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Sample records for candu steam generator

  1. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in 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)

  2. Advancing CANDU experience to the world steam generator market

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. CANDU 6 steam generator thermalhydraulic modeling and simulation

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Steam generator life management

    International Nuclear Information System (INIS)

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

  11. CANDU energy for steam assisted gravity drainage

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Steam generator tube failures

    International Nuclear Information System (INIS)

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

  14. 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.

  15. The third generation CANDU control room

    International Nuclear Information System (INIS)

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

  16. Replacement steam generators for pressurized water reactors

    International Nuclear Information System (INIS)

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

  17. Candu technology: the next generation now

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. CANDU co-generation opportunities

    International Nuclear Information System (INIS)

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

  20. Trends in the capital costs of CANDU generating stations

    International Nuclear Information System (INIS)

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

  1. Key thrusts in next generation CANDU. Annex 10

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. Nuclear steam generator

    International Nuclear Information System (INIS)

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

  4. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

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

  5. Replacement nuclear steam generators

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. Steam generators - problems and prognosis

    Energy Technology Data Exchange (ETDEWEB)

    Tapping, R.L

    1997-05-01

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

  8. Steam generator life management

    International Nuclear Information System (INIS)

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

  9. Steam generator tube performance

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Steam generator module

    International Nuclear Information System (INIS)

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

  12. Steam generator materials

    International Nuclear Information System (INIS)

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

  13. Solar steam generation: Steam by thermal concentration

    Science.gov (United States)

    Shang, Wen; Deng, Tao

    2016-09-01

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

  14. Steam generator tube integrity program

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  15. 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.

  16. Application of a Zircaloy/Steam Oxidation Model to a CFX-10 Code for its Validation against a CANDU Fuel Channel Experiment: CS28-2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

    2007-11-15

    Oxidation of a Zircaloy cladding exposed to high-temperature steam is an important phenomenon in the safety analysis of CANDU reactors during a postulated loss-of-coolant accident (LOCA), since a Zircaloy/steam reaction is highly exothermic and results in hydrogen production.

  17. Environmental codes of practice for steam electric power generation

    International Nuclear Information System (INIS)

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

  18. Research program plan: steam generators

    International Nuclear Information System (INIS)

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

  19. Options for Steam Generator Decommissioning

    International Nuclear Information System (INIS)

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

  20. The CANDU contribution to environmentally friendly energy production

    International Nuclear Information System (INIS)

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

  1. Steam generator hand hole shielding.

    Science.gov (United States)

    Cox, W E

    2000-05-01

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

  2. Safety design of next generation SUI of CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-01-15

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

  3. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

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

  4. Proceedings of the fourth international conference on CANDU maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

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

  5. Proceedings of the fourth international conference on CANDU maintenance

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. CANDU operating experience

    International Nuclear Information System (INIS)

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

  8. Proceedings of the third international conference on CANDU maintenance

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Seung Yeol Yoo

    2015-01-01

    Full Text Available The current neutronics design methodology of CANDU-PHWRs based on the two-step calculations requires determining not only homogenized two-group constants for ordinary fuel bundle lattice cells by the WIMS-AECL lattice cell code but also incremental two-group constants arising from the penetration of control devices into the fuel bundle cells by a supercell analysis code like MULTICELL or DRAGON. As an alternative way to generate the two-group constants necessary for the CANDU-PHWR core analysis, this paper proposes utilizing a B1 theory augmented Monte Carlo (MC few-group constant generation method (B1 MC method which has been devised for the PWR fuel assembly analysis method. To examine the applicability of the B1 MC method for the CANDU 6 core analysis, the fuel bundle cell and supercell calculations are performed using it to obtain the two-group constants. By showing that the two-group constants from the B1 MC method agree well with those from WIMS-AECL and that core neutronics calculations for hypothetical CANDU 6 cores by a deterministic diffusion theory code SCAN with B1 MC method generated two-group constants also agree well with whole core MC analyses, it is concluded that the B1 MC method is well qualified for both fuel bundle cell and supercell analyses.

  10. 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.

  11. Quality Products - The CANDU Approach

    International Nuclear Information System (INIS)

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

  12. RPV steam generator pressure boundary

    Energy Technology Data Exchange (ETDEWEB)

    Strosnider, J.

    1996-03-01

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

  13. Strategic management of steam generators

    International Nuclear Information System (INIS)

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

  14. Repair technique for steam generator tubes using electroforming

    International Nuclear Information System (INIS)

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

  15. Steam generator tubing NDE performance

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  16. Innovations relative to steam generators

    International Nuclear Information System (INIS)

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

  17. Steam generator tubing NDE performance

    International Nuclear Information System (INIS)

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

  18. Design of PFBR steam generator

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Third international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

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

  1. 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.

  2. CAREM-25 Steam Generator Stability Analysis

    International Nuclear Information System (INIS)

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

  3. Strategic maintenance plan for Cernavoda steam generators

    International Nuclear Information System (INIS)

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

  4. Circumferential cracking of steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Karwoski, K.J.

    1997-04-01

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

  5. Reliability of steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  6. Centrifugal steam-water separator for steam generators

    International Nuclear Information System (INIS)

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

  7. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-15

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

  8. Steam generator tube laser sleeving

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

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

    International Nuclear Information System (INIS)

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

  13. Ultrasonic testing of steam generator tubes

    International Nuclear Information System (INIS)

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

  14. US PWR steam generator management: An overview

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  15. US PWR steam generator management: An overview

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

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

  18. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. Electric-arc steam plasma generator

    Science.gov (United States)

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

    2015-01-01

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

  1. 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)

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

    International Nuclear Information System (INIS)

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

  3. 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.

  4. Fourth international seminar on horizontal steam generators

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. 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...

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

    International Nuclear Information System (INIS)

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

  8. Revised evaluation of steam generator testing alternatives

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  10. Failure Analysis of Retired Steam Generator Tubings

    International Nuclear Information System (INIS)

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

  11. Heat Recovery Steam Generator by Using Cogeneration

    Directory of Open Access Journals (Sweden)

    P.Vivek, P. Vijaya kumar

    2014-01-01

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

  12. Dynamic simulation of steam generator failures

    International Nuclear Information System (INIS)

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

  13. Leakage experiences with 1 MW steam generator

    International Nuclear Information System (INIS)

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

  14. 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.

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

    International Nuclear Information System (INIS)

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

  16. COMMAND AND CONTROL STRATEGIES APPLIED TO HIGHPOWER STEAM GENERATORS

    OpenAIRE

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

    2015-01-01

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

  17. COMMAND AND CONTROL STRATEGIES APPLIED TO HIGHPOWER STEAM GENERATORS

    Directory of Open Access Journals (Sweden)

    DUINEA. A.M.

    2015-06-01

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

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

    Science.gov (United States)

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

    2012-05-01

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

  19. Artificial Intelligence Techniques for Steam Generator Modelling

    CERN Document Server

    Wright, Sarah

    2008-01-01

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

  20. Solar steam generation by heat localization.

    Science.gov (United States)

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

    2014-01-01

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

  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. Mathematical models for steam generator accident simulation

    International Nuclear Information System (INIS)

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

  3. AGR operational experience - steam generator materials constraints

    International Nuclear Information System (INIS)

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

  4. Perspectives of conventional and nuclear steam generation

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2016-01-01

    A steam generator is an essential unit of each cogeneration system using steam machines. Currently one of the cheapest ways of the steam generation can be application of old steam generators came from army surplus store. They have relatively simple construction and in case of not so exploited units – quite good general conditions, and functionality of mechanical components. By contrast, electrical components and control systems (mostly based on relay automatics) are definitely obsolete. It is...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  8. Corrosion Evaluation and Corrosion Control of Steam Generators

    International Nuclear Information System (INIS)

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

  9. 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.

