inc. The actual schedule of the replacement works is shown. The basic procedure of the replacement works is the preparation works, taking all fuel out from the reactor, making the temporary opening on the reactor containment vessel, temporarily installing a crane, cutting main steam, main feedwater and primary coolant pipings, taking supporting structures off, carrying the old steam generators out, carrying new steam generators in and installing them, restoring support structures and pipings, removing the temporary crane, restoring the temporary opening of the reactor containment vessel and loading fuel. The steam generator was replaced in a lump. The methods of the above main processes, the dose equivalent of workers, and the amount of waste and the method of its disposal are reported. (K.I.)

  6. Condition monitoring of steam generator by estimating the overall heat transfer coefficient

    International Nuclear Information System (INIS)

    Furusawa, Hiroaki; Gofuku, Akio

    2013-01-01

    This study develops a technique for monitoring in on-line the state of the steam generator of the fast-breeder reactor (FBR) “Monju”. Because the FBR uses liquid sodium as coolant, it is necessary to handle liquid sodium with caution due to its chemical characteristics. The steam generator generates steam by the heat of secondary sodium coolant. The sodium-water reaction may happen if a pinhole or crack occurs at the thin metal tube wall that separates the secondary sodium coolant and water/steam. Therefore, it is very important to detect an anomaly of the wall of heat transfer tubes at an early stage. This study aims at developing an on-line condition monitoring technique of the steam generator by estimating overall heat transfer coefficient from process signals. This paper describes simplified mathematical models of superheater and evaporator to estimate the overall heat transfer coefficient and a technique to diagnose the state of the steam generator. The applicability of the technique is confirmed by several estimations using simulated process signals with artificial noises. The results of the estimations show that the developed technique can detect the occurrence of an anomaly. (author)

  7. Optimization of steam generators of NPP with WWER in operation with variable load

    Science.gov (United States)

    Parchevskii, V. M.; Shchederkina, T. E.; Gur’yanova, V. V.

    2017-11-01

    The report addresses the issue of the optimal water level in the horizontal steam generators of NPP with WWER. On the one hand, the level needs to be kept at the lower limit of the allowable range, as gravity separation, steam will have the least humidity and the turbine will operate with higher efficiency. On the other hand, the higher the level, the greater the supply of water in the steam generator, and therefore the higher the security level of the unit, because when accidents involving loss of cooling of the reactor core, the water in the steam generators, can be used for cooling. To quantitatively compare the damage from higher level to the benefit of improving the safety was assessed of the cost of one cubic meter of water in the steam generators, the formulated objective function of optimal levels control. This was used two-dimensional separation characteristics of steam generators. It is demonstrated that the security significantly shifts the optimal values of the levels toward the higher values, and this bias is greater the lower the load unit.

  8. Conceptual design of a bayonet-tube steam generator for the ALFRED lead-cooled reactor

    International Nuclear Information System (INIS)

    Damiani, Lorenzo; Montecucco, Massimo; Pini Prato, Alessandro

    2013-01-01

    Highlights: • Conceptual design of a bayonet-tube steam generator for a lead-cooled reactor demonstrator. • Steady-state simulations effected through RELAP 5. • Performance evaluation of the steam generator for different configurations of the bayonet tubes. - Abstract: The present paper is centred on the design of a bayonet tube steam generator, fundamental part of an innovative lead-cooled fast nuclear reactor (LFR). The construction of the LFR is the main objective of the European project named LEADER, of which Ansaldo Nucleare is an important member. The steam generator described in this paper is expected to be installed in a 300 MW thermal power demonstrator plant, named ALFRED. The investigations carried out in this work through the RELAP 5 code have first faced the sizing of the single bayonet tube and then the design of the whole heat exchanger. The configurations of the four coaxial tubes composing the single bayonet, the length of the bayonets and the materials employed have been investigated; the final heat exchanger configuration provides 510 bayonet tubes of 6 m active length with a thermal insulation between the inner descending tube and the rising annulus, assured by a special extremely insulating paint. The whole steam generator has shown its capability to reach the required exchanged power of 37.5 MW th , providing as output dry superheated steam at the desired temperature of 450 °C

  9. Rapid Generation of Superheated Steam Using a Water-containing Porous Material

    Science.gov (United States)

    Mori, Shoji; Okuyama, Kunito

    Heat treatment by superheated steam has been utilized in several industrial fields including sterilization, desiccation, and cooking. In particular, cooking by superheated steam is receiving increased attention because it has advantages of reducing the salt and fat contents in foods as well as suppressing the oxidation of vitamin C and fat. In this application, quick startup and cut-off responses are required. Most electrically energized steam generators require a relatively long time to generate superheated steam due to the large heat capacities of the water in container and of the heater. Zhao and Liao (2002) introduced a novel process for rapid vaporization of subcooled liquid, in which a low-thermal-conductivity porous wick containing water is heated by a downward-facing grooved heating block in contact with the upper surface of the wick structure. They showed that saturated steam is generated within approximately 30 seconds from room-temperature water at a heat flux 41.2 kW⁄m2. In order to quickly generate superheated steam of approximately 300°C, which is required for cooking, the heat capacity of the heater should be as small as possible and the imposed heat flux should be so high enough that the porous wick is able to dry out in the vicinity of the contact with the heater and that the resulting heater temperature becomes much higher than the saturation temperature. The present paper proposes a simple structured generator to quickly produce superheated steam. Only a fine wire heater is contacted spirally on the inside wall in a hollow porous material. The start-up, cut-off responses and the rate of energy conversion for input power are investigated experimentally. Superheated steam of 300°C is produced in approximately 19 seconds from room-temperature water for an input power of 300 W. The maximum rate of energy conversion in the steady state is approximately 0.9.

  10. Efficient running of steam generator trims fuel cost

    International Nuclear Information System (INIS)

    Selim, M.; Eltouny, S.A.