  10. Incentives for improvement of CANDU

    International Nuclear Information System (INIS)

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

  11. LMFBR steam generators in the United Kingdom

    International Nuclear Information System (INIS)

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

  12. Hideout in steam generator tube deposits

    International Nuclear Information System (INIS)

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

  13. 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.

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

    Science.gov (United States)

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

    2016-03-01

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

  15. 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.

  16. Repair technology for steam generator tubes

    International Nuclear Information System (INIS)

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

  17. Ultrasonic examination techniques for steam generator tubing

    International Nuclear Information System (INIS)

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

  18. Selection of the design basis leak for LMFBR steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R A; Pfefferlen, H C; Roberts, J M; Sane, J O

    1977-11-01

    Steam generator tube failure mechanisms that have been observed in experiments or that have been postulated are discussed. The DBL for CRBRP and a proposed DBL for future large LMFBR steam generators are described. Safety considerations and philosophy in selection of Design Basis Leaks (DBL) are presented. A discussion of the Large Leak Test Rig (LLTR) Series II program support of the DBL selection for future large LMFBR steam generators is included.

  19. Steam generator operating experience update, 1982-1983

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  1. Mitigating aging in CANDU plants

    International Nuclear Information System (INIS)

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

  2. 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

  3. Joint studies on large CANDU

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Steam generator issues in the United States

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  6. Description of a program for steam generators

    International Nuclear Information System (INIS)

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

  7. Steam generator issues in the United States

    International Nuclear Information System (INIS)

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

  8. Internal ultrasonic testing of steam generator tubes

    International Nuclear Information System (INIS)

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

  9. Water chemistry of breeder reactor steam generators

    International Nuclear Information System (INIS)

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

  10. Crevice chemistry control in PWR steam generators

    International Nuclear Information System (INIS)

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

  11. Steam generators life extension experience at KANUPP

    International Nuclear Information System (INIS)

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

  12. Steam generator tube inspection in Japan

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  13. Comparison of steam generator methods in PISC

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2012-01-01

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

  16. 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...

  17. Study on Technology Solutions of CEFR Steam Generator

    Institute of Scientific and Technical Information of China (English)

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

    2012-01-01

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

  18. Generator of steam plasma for gasification of solid fuels

    Science.gov (United States)

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

    2013-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

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

    International Nuclear Information System (INIS)

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

  1. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

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

  2. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  4. Direct steam generation in line-focus solar collectors

    Science.gov (United States)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

  6. Mechanical design of a sodium heated steam generator

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Steam generator tube integrity flaw acceptance criteria

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  10. Welding for the CRBRP steam generators

    International Nuclear Information System (INIS)

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

  11. Life Assurance Strategy for CANDU NPP

    International Nuclear Information System (INIS)

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

  12. Next Generation Steam Cracking Reactor Concept

    OpenAIRE

    Van Goethem, M.W.M.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

  14. Methods of inspecting and repairing steam generator heating tubes

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. KANUPP operation with four out of six steam generators

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  20. 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.

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

    International Nuclear Information System (INIS)

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

  2. Dynamic and control of a once through steam generator

    International Nuclear Information System (INIS)

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

  3. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. MINET validation study using steam generator test data

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-01-01

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

  7. Modelling of a Coil Steam Generator for CSP applications

    OpenAIRE

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

    2014-01-01

    The project investigates a new design for a CSP plant steam generation system, the Coil Steam Generator (CSG). This system allows faster start-ups and therefore higher daily energy production from the Sun. An analytical thermodynamic simulation model of the evaporator and a mechanical analysis are developed to optimize the behavior of the system in different start-up scenarios. The results improve the effective life time (ELT) of the CSG, the thermal exibility of the overall CSP plant to have...

  8. Development of data management system for steam generator inspection

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

  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: e

  10. CANDU development

    International Nuclear Information System (INIS)

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

  11. Steam generator tube fretting - Darlington NGS experience

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-10-01

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

  15. Novel down-hole combustor for steam generation

    Energy Technology Data Exchange (ETDEWEB)

    Weissman, J.G.; Baird, B.; Alavandi, S.; Pfefferle, W.C. [Precision Combustion, Inc, North Haven, Connecticut, USA 06437 (United States); Etemad, S. [Fairfield University, Fairfield, Connecticut (United States)

    2011-07-01

    In the heavy oil industry, steam injection methods are often used to enhance oil recovery. Steam is usually generated on surface resulting in important heat loss during transfer to the reservoir; this impacts the economics and releases significant emissions of CO2. Precision Combustion, Inc. (PCI) is developing a novel downhole catalytic combustor for steam generation which will address the heat loss issue; the aim of this paper is to present this device and its advantages. Bench scale tests were performed under high pressure and low temperature conditions. Results proved the device to be durable and to produce clean steam for a wide range of applications. In addition it was demonstrated that this device provides a safe and controllable environment for enhanced oil recovery. This study highlighted the advantages of this downhole combustor which could be applied to heavy oil, shale oil, conventional oil or methane hydrate production.

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

    International Nuclear Information System (INIS)

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

  17. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  18. Hydrogen-based power generation from bioethanol steam reforming

    International Nuclear Information System (INIS)

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

  19. 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.

  20. 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.

  1. Steam generator assessment for sustainable power plant operation

    International Nuclear Information System (INIS)

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

  2. Detailed design of 700 MWe steam generator features

    Energy Technology Data Exchange (ETDEWEB)

    Korde, M.; John, B. [Nuclear Power Corp. of India Ltd., Mumbai, Maharashtra (India)]. E-mail: bjohn@npcil.co.in

    2006-07-01

    The next stage in the Indian nuclear power programme consists of building 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) units. This involves up-rating of all the plant equipment like reactor, steam generators (SGs), turbo-generator, major pumps, etc. The SG used in the current generation of IPHWRs, which have an electrical power of 540 MWe, is a mushroom type, inverted U-tube, natural-circulation SG. The 700 MWe SG design has the same the tube diameter, tube pitch and outer diameter of the steam generator sections as the 434 MWth SG, with certain changes in geometry of the feed header, flow restrictor in the downcomer and flow distribution plate. The changes resulted in a 30% increase in steam flow rate while maintaining the same circulation ratio. The paper describes detailing of these changes using a CFD code for optimizing the flow field. (author)

  3. 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.).

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2008-10-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    2010-10-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Forced circulation type steam generator simulation code: HT4

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. 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

  15. Evaluation of steam generator WWER 440 tube integrity criteria

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  16. Health and safety impact of steam generator tube degradation

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    1999-01-01

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

  19. 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.)

  20. A balanced strategy in managing steam generator thermal performance

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Steam generators regulatory practices and issues in Spain

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  3. In-service inspection of steam generator pipes

    International Nuclear Information System (INIS)

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

  4. 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.

  5. Status of the CRBRP steam-generator design

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-08-15

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

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

    OpenAIRE

    Folaranmi, Joshua

    2009-01-01

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

  8. Direct steam generation using a water injection system

    International Nuclear Information System (INIS)

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

  9. Effect of liquid waste discharges from steam generating facilities

    Energy Technology Data Exchange (ETDEWEB)

    McGuire, H.E. Jr.

    1977-09-01

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

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

    International Nuclear Information System (INIS)

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

  11. 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.

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  14. Steam generators clogging diagnosis through physical and statistical modelling

    International Nuclear Information System (INIS)

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

  15. CANDU nuclear power system

    International Nuclear Information System (INIS)

    This report provides a summary of the components that make up a CANDU reactor. Major emphasis is placed on the CANDU 600 MW(e) design. The reasons for CANDU's performance and the inherent safety of the system are also discussed

  16. Tests and analysis on steam generator tube failure propagation

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2015-08-01

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

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

    Science.gov (United States)

    Bergunker, V. D.

    2011-03-01

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

  19. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU-PHW program is based upon 38 years of heavy water reactor experience with 35 years of operating experience. Canada has had 72 reactor years of nuclear-electric operations experience with 10 nuclear units in 4 generating stations during a period of 18 years. All objectives have been met with outstanding performance: worker safety, public safety, environmental emissions, reliable electricity production, and low electricity cost. The achievement has been realized through total teamwork involving all scientific disciplines and all project functions (research, design, manufacturing, construction, and operation). (auth)

  20. Advanced CANDU reactor: an optimized energy source of oil sands application

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) is developing the ACR-700TM (Advanced CANDU Reactor-700TM) to meet customer needs for reduced capital cost, shorter construction schedule, high capacity factor while retaining the benefits of the CANDU experience base. The ACR-700 is based on the concept of CANDU horizontal fuel channels surrounded by heavy water moderator. The major innovation of this design is the use of slightly enriched uranium fuel in a CANFLEX bundle that is cooled by light water. This ensures: higher main steam pressures and temperatures providing higher thermal efficiency; a compact and simpler reactor design with reduced capital costs and shorter construction schedules; and reduced heavy water inventory compared to existing CANDU reactors. ACR-700 is not only a technically advanced and cost effective solution for electricity generating utilities, but also a low-cost, long-life and sustainable steam source for increasing Alberta's Oil Sand production rates. Currently practiced commercial surface mining and extraction of Oil Sand resources has been well established over the last three decades. But a majority of the available resources are somewhat deeper underground require in-situ extraction. Economic removal of such underground resources is now possible through the Steam Assisted Gravity Drainage (SAGD) process developed and proto-type tested in-site. SAGD requires the injection of large quantities of high-pressure steam into horizontal wells to form reduced viscosity bitumen and condensate mixture that is then collected at the surface. This paper describes joint AECL studies with CERI (Canadian Energy Research Institute) for the ACR, supplying both electricity and medium-pressure steam to an oil sands facility. The extensive oil sands deposits in northern Alberta are a very large energy resource. Currently, 30% of Canda's oil production is from the oil sands and this is expected to expand greatly over the coming decade. The bitumen deposits in the

  1. Gamma spectrometry application for steam generators radiological characterization

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    Science.gov (United States)

    Gallardo, Sergio; Querol, Andrea; Verdú, Gumersindo

    2014-06-01

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

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

    International Nuclear Information System (INIS)

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

  5. 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)

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

    International Nuclear Information System (INIS)

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

  7. 76 FR 74834 - Interim Staff Guidance on Aging Management Program for Steam Generators

    Science.gov (United States)

    2011-12-01

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

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

    International Nuclear Information System (INIS)

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

  9. An innovative piping verification program for steam generator replacement

    International Nuclear Information System (INIS)

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

  10. Ultrasonic downcomer flow measurements for recirculating steam generators

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

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

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

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

  12. 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.

  13. Homogenized behaviour of the steam generator perforated plates

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    居怀明; 左开芬; 刘志勇; 徐元辉

    2001-01-01

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

  15. Hitachi turbine generator technology for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, T.; Kudo, T. [Hitachi, Ltd., Power Systems, Hitachi Works, Hitachi (Japan); Akane, N. [Hitachi, Ltd. Power Systems, Nuclear Systems Division, Hitachi (Japan)

    2009-07-01

    Hitachi has supplied more than 1200 steam turbines and generators in the past 70 years for both thermal and nuclear applications. Hitachi nuclear steam turbines have been applied to all major reactor types including PWR's, BWR's and PHWR's (CANDU). Hitachi's recent experience has included supplying the steam turbines for Qinshan Phase III Unit 1 and 2 in China, powered by two CANDU 6 reactor, as well as several ABWR projects in Japan. Hitachi has focused significant R and D efforts on continuous improvement of nuclear steam turbine technology capitalizing on its continuous supply history and sound technical capability. This paper addresses some of the key developments and new technologies to be employed for new-build nuclear projects, including the ACR-1000 and Enhanced CANDU6, and focuses on longer Last Stage Blade (LSB) development, Continuous Cover Blades (CCB), and other enhancements in product reliability and performance. (author)

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

    International Nuclear Information System (INIS)

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

  17. Thermal hydraulic analysis of Alfred bayonet tube steam generator

    OpenAIRE

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

    2015-01-01

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

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

    Science.gov (United States)

    Hassanzadeh, Bazargan; Keshavarz, Ali; Ebrahimi, Masood

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    M. Głowacka

    2012-12-01

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

  1. Nuclear power plant modeling and steam generator stability analysis

    International Nuclear Information System (INIS)

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

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

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

    International Nuclear Information System (INIS)

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

  8. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results. Two LOF tests were analyzed in detail. Both tests were initiated from 100% power condition by shutting off the main feedwater flow. In LOF Test No. 1, the remaining boundary conditions were kept constant while in LOF Test No. 2, the power was rapidly reduced to 3%. The results show that the primary to secondary heat transfer becomes degraded when the collapsed water liquid level in the bundle region falls below approximately 50 inches. The SGTR test analyzed in detail - SGTR Test No. 2 - simulated the post-reactor-trip portion of the SGTR transient (T/sub prim/ = 5600F). The transient was initiated by starting the SGTR flow injection and simultaneously shutting off the auxiliary feedwater. The water level rose and flooded the dryer to its mid-elevation by the end of the test. The primary carry-over was shown to be less than 0.4% of the tracer mass injected into the secondary side by the SGTR flow. SGTR Test No. 3 investigated the response of the intact steam generator. Reverse heat transfer and low heat flow conditions were simulated. The results have demonstrated the occurrence of temperature stratification in the secondary water which lasted for about 800 seconds

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. 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.

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

    International Nuclear Information System (INIS)

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

  16. 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.

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

    International Nuclear Information System (INIS)

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

  18. Fabrication and inspection development for CRBRP steam generators

    International Nuclear Information System (INIS)

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

  19. The market for steam turbine generators around the world

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Yanjun Xiao

    2014-03-01

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

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

    Science.gov (United States)

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

    2012-05-01

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

  2. Thermoelectric generation coupling methanol steam reforming characteristic in microreactor

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

  5. Economics of CANDU

    International Nuclear Information System (INIS)

    The cost of producing electricity from CANDU reactors is discussed. The total unit energy cost of base-load electricity from CANDU reactors is compared with that of coal-fired plants in Ontario. In 1980 nuclear power was 8.41 m$/kW.h less costly for plants of similar size and vintage. Comparison of CANDU with pressurized water reactors indicated that the latter would be about 26 percent more costly in Ontario

  6. Alternate tube plugging criteria for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  7. Radiological assessment of steam generator repair and replacement

    International Nuclear Information System (INIS)

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

  8. Influence of sodium water reaction on MONJU steam generator

    International Nuclear Information System (INIS)

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

  9. Criteria for maintenance and repair - LMFBR steam generators

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    M. Głowacka; J. Łabanowski; S. Topolska

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

  12. 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.