    1993-01-01

    E scaling energy prices have led to drastic changes in the operating philosophy of the worldwide industry. About 50% of the thermal energy in industry is being consumed in steam boilers. The new energy reduction programs that have been adopted in egypt draw attention to the boilers, not only to trim energy consumption and improve the production of steam but also to save as much money as possible in doing it. Organization for energy planning (OEP) has started a program for 'Boiler efficiency improvement' in industry since 2 years. The program aimed at performing energy audits in a selective number of industries in both public and private sectors using fire tube boilers produced locally by El Nasr company. As a result of audits an evaluation of performance of this type of boilers was,performed. The energy profiles and the common problems affecting the efficiency of boilers were determined. Energy conservation opportunities (ECO) were identified. 9 figs

  11. Chemical cleaning as an essential part of steam generator asset management

    International Nuclear Information System (INIS)

    Stiepani, C.; Ammann, F.; Jones, D.; Evans, S.; Harper, K.

    2010-01-01

    Accumulation of deposits is intrinsic for the operation of Steam Generators in PWRs. Such depositions often lead to reduction of thermal performance, loss of component integrity and, in some cases to power restrictions. Accordingly removal of such deposits is an essential part of the asset management of the Steam Generators in a Nuclear Power Plant. Every plant has its individual condition, history and constraints which need to be considered when planning and performing a chemical cleaning. Typical points are: Sludge load amount and constitution of the deposits; Sludge distribution in the steam generator; Existing or expected corrosion problems; Amount and treatment possibilities for the waste generated. Depending on these points the strategy for chemical cleaning shall be evolved. The range of treatment starts with very soft cleanings with a removal of approx 100 kg per steam generator and goes to a full scale cleaning which can remove up to several thousand kilograms of deposits from a steam generator. Depending on the goal to be achieved and the steam generator present an adequate cleaning method shall be selected. Flexible and 'customizable' cleaning methods that can be adapted to the individual needs of a plant are therefore a must. Particular for the application of preventive cleanings where repeated or even regular application are intended, special focus has to be put on low corrosion and easy waste handling. Therefore AREVA has developed the 'C3' concept, Customized Chemical Cleaning concept. This concept covers the entire range of steam generator cleaning. Particular for the preventive maintenance cleanings processes with extreme low corrosion rates and easy waste handling are provided which make repeated applications safe and cost efficient. (author)

  12. Fail-safety of the EBR-II steam generator system

    International Nuclear Information System (INIS)

    Chopra, P.S.; Stone, C.C.; Hutter, E.; Barney, W.K.; Staker, R.G.

    1976-01-01

    Fail-safe analyses of the EBR-II steam-generator system show that a postulated non-instantaneous leak of water or steam into sodium, through a duplex tube or a tubesheet, at credible leak rates will not structurally damage the evaporators and superheaters. However, contamination of the system and possible shell wastage by sodium-water reaction products may render the system inoperable for a period exceeding six months. This period would be shortened to three months if the system were modified by adding a remotely operated water dump system, a steam vent system, a secondary sodium superheater relief line, and a tubesheet leak-detection system

  13. EPR: steam generator tube rupture analysis in Finland and in France

    International Nuclear Information System (INIS)

    Israel, S.

    2006-01-01

    Different requirements between Finland and France lead EPR designer to define different features (system or action) for management of accidents on Olkiluoto 3 EPR that is under-construction in Finland compared to Flamanville 3 EPR that is foreseen in France. One of these differences concerns the management of Steam Generator Tube Rupture since no primary coolant (liquid and steam) release to the environment is allowed in Finland dislike in France where primary steam releases are not forbidden. This leads to define on Finnish EPR a strategy that anticipates mitigation action compared to French EPR and that only uses the unaffected steam generators. This strategy is intended to reduce the release to the environment. IRSN has analysed an other aspect of the Steam Generator Tube Rupture: the back-flow (flow of un-borated water from steam generator to the primary circuit). Indeed, if the Reactor Coolant Pumps have been shut down, the creation an un-borated water plug because of the back-flow could lead to reactivity accident in case of Reactor Coolant Pump restart. IRSN analysis shows that, using the current Olkiluoto 3 SGTR mitigation strategy and very penalizing assumptions, the amount of un-borated water transferred to the primary circuit on the Finnish EPR could be higher than on the French EPR in the long term. Discussions are going on between STUK and TVO to finalize the SGTR strategy so that both releases into the environment and risk of back-flow can be minimized. (author)

  14. Simulation of a Standalone, Portable Steam Generator Driven by a Solar Concentrator

    Directory of Open Access Journals (Sweden)

    Mohamed Sabry

    2015-05-01

    Full Text Available Solar energy is a good solution for energy-deficiency problems, especially in regions such ‎as rural areas in the Middle East that have not been electrified yet or are ‎under electrification. In ‎this paper, with the aid of a Computational Fluid Dynamics simulation, we propose a ‎system that comprises a trough solar concentrator and a pipe—with flowing water—that ‎is set in the concentrator focus. The aim of this work is to investigate the feasibility of generating steam ‎from such a system as well as analyzing the generated steam quantitatively ‎and qualitatively. Effects of variation of solar radiation intensity, ambient temperature, water ‎flow rate and pipe diameter on the quantity and quality of the generated steam have been investigated. The results ‎show that a quantity of about 130 kg of steam could be generated per day with a 0.01 m diameter with 0.0042 kg/s flowing water, although qualitatively, a narrower pipe achieves better performance than a wider one. About 74 kg of daily accumulated steam mass with a temperature >423 K could be achieved for a 0.005 m diameter tube compared to about 50 kg for the 0.01 m diameter tube. Steam quality factor is higher at all flow rates for the 0.005 m diameter tube compared to that of 0.01 m.

  15. SCC susceptibility and flaw tolerance evaluation for steam generator channel head materials

    International Nuclear Information System (INIS)

    Cothron, H.; Wolfe, R.

    2015-01-01

    Primary water stress corrosion cracking (PWSCC) has been reported in the divider plate assemblies of steam generators in operation outside of the United States. Evaluations have been performed to assess the susceptibility of U.S. steam generator channel head materials to PWSCC and the flaw tolerance of the channel head assembly in instances where cracking is assumed to occur. Earlier work concluded that the cracks reported in the foreign steam generators could not cause failure of the divider plate in the limiting U.S. steam generators during the design basis accident or normal operating conditions. Three additional cracking scenarios represent a potential breach of the primary pressure boundary of the channel head assembly: cracks initiating in the divider plate then propagating through the channel head cladding and into the low alloy steel shell material, cracks initiating in the tubesheet cladding then propagating to the tube-to-tubesheet weldments, and cracks initiating in the tube-to-tubesheet weld. Operating experience and literature were reviewed to determine the likelihood that cracks will propagate into the carbon steel channel head material and cause a breach in the primary pressure boundary. A flaw tolerance evaluation demonstrated that the structural integrity of the steam generator channel head is not compromised by a crack originating in the divider plate. Assumed axial and circumferential flaws in the steam generator channel head material remain well below the allowable flaw depths after 40 years of operation. Utilities can use the results from this analysis to determine the need to inspect steam generator Alloy 600 material in the divider plate assembly or the low-alloy steel channel head material in the period of operation beyond 40 years. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  17. Ultrasonic wall thickness gauging for ferritic steam generator tubing as an in-service inspection tool

    International Nuclear Information System (INIS)

    Haesen, W.M.J.; Tromp, Th.J.

    1980-01-01

    In-service inspection of LWR steam generators is more or less a standard routine operation. The situation can be very different for LMFBRs. For the SNR 300 (Kalkar Power Station) the situation is different because the steam generators have ferritic tubing. The tube walls are comparatively thick, 2 to 4.5 mm. During inservice examinations the steam generators will be drained on both sides, however on the sodium side a sodium film will be present. Furthermore the SNR 300 will have two types of steam generator. A straight tube design and a helical coil design will be used. Both types consist of a evaporator and superheater. The steam generators are of course not radioactive. It is obvious that in this case the eddy current (EC) technique is not an enviable inservice inspection tool. Basically EC is a surface flaw detection technique. Only the saturation magnetisation method will improve the EC technique sufficiently for ferritic material. However the 'in bore examination' with the saturation technique was, in case of the SNR 300 steam generator tubing, considered impossible since the inner diameters are fairly small. Furthermore sodium traces may influence the EC method. Although multifrequency methods can solve this problem, EC is not considered as a useful tool for examining ferritic tubing. Another method is to employ the 'stray flux' method which is under development with the TNO organization in Holland. The EC and stray flux method do have one drawback, these methods do not detect gradual changes in wall thickness. Ultrasonic examinations will be used in the SNR 300 as the main inspection tool for the steam generators. In this paper the reasons why ultrasonic examination was selected are explained. The results of the development work on this subject are discussed

  18. Evaluation of EDTA based chemical formulations for the cleaning of monel-400 tubed steam generators

    International Nuclear Information System (INIS)

    Velmurugan, S.; Rufus, A.L.; Sathyaseelan, V.S.; Kumar, P.S.; Veena, S.N.; Srinivasan, M.P.; Narasimhan, S.V.

    1998-01-01

    The Steam Generator (SG) is an important component in any nuclear power plant which contributes significantly for the over all performance of the reactor. The failure of SG tubes occurs mainly by corrosion under accelerated conditions caused by fouling. There is continuous ingress of the corrosion products and ionic impurities from the condenser and feed train of the secondary heat transfer system. The corrosion products accumulate in the stagnant areas near the tube sheet, over the tube support plates and in the tube to tube support plate crevices. These accumulated deposits help to concentrate the aggressive impurities and induce a variety of corrosion processes affecting the structural materials and finally leading to failure of the SG tube. Scale forming impurities can deposit over the tube surfaces and result in reduction of heat transfer efficiency and over heating of the surfaces. Every effort is being made to control the transport of impurities to the steam generator. Increased blow down, installation of condensate polishers and use of all volatile amines have helped to reduce the corrosion product and ionic impurities input into the steam generators of PHWRs. Despite these efforts, failures of SG tubes in PHWRs have been reported. Hence, attempts are being made to develop chemical formulations to clean the deposits accumulated in the steam generators. The EPRI-SGOG chemical cleaning process has been tried with good success in steam generators of different designs including the steam generators of PHWRs. This paper discusses the work on the evaluation of EDTA based chemical cleaning formulations for monel-400 tubed steam generators of PHWRs. (author)

  19. Some engineering aspects of the steam generator system for the United States LMFBR demonstration plant

    International Nuclear Information System (INIS)

    Tippets, F.E.

    1975-01-01

    This paper describes the main design features of the steam generator system for the Clinch River Breeder Reactor Plant and the engineering approach being employed for some of the critical elements of this system, including in particular the sodium-steam/water boundary, the efforts to have this boundary be of highest integrity, and the system features to safely accommodate any failure of the boundary. (author)

  20. On the reliability of steam generator performance at nuclear power plants with WWER type reactors

    International Nuclear Information System (INIS)

    Styrikovich, M.A.; Margulova, T.Kh.