  13. Application of ultrasonic shot peening to steam generator nozzles

    International Nuclear Information System (INIS)

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

  14. Radiological assessment of steam generator repair and replacement

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-12-01

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

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

    OpenAIRE

    Vladimir Melikhov; Oleg Melikhov; Yury Parfenov; Alexey Nerovnov

    2011-01-01

    The horizontal steam generator (SG) is one of specific features of Russian-type pressurized water reactors (VVERs). The main advantages of horizontal steam generator are connected with low steam loads on evaporation surface, simple separation scheme and high circulation ratio. The complex three-dimensional steam-water flows in the steam generator vessel influence significantly the processes of the steam separation, distribution, and deposition of the soluble and nonsoluble impurities and dete...

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  18. Analysis of once-through steam generator instability

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  19. Recent operating experiences with steam generators in Japanese NPPs

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  20. Corrosion evaluation of alternate nuclear steam generator tubing materials

    International Nuclear Information System (INIS)

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

  1. Steam generator life cycle management challenges - on-going and new build

    International Nuclear Information System (INIS)

    Ontario Power Generation (OPG) is committed to the safe, reliable, and cost-effective operation of its fleet of CANDU plants. Steam Generators (SGs) are a major component of the heat transport system in these plants and maintaining their health is an essential element to achieving plant safety, reliability and economic performance. OPG has been actively engaged in formal life cycle management of its SGs for about 15 years. Over this time, we have developed stable, mature, detailed life cycle plans for each of our plants on a unit by unit, and in some cases, SG by SG, basis. These plans have been externally reviewed over the years by our regulator and by other third-party experts, and they've been acknowledged as being among the best life cycle plans anywhere. Although we are pleased that our life cycle plans are as detailed and mature as they are, we certainly aren't fully satisfied because they're not perfect. Even if they were perfect at any point in time, they wouldn't be for very long because the environment is constantly changing, both the technical environment and the business environment. This paper presents some of these challenges and offers some possible solutions or suggestions based on OPG's experience. The paper describes the background on SG life cycle management in OPG, i.e. what it is and how we do it. Then it presents challenges in the following areas: despite having some very detailed and technically strong life cycle plans, we still face some technical issues; in addition, we face challenges in integrating these plans into the overall business processes within the company; up until now, our life cycle planning has been aimed at early-and mid-life in our units. But our units are aging and we are now within sight, at least in a life cycle management sense, of a point at which decisions need to be made on refurbishment, life extension or retirement of the units. We need to adjust our life cycle management approach as we approach those major

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

    Science.gov (United States)

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

    2014-02-01

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

  3. Seismic analysis of steam generator and parameter sensitivity studies

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  6. CFD Analysis of Aspirator Region in a B&W Enhanced Once-Through Steam Generator

    OpenAIRE

    Spontarelli, Adam Michael

    2013-01-01

    This analysis calculates the velocity profile and recirculation ratio in the aspirator region of an enhanced once-through steam generator of the Babcock & Wilcox design. This information is important to the development of accurate RELAP5 models, steam generator level calculations, steam generator downcomer models, and flow induced vibration analyses. The OpenFOAM CFD software package was used to develop the three-dimensional model of the EOTSG aspirator region, perform the calculations, and p...

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

    OpenAIRE

    Esch Markus; Knoche Dietrich; Hurtado Antonio

    2014-01-01

    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 bein...

  8. A CFD Simulation of a 1/5-Scale Steam Generator Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Euh, Dong Jin; Shin, Byung Soo; Ko, Yung Joo; Kwon, Tae Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    A 1/5-scale steam generator simulator was developed for core flow test of SMART. Its overall layout is illustrated in Fig.1. To simulate a pressure drop induced by a helical tube of steam generator, an orifice is installed in the lower part of the 1/5-scale steam generator simulator. The axial flow rate is measured by a venturi flow. The SG bypass fraction is ignored because it is quite a few amount compared with the total flow amount

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    OpenAIRE

    Yanjun Xiao; Xuewei Ma; Wei Shao; Yuming Guan

    2014-01-01

    The object of this study is a kind of 3 kW small steam generator, which can recover waste heat through making use of 0.1~0.3 MPa steam. This can exploit secondary energy efficiently. The electricity generated can be commonly used as factory lighting, heating, fan and emergency power supply. But the generation voltage of the existed steam turbine is instable, especially when the steam pressure and the load of the generator changes suddenly. This can pose a threat to electrical safety and great...

  12. The new steam generators for the Krško nuclear power plant

    OpenAIRE

    Nemčić, Krešimir

    2015-01-01

    This paper presents the new steam generators for the Krško nuclear power plant and describes the main design and fabrication improvements which are an ongoing research and development effort in steam generator technology, aimed at improving the reliability and maintainability of the new steam generators. The paper also provides basic information relating to the manufacturing of the new steam generators. Članek predstavlja nova uparjalnika za JE Krško in opisuje glavne izboljšave projekta i...

  13. [A study on steam generator in solid amine CO2 purification system].

    Science.gov (United States)

    Zhou, K H; Liu, X Y; Lu, X Y; Ai, S K; Li, S L; Huang, Y

    2001-04-01

    Objective. To solve the key problems of power matching between process of CO2 steam desorption and process of steam generation, as well as water/vapor separation. Method. Solid amine desorption process was studied by thermodynamic analysis and experiments. The distribution rule of desorption energy was found out and then the power consumption of steam generator was decided. Ceramic insert was designed to separate water and vapor making use of surface tension. Finally, the steam generator was designed on system requirements. Result. Experiments proved that the steam generator can satisfy the demand of the system as well as successfully separate water and vapor, in addition, the selected power is suitable. Conclusion. The design on steam generator was right and practicable. PMID:11808568

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  16. Steam Generator Chemical Cleaning Application: Korean Experience in PWR NPP

    International Nuclear Information System (INIS)

    Korea Hydro and Nuclear Power (KHNP) performed an EPRI/SGOG chemical cleaning of the secondary side of the steam generators at Ulchin Unit 3 (UCN3) in March 2011 and at Ulchin Unit 4 (UCN4) in September 2011. The steam generator chemical cleaning (SGCC) was performed with venting at the top-of-tube sheet (TTS) and at tube support plates (TSPs) 4, 5, 6, 7, 8, 9, and 10. A primary objective of this SGCC was to address outer diameter stress corrosion cracking (ODSCC), which has been observed at the TTS and TSPs in the UCN3 SGs. The EPRI/SGOG process has been shown to effectively reduce prevailing ODSCC rates at the TTS and TSPs, particularly when applied with periodic venting in this application. This was the first full-length SGCC campaign with venting performed in Korea. Ulchin Unit 3 commenced commercial operation in August 1998 and Ulchin Unit 4 commenced commercial operation in December 1999. UCN3 and UCN4 are a two-loop pressurized water reactor (PWR) of the Korea Standard Nuclear Plant (KSNP) design. The SGs contain high-temperature mill annealed (HTMA) Alloy 600 tubing and are similar in design to the Combustion Engineering CE-80. The KSNP SGs have been susceptible to outer diameter stress corrosion cracking (ODSCC), which is consistent with operating experience for other SGs containing Alloy 600HTMA tubing material. The UCN3/4 SGs have recently begun to experience ODSCC. Hankook Jungsoo Industries Co., Ltd (HaJI) was selected as the cleaning vendor by KHNP. To date, HaJI has completed five Advanced Scale Conditioning Agent (ASCA) cleaning applications and two EPRI/SGOG Steam Generator Chemical Cleaning (SGCC) campaigns for KHNP. The goal of total deposit removal of the applications were successfully achieved and the amounts are 3,579 kg at UCN3 and 3,786 kg at UCN4 which values were estimated before each cleaning by analysing ECT signal and liquid samples from the SGs. The deposits from the SGs were primarily composed of magnetite. There were no chemical