    1974-01-01

    The problem of ensuring reliable operation of steam generators in a nuclear power plant with a water-cooled, water-moderated reactor (WWER) was studied. At a nuclear power plant with a vertical steam generator (specifically, a Westinghouse product) the steam generator tubes were found to have been penetrated. Shutdown was due to corrosion disintegration of the austenitic stainless steel, type 18/8, used as pipe material for the heater surface. The corrosion was the result of the action of chlorine ions concentrated in the moisture contained in the iron oxide films deposited in low parts of the tube bundle, directly at the tube plate. Blowing through did not ensure complete removal of the film, and in some cases the construction features of the steam generator made removal of the film practically impossible. Replacement of type 18/8 stainless steel by other construction material, e.g., Inconel, did not give good results. To ensure reliable operation of vertical steam generators in domestic practice, the generators are designed without a low tube plate (a variant diagram of the vertical steam generator of such construction for the water-cooled, water-moderated reactor 1000 is presented). When low tube plates are used the film deposition is intolerable. For organization of a non-film regime a complex treatment of the feed water is used, in which the amount of complexion is calculated from the stoichmetric ratios with the composition of the feed water. It is noted that, if 100% condensate purification is used with complexon processing of the feed water to the generator, we can calculate the surface of the steam-generator heater without considering the outer placement on the tubes. In this the cost of the steam generator and all the nuclear power plants with WWER type reactors is decreased even with installation of a 100% condensate purification. It is concluded that only simultaneous solution of construction and water-regime problems will ensure relaible operation of

  1. On possibility of application of the parallel-mixed type coolant flow scheme to NPP steam generators linked with superheaters

    International Nuclear Information System (INIS)

    Malkis, V.A.; Lokshin, V.A.

    1983-01-01

    Optimum distribution of the coolant straight-through flow between the superheater, evaporator and economizer is determined and the parallel-mixed type flow scheme is compared with other schemes. The calculations are performed for the 250 MW(e) steam generator for the WWER-1000 reactor unit the inlet and outlet primary coolant temperature of which is 324 and 290 deg C, respectively, while the feed water and saturation temperatures are 220 and 278.5 deg C, respectively. The rated superheating temperature is 300 deg C. The comparison of different schemes has been performed according to the average temperature head value at the steam-generator under the condition of equality as well as essential difference in the heat transfer coefficients in certain steam-generator sections. The calculations have shown that the use of parallel-mixed type flow permits to essentially increase the temperature head of the steam generator. At a constant heat transfer coefficient in all steam generator sections the highest temperature head is reached. At relative flow rates in the steam generator, economizer and evaporator equal to 6, 8 and 86%, respectively. The superheated steam generator temperature head in this case by 12% exceeds the temperature head of the WWER-1000 reactor unit wet steam generator. In case of heat transfer coefficient reduction in the superheater by a factor of three, the choice of the primary coolant, optimum distribution permits to maintain the steam generator temperature head at the level of the WWER-1000 reactor unit wet-steam steam generator. The use of the parallel-mixed type flow scheme permits to design a steam generator of slightly superheated steam for the parameters of the WWER-1000 unit

  2. Steam generator performance improvements for integral small modular reactors

    Directory of Open Access Journals (Sweden)

    Muhammad Ilyas

    2017-12-01

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

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

    International Nuclear Information System (INIS)

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-07-01

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

  5. Wasteless combined aggregate-coal-fired steam-generator/melting-converter

    International Nuclear Information System (INIS)

    Pioro, L.S.; Pioro, I.L.

    2003-01-01

    A method of reprocessing coal sludge and ash into granulate for the building industry in a combined wasteless aggregate-steam-generator/melting-converter was developed and tested. The method involves melting sludge and ash from coal-fired steam-generators of power plants in a melting-converter installed under the steam-generator, with direct sludge drain from the steam generator combustion chamber. The direct drain of sludge into converter allows burnup of coal with high ash levels in the steam-generator without an additional source of ignition (natural gas, heating oil, etc.). Specific to the melting process is the use of a gas-air mixture with direct combustion inside a melt. This feature provides melt bubbling and helps to achieve maximum heat transfer from combustion products to the melt, to improve mixing, to increase rate of chemical reactions and to improve the conditions for burning the carbon residue from the sludge and ash. The 'gross' thermal efficiency of the combined aggregate is about 93% and the converter capacity is about 18 t of melt in 100 min. The experimental data for different aspects of the proposed method are presented. The effective ash/charging materials feeding system is also discussed. The reprocessed coal ash and sludge in the form of granules can be used as fillers for concretes and as additives in the production of cement, bricks and other building materials

  6. Non-intrusive downcomer flow measurements: a means of monitoring steam generator performance

    International Nuclear Information System (INIS)

    Taylor, C.E.; McGregor, J.E.; Kittmer, C.A.; Pettigrew, M.J.; Jelinski, E.F.; Seppala, D.A.

    1995-01-01

    Nuclear plant reliability depends directly on steam generator performance. Downcomer flow is a good monitor of steam generator performance. It provides information critical to the efficient and safe operation of steam generators as determined by the recirculation ratio and water inventory. In addition, reduced downcomer flow may indicate steam generator crudding or inadequate chemical cleaning. This paper describes the application of ultrasonic technology to measure flow velocity in the downcomer annulus during operation. The technique is non-intrusive since the measurements are taken with ultrasonic transducers mounted on the outer shell of the steam generator. Successful application of this technique required development in several areas - high temperature couplants, signal quality, transducer performance and reliability, and remote monitoring. The effects of carry under, obstacles in the downcomer annulus, temperature variation, and wall thickness are also discussed in this paper. The results of measurements from 0 to 1 00% power in the Darlington nuclear station are presented. The results are compared to thermalhydraulic calculations. A second ultrasonic technique has recently been successfully tested at operating conditions with void in the flow. This new technique is also presented in this paper. (author)

  7. Dynamic heat transfer performance study of steam generator based on distributed parameter method

    International Nuclear Information System (INIS)

    Zhang, Guolei; Zhang, Yu; Yang, Yuanlong; Li, Yanjun; Sun, Baozhi

    2014-01-01

    Highlights: • One-dimensional mathematical model is built based on the distributed parameter method. • Dynamic simulation program is applied based on MATLAB using Runge–Kutta method. • The variations of primary and secondary parameters with power and space is discussed. • The highest temperature positions for the u-tube inner and outer wall are obtained. - Abstract: Using the steam generator of Daya Bay nuclear power plant as prototype, a one-dimensional dynamic mathematical model of nuclear-powered steam generator is built addressing the primary side fluid, the secondary side fluid and the inner and outer walls of the u-tubes based on distributed parameter method and reasonable assumptions. A dynamic simulation program is developed based on MATLAB using Runge–Kutta method and dynamic heat transfer performance simulation of steam generator is conducted under varying power. The calculation results show that the outlet temperature of primary side, the vapor saturation temperature and the mass fraction of secondary side agree with actual operating data of Daya Bay Nuclear Power Plant. Outer wall temperature at interface between parallel flow preheating-section and boiling-section is the highest. It provides a theoretical basis for the analysis of steam generator actual operating condition to build a one-dimensional mathematical model of steam generator based on the distributed parameter method and apply in simulation successfully

  8. Lightweight, Mesoporous, and Highly Absorptive All-Nanofiber Aerogel for Efficient Solar Steam Generation.

    Science.gov (United States)

    Jiang, Feng; Liu, He; Li, Yiju; Kuang, Yudi; Xu, Xu; Chen, Chaoji; Huang, Hao; Jia, Chao; Zhao, Xinpeng; Hitz, Emily; Zhou, Yubing; Yang, Ronggui; Cui, Lifeng; Hu, Liangbing

    2018-01-10

    The global fresh water shortage has driven enormous endeavors in seawater desalination and wastewater purification; among these, solar steam generation is effective in extracting fresh water by efficient utilization of naturally abundant solar energy. For solar steam generation, the primary focus is to design new materials that are biodegradable, sustainable, of low cost, and have high solar steam generation efficiency. Here, we designed a bilayer aerogel structure employing naturally abundant cellulose nanofibrils (CNFs) as basic building blocks to achieve sustainability and biodegradability as well as employing a carbon nanotube (CNT) layer for efficient solar utilization with over 97.5% of light absorbance from 300 to 1200 nm wavelength. The ultralow density (0.0096 g/cm 3 ) of the aerogel ensures that minimal material is required, reducing the production cost while at the same time satisfying the water transport and thermal-insulation requirements due to its highly porous structure (99.4% porosity). Owing to its rationally designed structure and thermal-regulation performance, the bilayer CNF-CNT aerogel exhibits a high solar-energy conversion efficiency of 76.3% and 1.11 kg m -2 h -1 at 1 kW m -2 (1 Sun) solar irradiation, comparable or even higher than most of the reported solar steam generation devices. Therefore, the all-nanofiber aerogel presents a new route for designing biodegradable, sustainable, and scalable solar steam generation devices with superb performance.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  10. Technical basis for the CANDU steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

    Kozluk, M.J.; Scarth, D.A.; Graham, D.B.

    2002-01-01

    Active degradation mechanisms in steam generators and preheaters in Canadian CANDU T M generating stations are managed through Steam Generator Programs that incorporate tube inspection, maintenance (cleaning), fitness-for-service assessment, and preventative plugging as part of the overall steam generator management strategy. Steam generator and preheater tubes are inspected in accordance with the CSA Standard CAN/CSA-N285.4-94[l]. When a detected flaw indication does not satisfy the criteria of acceptance by examination, CSA-N285.4-94 permits a fitness-for-service assessment to determine acceptability. In 1999 Ontario Power Generation issued, for trial use, fitness-for-service guidelines for steam generator and preheater tubes in CANDU nuclear power plants. The main objectives of the Fitness-for-Service Guidelines are to provide reasonable assurance that tube structural integrity is maintained, and to provide reasonable assurance that there are adequate margins between estimated accumulated dose and applicable site dose limits. The Fitness-for-Service Guidelines are intended to provide industry-standard acceptance criteria and evaluation procedures for assessing the condition of steam generator and preheater tubes in terms of tube structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. This paper describes the technical basis for the minimum required safety factors specified in Table IC-1 of the Fitness-for-Service Guidelines and for the flaw models used to develop the flaw stability requirements in the nonmandatory, Appendix C of the Fitness-for-Service Guidelines. (author)

  11. Next Generation Engineered Materials for Ultra Supercritical Steam Turbines

    Energy Technology Data Exchange (ETDEWEB)

    Douglas Arrell

    2006-05-31

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

  12. Technical and economic feasibility of solar augmented process steam generation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1976-01-01

    A study of the technical and economic feasibility of solar augmented process steam generation was performed. This approach is analogous to a heat pump that extracts heat from a low temperature reservoir (provided by solar energy) and raises its temperature to a useful level via mechanical work. The shaft power required in the compressor is only one third to one quarter of total steam enthalpy for low pressure process steam (100 psig). This approach permits the use of low cost flat plate collectors. It was concluded that these systems have the potential of yielding payback periods of 5 to 8 years and 10 to 15 years for collector costs of $2/ft/sup 2/ and $5/ft/sup 2/, respectively, depending upon the location. A design study of various components indicated that these components are generally available or need only minor modifications for steam service. The component selection was largely a function of steam generation rate. In general, collector cost was the controlling factor. It was also concluded that additional incentives are probably required for increased utilization of solar energy for industrial process steam.