  17. STGEN computer program methodology and application to the steam line break tests in the RD-12 steam generator

    International Nuclear Information System (INIS)

    A computer model for calculating a U-tube stream generator response to a main stream line break accident is presented. The stream generator is nodalized into three regions: riser, steam drum, and downcomer. Mass and pressure (temperature) are the main state variables representing each region. Models for steam drum swelling, tube bundle heat transfer, liquid subcooling, void fraction distribution and two-phase flow in the various flow connections are all formulated based on the state variables of the regions. The model was tested in a series of experiments carried out with the steam generator RD-12. Comparison between the experimental data and code predictions are discussed for small and large breaks. The model correctly predicts the major quantitative trends and many of the quantitative features. This includes global system variables such as the pressure variations as well as local variables such as quality, void fractions, liquid levels, and dryout level. (orig.)

  18. Modeling local chemistry in PWR steam generator crevices

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  19. Denting of steam generator tubes in PWR plants

    International Nuclear Information System (INIS)

    Reactions of 12 Cr and carbon steels in approximately 300 C deoxygenated water containing various concentrations of metal chlorides while adjacent to or connected to Alloy 600 are reported. The tests were designed to determine the extent to which the 12 Cr (AISI 405) steel accumulated corrosion products similar to those found on carbon steel which have caused crevice denting of Alloy 600 tubes in steam generators. Static isothermal autoclave and capsule tests were made in Na, Mg, Fe, and Ni and the corrosion products analyzed. 12 Cr steel's resistance was superior to C steel by factors ranging from 1.5 to 80. 12 Cr steel corrosion showed little temperature dependence in the range 232 to 316 C and zero volume change resulting from oxides. 11 refs

  20. Some aspects of materials development for sodium heated steam generators

    International Nuclear Information System (INIS)

    A development program was undertaken to support the materials selection for steam generator piping and IHX which are to be used in Liquid Metal Fast Breeder Reactors (LMFBR). Four major topics were reviewed, describing the results obtained as well as the direction of future tests. These topics are: carbon transport in sodium, effect of carbon loss/gain upon materials in the reactor Intermediate Heat Transport System (IHTS), corrosion fatigue and aqueous corrosion. The results support the initial assumptions made in specifying the use of 2-1/4Cr-1Mo as the construction material for the evaporator and superheater and Type 316 piping of the IHT system. Future direction of the experimental programs is to further verify the materials choice and to also obtain information which will be essential during the plant installation, operation and reliability of the components

  1. Non linear identification applied to PWR steam generators

    International Nuclear Information System (INIS)

    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

  2. Adsorption of sulfate in PWR steam generators: Laboratory tests

    International Nuclear Information System (INIS)

    Following observation of an apparent difference in the hideout mechanism for sulfate compared to that of other highly soluble species during chemical injection tests at several PWRS, a laboratory test program, discussed in this report was implemented to quantify sulfate adsorption on metal surfaces. Approximately 350 ug/m2 of sulfate could be adsorbed on Alloy 600 from neutral solutions at 300 degree C. Less adsorption was observed at lower temperature as well as at increased pH. The adsorbed sulfate could be desorbed into pure water over a period of several days subsequent to termination of sulfate ingress. Thus, a prompt shutdown to hot standby with maximization of blowdown should minimize the long term impact of sulfate steam generator corrosion subsequent to a period of significant sulfate or cation resin ingress. The only other species which exhibited significant adsorption was phosphate which also has a tetrahedral ionic structure in solution

  3. Flow-induced vibration of steam generator tube bundles

    International Nuclear Information System (INIS)

    The vibrations induced in tube arrays by a transversal flow are of great practical interest because of their destructive effects especially on heat exchangers. Instabilities can appear beyond a critical flow velocity and induce very high vibratory levels which may involve fractures. These instabilities involve a fluidelastic coupling between the vibratory movement of the tubes and the flow round them. Studies are being carried out in France concerning steam generators. A lot of bundle mock-ups with various pitches have been tested and large range parameter domains have been investigated. In a second part, the C.E.A., FRAMATOME, E.D.F. and WESTINGHOUSE research program, which is being carried out, is presented

  4. Burnout power in once-through tubular steam generators

    International Nuclear Information System (INIS)

    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/cm2 and specific mass flowrates from 110 to 260 g/cm2s 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/cm2 and 110 less than p less than 160 kg/cm2 two different burnout mechanisms may set up. (U.S.)

  5. Oxides as barriers to tritium permeation in steam generators and tritium content in CTR coolants

    International Nuclear Information System (INIS)

    The primary release of tritium from a fusion reactor complex into the environment is via the steam generator system. Tritium in the coolant can permeate through the heat exchanger into the steam cycle, and is trapped in the steam as HTO. Subsequent recovery of tritium from the steam is impractical. The amount of tritium that permeates into the steam cycle will depend on the concentration of tritium in the coolant, or more significantly the amount of tritium that can be allowed in the coolant will depend on the rate of tritium permeation that can be tolerated

  6. 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.

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

    International Nuclear Information System (INIS)

    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

  8. CANDU 9 design

    International Nuclear Information System (INIS)

    AECL has made significant design improvements in the latest CANDU nuclear power plant (NPP) - the CANDU 9. The CANDU 9 operates with the energy efficient heavy water moderated reactor and natural uranium fuel and utilizes proven technology. The CANDU 9 NPP design is similar to the world leading CANDU 6 but is based upon the single unit adaptation of the 900 MWe class reactors currently operating in Canada as in integrated four-unit configurations. The evolution of the CANDU family of heavy water reactors (HAIR) is based on a continuous product improvement approach. Proven equipment and systems from operating stations are standardized and used in new products. As a result of the flexibility of the technology, evolution of the current design will ensure that any new requirements can be met, and there is no need to change the basic concept. This paper will provide an overview for some of the key features of the CANDU 9 NPP such as nuclear systems and equipment, advanced control and computer systems, safety design and protection features, and plant layout. The safety enhancements and operability improvements implemented in this design are described and some of the advantages that can be expected by the operating utility are highlighted. (author)

  9. SNR-steam generator design with respect to large sodium water reactions

    International Nuclear Information System (INIS)

    This paper deals with the experiences gained during the licensing procedure for the steam generators for the SNR 300 LMFBR regarding large sodium-water reactions. A description is given of the different calculations executed to investigate the effects of large leaks on the 85 MW helical coiled and straight tube steam generators. The investigations on the helical coiled steam generators are divided in the formulations of fluid behaviour, dynamic force calculations, dynamic response calculation and finally stress analyses. Several results are shown. The investigations on the straight tube steam generators are performed using models describing fluid-structure interaction, coupled with stress analyses. Several results are presented. A description is given of the problems and necessary construction changes during the licensing process. Advises are given for future analyses and design concepts for second generation commercial size LMFBR steam generators with respect to large leaks; based on the experience, gained with SNR 300, and using some new calculations for SNR 2. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  11. 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.