  13. Desain Sistem Pendeteksi untuk Citra Base Sub-assembly dengan Algoritma Backpropagation

    Directory of Open Access Journals (Sweden)

    Kasdianto Kasdianto

    2017-04-01

    Full Text Available Object identification technique using machine vision has been implemented in industrial of electronic manufacturers for years. This technique is commonly used for reject detection (for disqualified product based on existing standard or defect detection. This research aims to build a reject detector of sub-assembly condition which is differed by two conditions that are missing screw and wrong position screw using neural network backpropagation. The image taken using camera will be converted into grayscale before it is processed in backpropagation methods to generate a weight value. The experiment result shows that the network architecture with two layers has the most excellent accuracy level. Using learning rate of 0.5, target error 0.015%, and the number of node 1 of 100 and node 2 of 50, the successive rate for sub-assembly detection in right condition reached 99.02% while no error occurs in detecting the wrong condition of Sub-assembly (missing screw and wrong position screw.

  14. Support and tool displacement device for the attachment of a tube bundle on a tubular plate of a steam generator

    International Nuclear Information System (INIS)

    Morisot, M.; Werle, R.; Michaud, J.P.

    1983-01-01

    The steam generator is being assembled, disposed with its axis horizontal and its tubular plate vertical; the device described in this patent, allows to automatize the preparation stages of the tubular plate and the attachment of the bundle, to shorten the construction of the steam generator and to remove drudgeries done by hand on the tubular plate or the tubes of the bundle. The invention can be applied to the construction of PWR steam generators [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    International Nuclear Information System (INIS)

    Souza, A.L. de.

    1981-07-01

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

  17. Heat transfer efficient thermal energy storage for steam generation

    International Nuclear Information System (INIS)

    Adinberg, R.; Zvegilsky, D.; Epstein, M.

    2010-01-01

    A novel reflux heat transfer storage (RHTS) concept for producing high-temperature superheated steam in the temperature range 350-400 deg. C was developed and tested. The thermal storage medium is a metallic substance, Zinc-Tin alloy, which serves as the phase change material (PCM). A high-temperature heat transfer fluid (HTF) is added to the storage medium in order to enhance heat exchange within the storage system, which comprises PCM units and the associated heat exchangers serving for charging and discharging the storage. The applied heat transfer mechanism is based on the HTF reflux created by a combined evaporation-condensation process. It was shown that a PCM with a fraction of 70 wt.% Zn in the alloy (Zn70Sn30) is optimal to attain a storage temperature of 370 deg. C, provided the heat source such as solar-produced steam or solar-heated synthetic oil has a temperature of about 400 deg. C (typical for the parabolic troughs technology). This PCM melts gradually between temperatures 200 and 370 deg. C preserving the latent heat of fusion, mainly of the Zn-component, that later, at the stage of heat discharge, will be available for producing steam. The thermal storage concept was experimentally studied using a lab scale apparatus that enabled investigating of storage materials (the PCM-HTF system) simultaneously with carrying out thermal performance measurements and observing heat transfer effects occurring in the system. The tests produced satisfactory results in terms of thermal stability and compatibility of the utilized storage materials, alloy Zn70Sn30 and the eutectic mixture of biphenyl and diphenyl oxide, up to a working temperature of 400 deg. C. Optional schemes for integrating the developed thermal storage into a solar thermal electric plant are discussed and evaluated considering a pilot scale solar plant with thermal power output of 12 MW. The storage should enable uninterrupted operation of solar thermal electric systems during additional hours

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-02-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  20. Design and development of steam generators for the AGR power stations at Heysham II/Torness

    International Nuclear Information System (INIS)

    Charcharos, A.N.; Jones, A.G.

    1984-01-01

    The current AGR steam generator design is a development of the successful once-through units supplied for the Oldbury Magnox and Hinkley/Hunterston AGR power stations. These units have demonstrated proven control and reliability in service. In this paper the factors which have dictated the design and layout of the latest AGR steam generators are described and reference made to the latest high temperature design techniques that have been employed. Details of development work to support the design and establish the performance characteristics over the range of plant operating conditions are also given. To comply with current UK safety standards, the AGR steam generators and associated plant are designed to accommodate seismic loadings. In addition, provision is made for an independent heat removal system for post reactor trip operations. (author)

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

    International Nuclear Information System (INIS)

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

    1996-07-01

    The objective of the Improved Eddy-Current ISI for Steam Generator Tubing program is to upgrade and validate eddy-current inspections, including probes, instrumentation, and data processing techniques for inservice inspection of new, used, and repaired steam generator tubes; to improve defect detection, classification, and characterization as affected by diameter and thickness variations, denting, probe wobble, tube sheet, tube supports, copper and sludge deposits, even when defect types and other variables occur in combination; to transfer this advanced technology to NRC's mobile NDE laboratory and staff. This report describes the design of specialized high-speed 16-coil eddy-current array probes. Both pancake and reflection coils are considered. Test results from inspections using the probes in working steam generators are given. Computer programs developed for probe calculations are also supplied

  2. Thermal-hydraulic phenomena during reflux condensation cooling in steam generator tubes

    International Nuclear Information System (INIS)

    Hae, Yong Jeong; Bum, Nyun Kim; Kwangho, Lee

    1998-01-01

    The transitions of cooling mechanism in steam generator tubes during reflux condensation are studied. It is found that the transitions are closely related to the occurrence of flooding or counter-current flow limitation phenomena in steam generator tubes. As shown in the previous studies of other researchers, the transition from filmwise reflux condensation into total reflux condensation occurs when the flooding criterion suggested by Wallis is met. In this study, it is suggested that the transition from total reflux condensation to complete carry-over occurs depending on the tube height and cooling conditions. It is also shown that the flooding at SG tubes occurs before the flooding at hot leg when a reflux condensation mode is existing in steam generator. Though the thermal-hydraulic conditions during reflux condensation after a small-break loss-of-coolant accident have enough margin to the transition into carry-over, considerations for the prevention of primary coolant relocation should be provided

  3. Severe transient tests on operating steam generator: analysis of the fluid-structure dynamic thermal interaction

    International Nuclear Information System (INIS)

    David, J.; Procaccia, H.; Billon, F.