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

    International Nuclear Information System (INIS)

    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.)

  13. Replacement steam generators for Calvert Cliffs, Oconee and future replacement design

    International Nuclear Information System (INIS)

    After the completion of steam generators presently being fabricated, a total of forty replacement steam generators will have been built for fourteen reactor units located at ten reactor sites. This represents approximately $1 billion of manufacture excluding installation costs. Replacement steam generator work began with the initiation of the Millstone 2 steam generator replacement program for Northeast Utilities in 1989. Manufacture is presently underway on replacement recirculating steam generators for Calvert Cliffs Units 1 and 2 plants of Constellation Nuclear (OEM Combustion Engineering) and the once-through steam generators for the Oconee 1, 2 and 3 plants of Duke Power (OEM Babcock and Wilcox). These two sites are the first and second respectively to have applied for and received approval for a life extension of 20 years beyond their original operating license. The application and granting of these license extensions reflects a major change in the nuclear industry over the recent past. The attitude to nuclear power has changed from a relatively defensive strategy to a much more optimistic agenda of utility reorganization, purchase of well performing older plants, replacement of aging components, plant refurbishment, and upgrades and applications for license extension. Possible new plants are also being considered. The paper discusses specific features, attributes, performance and operating experience with replacement steam generators (RSGs) both in service and under construction. Industry issues and design features applicable to future replacement steam generators are also reviewed. (author)

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

    International Nuclear Information System (INIS)

    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

  15. Hydrogen analysis using MELCOR at CANDU plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Chu; Park, Jae Hong; Kim, Han Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-10-15

    As a part of a submitted Wolsong Severe Accident Management Plan(SAMP) at the end of 2009, KINS needs to the review of it. In this the focus is on the hydrogen behavior during the station blackout(SBO) because the hydrogen explosion during severe accident in CANDU plants such as Wolsong units has been safety issue. A SBO occurs with failure of the emergency power system when loss of class IV electric power happens because the loss of class IV power is the most dominant internal event causing core damage for Wolsong units. And following a SBO event, most of the engineered safety features(ESFs), including hydrogen igniters, are inoperable except the passive systems such as Dousing systems. Hydrogen is generated due to Zr-steam reaction in fuel channels and core debris oxidation in the suspended debris beds, jet breakup of molten debris in the water pool of the reactor vault and molten core-concrete interaction (MCCI). A combustible gas control system consisting of Passive Autocatalytic Recombiner(PAR) is currently installed at Wolsong unit 1. The results of MELCOR analysis are presented

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-09-15

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  19. Leak injection/detection input for B and W prototype steam generator test request

    International Nuclear Information System (INIS)

    The goal of the leak injection/detection phase of the test program on the prototype steam generator is to obtain data that can be used to specify the leak protection system for the plant unit steam generators. Both chemical and two acoustic leak detection methods (by GE and Rockwell International) are to be considered. The chemical system has been selected as the reference based on its more developed state. The acoustic methods have potential both as small leak detection systems and as intermediate leak protection/automatic shutdown systems. Simulated leak injections will be made at various locations within the steam generator to determine the performance of the chemical system as specifically applied to the B and W helical coil steam generator geometry. Acoustic tests will be made to characterize the various steam generator background noise sources and to record acoustic signals during smulated leak injections, in order to predict the performance of both systems

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  1. Marketing CANDU internationally

    International Nuclear Information System (INIS)

    The market for CANDU reactor sales, both international and domestic, is reviewed. It is reasonable to expect that between five and ten reactors can be sold outside Canada before the end of the centry, and new domestic orders should be forthcoming as well. AECL International has been created to market CANDU, and is working together with the Canadian nuclear industry to promote the reactor and to assemble an attractive package that can be offered abroad. (L.L.)

  2. Stress corrosion cracking experience in steam generators at Bruce NGS

    International Nuclear Information System (INIS)

    In late 1990 and through 1991, units 1 and 2 at the Bruce A Nuclear Generating Station (BNGS-A) experienced a number of steam generator tube leaks. Tube failures were identified by eddy current to be circumferential cracks at U-bend supports on the hot-leg side of the boilers. In late 1991, tubes were removed from these units for failure characterization. Two active failure modes were found: corrosion fatigue in both units 1 and 2 and stress corrosion cracking (SCC) in unit 2. In unit 2, lead was found in deposits, on tubes, and in cracks, and the cracking was mixed-mode: transgranular and intergranular. This convincingly indicated the involvement of lead in the stress corrosion cracking failures. A program of inspection and tube removals was carried out to investigate more fully the extent of the problem. This program found significant cracking only in lead-affected boilers in unit 2, and also revealed a limited extent of non-lead-related intergranular stress corrosion cracking in other boilers and units. Various aspects of the failures and tube examinations are presented in this paper. Included is discussion of the cracking morphology, measured crack size distributions, and chemical analysis of tube surfaces, crack faces, and deposits -- with particular emphasis on lead

  3. Short history of steam generators in the USSR

    International Nuclear Information System (INIS)

    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

  4. Improved eddy-current inspection for steam generator tubing

    International Nuclear Information System (INIS)

    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

  5. R and D in support of CANDU plant life management

    International Nuclear Information System (INIS)

    One of the keys to the long-term success of CANDUs is a high capacity factor over the station design life. Considerable R and D in underway at AECL to develop technologies for assessing, monitoring and mitigating the effect of plant ageing and for improving plant performance and extending plant life. To achieve longer service life and to realize high capacity factor from CANDU stations, AECL is developing new technologies to enhance fuel channel and steam generator inspection capabilities, to monitor system health, and to allow preventive maintenance and cleaning (e.g., on-line chemical cleaning processes that produce small volumes of wastes). The life management strategy for fuel channels and steam generators requires a program to inspect components on a routine basis to identify mechanisms that could potentially affect fitness-for-service. In the case of fuel channels, the strategy includes inspections for dimensional changes, flaw detection, and deuterium concentration. New techniques are been developed to enhance these inspection capabilities; examples include accurate measurement of the gap between a pressure tube and its calandria tube and rapid full-length inspections of steam generator tubes for all known flaw types. Central to life management of components are Fitness-for-Service Guidelines (FFSG) that have been developed with the CANDU Owners Group (COG) that provide a standardized method to assess the potential for propagation of flaws detected during in-service inspections, and assessment of any change in fracture characteristics of the material. FFSG continue to be improved with the development of new technologies such as the capability to credit relaxation of stresses due to creep and non-rejectable flaws in pressure tubes. Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that system health is continually monitored and managed. AECL has developed a system Health Monitor

  6. Support vector regression model based predictive control of water level of U-tube steam generators

    International Nuclear Information System (INIS)

    Highlights: • Water level of U-tube steam generators was controlled in a model predictive fashion. • Models for steam generator water level were built using support vector regression. • Cost function minimization for future optimal controls was performed by using the steepest descent method. • The results indicated the feasibility of the proposed method. - Abstract: A predictive control algorithm using support vector regression based models was proposed for controlling the water level of U-tube steam generators of pressurized water reactors. Steam generator data were obtained using a transfer function model of U-tube steam generators. Support vector regression based models were built using a time series type model structure for five different operating powers. Feedwater flow controls were calculated by minimizing a cost function that includes the level error, the feedwater change and the mismatch between feedwater and steam flow rates. Proposed algorithm was applied for a scenario consisting of a level setpoint change and a steam flow disturbance. The results showed that steam generator level can be controlled at all powers effectively by the proposed method

  7. Exergetic Optimization of the Heat Recovery Steam Generators by Imposing the Total Heat Transfer Area

    OpenAIRE

    Cenuşă, Victor-Eduard; Feidt, Michel; Badea, Adrian; Benelmir, Riad

    2004-01-01

    The paper presents an original and fast method for the heat recovery steam generator (HRSG) exergetic optimization. The objective is maximizing the exergy transfer to the water / steam circuit. The proposed approach, different from the classical method that fixes the pinch point, is essentially thermodynamic but it considers also the economics by imposing the total heat transfer area of HRSG. The HRSG may have one or two steam pressures, without reheat. The input data from the gas turbine are...

  8. Numerical investigation of three-dimensional flows of steam-water mixture in the housing of the PGV-1000 steam generator

    Science.gov (United States)

    Kroshilin, A. E.; Kroshilin, V. E.; Smirnov, A. V.

    2008-05-01

    Results are given of numerical simulation of three-dimensional pattern of flow of a two-phase steam-water mixture in the house of a PGV-1000 horizontal steam generator obtained using the BAGIRA best-estimate thermohydrodynamic computer codes. The space distributions of velocities and local void fractions in the steam generator housing for different modes of operation of power-generating unit are calculated and compared with available experimental data.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  11. 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.

  12. Development of weld plugging for steam generator tubes of FBR

    International Nuclear Information System (INIS)

    This study was undertaken to develop a method of weld plugging of the heat-exchanger tubes of steam generator of Prototype FBR 'MONJU' in case these tubes are damaged for some reason. We studied mainly the shape of plug, welding procedure and effect of postweld heat treatment (PWHT). Evaporator tube sheet, tube and plug are made of 2-1/4Cr-1Mo steel and usually preheating and PWHT will be required for welding of this steel. The results of this study is as follows. 1) Plug was designed to make butt joint welding with grooved tube sheet around the tube hole to satisfy the requirements of plug designing, stress analysis, and good weldability. 2) TIG welding process was selected and certified its good weldability and good performance. 3) PWHT can be done by using high frequency induction heating method locally and also designing the plug to weld joint with tube sheet which was grooved around the tube hole. 4) Mock up test was done and it was certified that this plugging procedure has good weldability and good performance ability for Non Destructive Inspection. (author)

  13. Specific ultrasonic inspection methods for steam generator tubes

    International Nuclear Information System (INIS)

    Framatome has developed a computerized equipment for inspecting PWR steam generator tubes using a rotating ultrasonic probe. Firstly devoted to the examination of the roll transition zone at the tube sheet secondary side level, the testing system can also operate now for inspections at the tube support plate levels. It is used independently for specific tube inspection, or it can be integrated into a broader-purpose system for sleeve weld testing, etc. The testing results are displayed in real time by means of two eight-level coded colored maps. Some applications, ranging from mockup testing to on-site inspection, are presented in this paper. In conclusion: An automated ultrasonic real-time imaging system for tube-wall thickness measurement, internal profilometry, and flaw detection, has been developed. This system has been successfully applied for on-site specific inspections within an industrial environment. In all cases the ultrasonic acquisition time was less than two minutes per tube. It should be pointed out that analysis of this new set of ultrasonic inspection results should also improve understanding of the in-service behaviour of these materials and components

  14. Analysis of the State of Steam Generator Tubes

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  15. 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)

  16. 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.

  17. Iodide volatility under condition relevant to PWR steam generator faults

    International Nuclear Information System (INIS)

    The evaluation of iodine volatility during steam generator tube rupture (SGTR) is hampered by three factors: (i) lack of suitable plant data under fault conditions, (ii) lack of experimental data (mainly due to the difficulty of performing experiments under the conditions required) and (iii) uncertainty in theoretical methods to extrapolate experimental data to the required conditions. This report summarises methods of estimating the volatility of hydrogen iodide and iodide salts at the required conditions of temperature and pressure. A thermodynamic method has been used to estimate HI volatility and the density correlation method for iodide salt volatility. It is assumed throughout that it is more conservative to predict higher volatility. Consideration is given to two explanations of experiments carried out at Oak Ridge National Laboratory (ORNL) on the influence of boric acid concentration and pH on the volatility of radioiodine ostensibly under SGTR conditions: (i) the results have been interpreted in terms of reactions involving volatility of iodide salt/ion-pairs and complexation by boric acid in the gas phase and (ii) the possibility is explored that the observed results are due to the influence of oxidation leading to the formation of much more volatile iodine species. (author)

  18. Lessons learned from tubes pulled from French steam generators

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  19. Innovations During Surface Treatment of PFBR Steam Generators in 91 Grade Material

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is a 500MWe pool type, mixed oxide fuelled, sodium cooled nuclear reactor, which is in advanced stage of construction at Kalpakkam, India. The Steam Generator (SG) is a vertical, once through, shell & tube type heat exchanger with liquid sodium flowing in shell side and water/steam flowing in the tube side. Due to very high reactivity of sodium with water/steam, the boundaries of sodium to water/steam in SG must possess a high degree of integrity & reliability against failure. This is achieved by precise design, correct material selection and high standard quality control & quality assurance during manufacture. Modified 9Cr-1Mo material is selected as principal material of construction. Application of suitable corrosion protection technology is extremely important during fabrication for safe and reliable operation due to critical nature of component. After completion of manufacture of Steam Generators, degreasing, pickling and passivation is carried out on inside surfaces to ensure passive chromium oxide layer. Due to asymmetric shape, configuration and complex constructional features, the surface treatment of Steam Generators is extremely difficult & really challenging task. Huge quantity of nitric acid (HNO3), HF and Demineralized (DM) water is used during surface treatment of each Steam Generators. Due to hazardous nature of acids, a closed loop of mini chemical plant is constructed for large scale circulation of picking & passivation solution inside the Steam Generators. Enormous efforts were put and many trials were conducted and various innovations/new techniques were followed for the first time in the nuclear history for effective surface treatment of Steam Generators for 40 years design service life. This paper explains the challenges faced and experience gained during surface treatment on modified 9Cr-1Mo surfaces of PFBR Steam Generators in detail. (author)

  20. Steam generator materials performance in high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    This paper reviews the materials technology aspects of steam generators for HTGRs which feature a graphite-moderated, uranium-thorium, all-ceramic core and utilizes high-pressure helium as the primary coolant. The steam generators are exposed to gas-side temperatures approaching 7600C and produce superheated steam at 5380C and 16.5 MPa (2400 psi). The prototype Peach Bottom I 40-MW(e) HTGR was operated for 1349 EFPD over 7 years. Examination after decommissioning of the U-tube steam generators and other components showed the steam generators to be in very satisfactory condition. The 330-MW(e) Fort St. Vrain HTGR, now in the final stages of startup, has achieved 70% power and generated more than 1.5 x 106 MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, requiring a number of new materials factors including creep-fatigue and water chemistry control. Current designs of larger HTGRs also feature steam generators of helical once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers (ASME) Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of large forgings, etc

  1. Dilute chemical cleaning of PWR steam generators off-line cleaning process evaluation

    International Nuclear Information System (INIS)