    1983-08-01

    The operating efficiency of steam generators (S.G.s) and their structural integrity depend on the design configurations of the feedwater spray within the S.G., and on the operating procedure. To check the merit of some design modifications, and to verify the fluid-structure interaction with a view to preserve the S.G.s integrity during severe operating transients, a special instrumentation that admits the determination of the instantaneous thermal hydraulic characteristics of the flow in the secondary water and the corresponding metal temperature, in particular in the S.G. downcomer and near the S.G. tube sheet, has been installed by EDF on one steam generator of Tricastin unit 1 power plant. In parallel, FRAMATOME has developped a computer code, TEMPTRON, that allows the calculation of the thermal loads and the consequent stresses in the most sollicited zones of the steam generator during transient operation of the plant

  4. Simulating Porous Magnetite Layer Deposited on Alloy 690TT Steam Generator Tubes.

    Science.gov (United States)

    Jeon, Soon-Hyeok; Son, Yeong-Ho; Choi, Won-Ik; Song, Geun Dong; Hur, Do Haeng

    2018-01-02

    In nuclear power plants, the main corrosion product that is deposited on the outside of steam generator tubes is porous magnetite. The objective of this study was to simulate porous magnetite that is deposited on thermally treated (TT) Alloy 690 steam generator tubes. A magnetite layer was electrodeposited on an Alloy 690TT substrate in an Fe(III)-triethanolamine solution. After electrodeposition, the dense magnetite layer was immersed to simulate porous magnetite deposits in alkaline solution for 50 days at room temperature. The dense morphology of the magnetite layer was changed to a porous structure by reductive dissolution reaction. The simulated porous magnetite layer was compared with flakes of steam generator tubes, which were collected from the secondary water system of a real nuclear power plant during sludge lancing. Possible nuclear research applications using simulated porous magnetite specimens are also proposed.

  5. Robust and Low-Cost Flame-Treated Wood for High-Performance Solar Steam Generation.

    Science.gov (United States)

    Xue, Guobin; Liu, Kang; Chen, Qian; Yang, Peihua; Li, Jia; Ding, Tianpeng; Duan, Jiangjiang; Qi, Bei; Zhou, Jun

    2017-05-03

    Solar-enabled steam generation has attracted increasing interest in recent years because of its potential applications in power generation, desalination, and wastewater treatment, among others. Recent studies have reported many strategies for promoting the efficiency of steam generation by employing absorbers based on carbon materials or plasmonic metal nanoparticles with well-defined pores. In this work, we report that natural wood can be utilized as an ideal solar absorber after a simple flame treatment. With ultrahigh solar absorbance (∼99%), low thermal conductivity (0.33 W m -1 K -1 ), and good hydrophilicity, the flame-treated wood can localize the solar heating at the evaporation surface and enable a solar-thermal efficiency of ∼72% under a solar intensity of 1 kW m -2 , and it thus represents a renewable, scalable, low-cost, and robust material for solar steam applications.

  6. Method and apparatus for enhanced heat recovery from steam generators and water heaters

    Science.gov (United States)

    Knight, Richard A.; Rabovitser, Iosif K.; Wang, Dexin

    2006-06-27

    A heating system having a steam generator or water heater, at least one economizer, at least one condenser and at least one oxidant heater arranged in a manner so as to reduce the temperature and humidity of the exhaust gas (flue gas) stream and recover a major portion of the associated sensible and latent heat. The recovered heat is returned to the steam generator or water heater so as to increase the quantity of steam generated or water heated per quantity of fuel consumed. In addition, a portion of the water vapor produced by combustion of fuel is reclaimed for use as feed water, thereby reducing the make-up water requirement for the system.

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

    Science.gov (United States)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Bolt, P.R.

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  10. In situ ultrasonic examination of high-strength steam generator support bolts

    International Nuclear Information System (INIS)

    Jusino, A.

    1985-01-01

    Currently employed high-strength steam generator support bolting material (designed prior to ASME Section III Part NF or Component Supports), 38.1 mm in diameter, in combination with high preloads are susceptible to stress corrosion cracking because of the relatively low stress corrosion resistance (K/sub ISCC/) properties. These bolts are part of the pressurized water reactor steam generator supports at the integral support pads (three per steam generator, with each pad housing six, eight, or ten bolts depending on the design). The US Nuclear Regulatory Commission concerns for high-strength bolting were identified in NUREG-0577, ''Potential for Low Fracture Toughness and Laminar Tearing in PWR Steam Generator and Reactor Coolant Pump Supports,'' which was issued for comment on unresolved safety issue A-12. Subsequently, the bolting issues were addressed in generic issue B29. One of the issues deals specifically with high-strength bolting materials, which are vulnerable to stress corrosion cracking. A Westinghouse Owners Group funded program was established to develop in situ ultrasonic examination techniques to determine steam generator support bolting integrity at the head-to-shank and first-thread locations. This program was established in order to determine bolting integrity in place. Ultrasonic techniques were developed for both socket-head and flat-head bolt configurations. As a result of this program, in situ ultrasonic examination techniques were developed for examination of PWR steam generator support bolts. By employing these techniques utilities will be able to ensure the integrity of this in-place bolting without incurring the costs previously experienced during removal for surface examinations

  11. Efficiency of steam reheating in heat recovery steam generator at thermal power plant with gas turbine topping

    Directory of Open Access Journals (Sweden)

    Romashova Olga

    2017-01-01

    Full Text Available Possibility of using alternative option to the displacement of regenerative heating is analyzed in this paper: application of heat recovery steam generator’s (HRSG thermal power for reheating of working medium that has passed steam turbine.

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

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  13. Digitization and simulation realization of full range control system for steam generator water level

    International Nuclear Information System (INIS)

    Qian Hong; Ye Jianhua; Qian Fei; Li Chao

    2010-01-01

    In this paper, a full range digital control system for the steam generator water level is designed by a control scheme of single element control and three-element cascade feed-forward control, and the method to use the software module configuration is proposed to realize the water level control strategy. This control strategy is then applied in the operation of the nuclear power simulation machine. The simulation result curves indicate that the steam generator water level maintains constant at the stable operation condition, and when the load changes, the water level changes but finally maintains the constant. (authors)

  14. Some economic aspects of steam generator replacements in nuclear power plants

    International Nuclear Information System (INIS)

    Lebegner, J.

    1995-01-01

    The steam generator replacements performed over last decade (about 25 replacements until now), indicate trends towards improved techniques, shorter schedules and reduced total exposure and total costs. The goal of this paper is to give a worldwide review of SG replacement experience with accent on the economic aspect of the SG replacement. The main information about carried out replacements will be presented: cost, schedules, exposures, SG supplier and type, date of replacement, etc. Furthermore, the paper will contain the list of planned steam generator replacements in Europe, Japan and US future replacement plans. Finally, some of NPPs will be described whose initial nominal power has been increased along with SG replacement. (author)

  15. Removal of secondary sludge from steam generators used in French 900 class nuclear power plants

    International Nuclear Information System (INIS)

    Lebouc, B.

    1982-09-01

    The objective is to remove magnetite deposits which have formed on a steam generator tubesheet during plant operation. The deposits are separated from the tubesheet by spraying water at high pressure (about 200 bar at lance nozzle outlets) on each tube bundle ligament, i.e. the spaces between steam generator tubes. The water is recovered in suction lines and then filtered in two seperate units. The residue obtained after settling is removed in the form of solid waste. This paper presents the sludge lancing technique (spray lances, sludge recovery, liquid waste, cooling). A typical operating sequence is detailed (duration, personnel). Specifications for the equipment used are given

  16. Investigation of reliability of EC method for inspection of VVER steam generator tubes

    International Nuclear Information System (INIS)

    Corak, Z.

    2004-01-01

    Complete and accurate non-destructive examinations (NDE) data provides the basis for performing mitigating actions and corrective repairs. It is important that detection and characterization of flaws are done properly at an early stage. EPRI Document PWR Steam Generator Examination Guidelines recommends an approach that is intended to provide the following: Ensure accurate assessment of steam generator tube integrity; Extend the reliable, cost effective, operating life of the steam generators, and Maximize the availability of the unit. Steam Generator Eddy Current Data Analysis Performance Demonstration represents the culmination of the intense two-year industry effort in the development of a performance demonstration program for eddy current testing (ECT) of steam generator tubing. It is referred to as the Industry Database (IDB) and provides a capability for individual organizations to implement SG ECT performance demonstration programs in accordance with the requirements specified in Appendices G and H of the ISI Guidelines. The Appendix G of EPRI Document PWR Steam Generator Examination Guidelines specifies personnel training and qualification requirements for NDE personnel who analyze NDE data for PWR steam generator tubing. Its purpose is to insure a continuing uniform knowledge base and skill level for data analysis. The European methodology document is intended to provide a general framework for development of qualifications for the inspection of specific components to ensure they are developed in a consistent way throughout Europe while still allowing qualification to be tailored in detail to meet different nation requirements. In the European methodology document one will not find a detailed description of how the inspection of a specific component should be qualified. A recommended practice is a document produced by ENIQ to support the production of detailed qualification procedures by individual countries. VVER SG tubes are inspected by EC method but a

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

    CERN Document Server

    Ganapathy, V

    2014-01-01

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

  18. Electropolishing of replacement steam generator channel heads at Millstone-2 PWR

    International Nuclear Information System (INIS)

    Hudson, M.J.B.; Raney, H.; Raney, D.; Spalaris, C.N.

    1992-07-01

    A field application of EPRI-developed steam generator electropolishing technique was performed at Millstone-2 PWR. The process was qualified under previous programs on a laboratory scale, but it was thought appropriate to scale up application to full size components. Replacement of steam generators at Millstone-2 provided a unique opportunity to demonstrate that electropolishing can be applied safely and at a cost which was judged to be recoverable after a small number of fuel cycles. The project, preparation, electropolishing and cleanup, was completed at the reactor site in 25 working days. An alternate, less costly electrolyte solution was qualified for use in future applications

  19. Stress corrosion cracking of the tubing materials for nuclear steam generators in an environment containing lead

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Kim, Uh Chul; Lee, Eun Hee; Hwang, Seong Sik

    2004-01-01

    Steam generator tube materials show a high susceptibility to stress corrosion cracking (SCC) in an environment containing lead species and some nuclear power plants currently have degradation problems associated with lead-induced stress corrosion cracking in a caustic solution. Effects of an applied potential on SCC is tested for middle-annealed Alloy 600 specimens since their corrosion potential can be changed when lead oxide coexists with other oxidizing species like copper oxide in the sludge. In addition, all the steam generator tubing materials used for nuclear power plants being operated and currently under construction in Korea are tested in a caustic solution with lead oxide. (author)

  20. Analytical treatment of large leak pressure behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Hori, Masao; Miyake, Osamu

    1980-07-01

    Simplified analytical methods applicable to the estimation of initial pressure spike in case of a large leak accident in LMFBR steam generators were devised as follows; (i) Estimation of the initial water leak rate by the centered rarefaction wave method, (ii) Estimation of the initial pressure spike by the one-dimensional compressible method with either the columnar bubble growth model or the spherical bubble growth model. These methods were compared with relevant experimental data or other more elaborate analyses and validated to be usable in simple geometry and limited time span cases. Application of these methods to an actual steam generator case was explained and demonstrated. (author)