    This project evaluated the feasibility of using a low-concentration (approx. 0.5 wt %) chemical cleaning process to remove corrosion product deposits from steam generator surfaces and magnetite from tube-to-support plate crevices of PWR steam generators. The primary objective was to develop a dilute process that could be safely applied at scheduled intervals, such as during normal refueling outages, to maintain a clean operating condition in the steam generator. The dilute chemical cleaning process developed in this project was demonstrated successfully on two model generators which were operated on faulted chemistry by DOE/CRC at Commonwealth's State Line Facility. Unit 5 was cleaned after 48 days of operation with 1% seawater fouling, and Unit 6 was cleaned after 112 days of operations with Lake Michigan water. This report describes work leading to the model generator cleaning demonstrations and provides details of the cleaning operation for each model steam generator

  2. U-tube steam generator model design based upon the transient modelling request

    International Nuclear Information System (INIS)

    Scaling factors which have to be equal for steam generator model and object(steam generator of NPP Fessenheim) are derived upon the modelling laws. Working fluids and main parameters are determined in advance by purpose and location of model, so the great deal of factors couldn't be satisfied. It is carried out that the similar transient behaviour of model and object can be obtained if the scaling factors which define time-dependent pressure and heat-exchange rate are satisfied. The vertical U-tube steam generator model is designed in this way, as a part of nuclear power plant model at Mechanical Engineering Faculty in Belgrade. (author)

  3. Structured singular value synthesis based steam generator water level controller design

    International Nuclear Information System (INIS)

    An uncertainty kinetic model for valve position of feed water to steam generator water level plant was built which can express parameter perturbation and unmodeled dynamic due to operation condition change of steam generator and feed water pump. Robust controller was designed based on structured singular value synthesis method. Robust stability of the water control system can be guaranteed under parameter perturbation and unmodeled dynamic, and robust performance can be acquired via suitable performance weight function. Simulation results show that good control performances are obtained under all combinations of steam generator operation condition and feed water pump operation condition. (authors)

  4. CANDU reactors. Experience and innovation

    International Nuclear Information System (INIS)

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

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

    Energy Technology Data Exchange (ETDEWEB)

    Sandon, G.

    1997-02-01

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

  6. Optimization of steam generator replacement with virtual reality modeling

    International Nuclear Information System (INIS)

    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)

  7. Steam generator local water chemistry and SCC of austenitic steel

    International Nuclear Information System (INIS)

    The titanium stabilized austenitic steel similar to the type of 321 is sensitive to the stress corrosion cracking under horizontal steam generator operating condition. SCC was observed under crevice corrosion parameters and has resulted in the transgranular or intergranular cracking at the both, components primary collectors and heat exchange tubes. The crevice environment is characterized by aggressive impurities and 'non aggressive' compounds. Sulfates and chlorides as aggressive species and silicates and alumino-silicates as 'non aggressive' species on the other hand are present in significant amount in the crevice environment under operating condition. Local water chemistry parameters were evaluated with MULTEQ Code. As input data the measured operational values of local and bulk environments have been used. The determined parameters were compared with the results of thread hole environment analyses and tube surface investigations respectively. Results of the hideout return profiles measurement showed an increase of sulfate concentration by one order of magnitude. Increase of the chloride content was not been observed, its value remains at operation levels. Examination of surface layers showed the preferential accumulation of sulfates, silicates and alumino-silicates in the deposit at tube support plates and in thread holes comparing relative to free span surfaces. The content of species in the water and deposits and the crystallographic structure of deposits correspond to MULTEQ results. Rising displacement tests were carried out with 0.5T CT specimens at a temperature 275 degrees C in the model water environment which simulated the crevice conditions. The experimental values are presented for crack growth rate versus stress intensity factor. Corrosion damage of the titanium stabilized austenitic steel is likely to be determined by the presence of sulfates and chlorides and other aggressive agents, as Cu. It is supposed that other decisive factor is the

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

    International Nuclear Information System (INIS)

    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

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

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

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    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

  14. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  16. An integrated CANDU system

    International Nuclear Information System (INIS)

    Twenty years of experience have shown that the early choices of heavy water as moderator and natural uranium as fuel imposed a discipline on CANDU design that has led to outstanding performance. The integrated structure of the industry in Canada, incorporating development, design, supply, manufacturing, and operation functions, has reinforced this performance and has provided a basis on which to continue development in the future. These same fundamental characteristics of the CANDU program open up propsects for further improvements in economy and resource utilization through increased reactor size and the development of the thorium fuel cycle

  17. Simulation-based reactor control design methodology for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Kattan, M.K.; MacBeth, M.J. [Atomic Energy of Canada Limited, Saskatoon, Saskatchewan (Canada); Chan, W.F.; Lam, K.Y. [Cassiopeia Technologies Inc., Toronto, Ontario (Canada)

    1996-07-01

    The next generation of CANDU nuclear power plant being designed by AECL is the 900 MWe CANDU 9 station. This design is based upon the Darlington CANDU nuclear power plant located in Ontario which is among the world leading nuclear power stations for highest capacity factor with the lowest operation, maintenance and administration costs in North America. Canadian-designed CANDU pressurized heavy water nuclear reactors have traditionally been world leaders in electrical power generation capacity performance. This paper introduces the CANDU 9 design initiative to use plant simulation during the design stage of the plant distributed control system (DCS), plant display system (PDS) and the control centre panels. This paper also introduces some details of the CANDU 9 DCS reactor regulating system (RRS) control application, a typical DCS partition configuration, and the interfacing of some of the software design processes that are being followed from conceptual design to final integrated design validation. A description is given of the reactor model developed specifically for use in the simulator. The CANDU 9 reactor model is a synthesis of 14 micro point-kinetic reactor models to facilitate 14 liquid zone controllers for bulk power error control, as well as zone flux tilt control. (author)

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

    International Nuclear Information System (INIS)

    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

  19. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  1. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  2. 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.

  3. Experimental model of the level dynamics in steam generator PGV-4EM

    International Nuclear Information System (INIS)

    In the paper is presented a model of the water level in steam generator PGV-4EM, obtained by active identification of the investigated plant properties. It is carried out a series of experiments with the steam generators of unit 3 at the 'Kozloduy' NPP. The experimental information (records) is processed to obtaining of continued transfer functions, with using of the MATLAB resources. (authors)

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

    OpenAIRE

    Susyadi; T. Yonomoto

    2007-01-01

    Investigation on non uniform flow behavior among U-tube in steam generator during natural circulation cooling has been conducted using RELAP5. The investigation is performed by modeling the steam generator into multi channel models, i.e. 9-tubes model. Two situations are implemented, high pressure and low pressure cases. Using partial model, the calculation simulates situation similar to the natural circulation test performed in LSTF. The imposed boundary conditions are flow rate, quality, pr...

  5. Themoeconomic optimization of triple pressure heat recovery steam generator operating parameters for combined cycle plants

    OpenAIRE

    Mohammd Mohammed S.; Petrović Milan V.

    2015-01-01

    The aim of this work is to develop a method for optimization of operating parameters of a triple pressure heat recovery steam generator. Two types of optimization: (a) thermodynamic and (b) thermoeconomic were preformed. The purpose of the thermodynamic optimization is to maximize the efficiency of the plant. The selected objective for this purpose is minimization of the exergy destruction in the heat recovery steam generator (HRSG). The purpose of the ther...

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

    International Nuclear Information System (INIS)

    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

  7. Safety-Evaluation Report related to the D2/D3 steam-generator design modification

    International Nuclear Information System (INIS)

    This Safety Evaluation Report (SER) related to the D2/D3 steam generator design modification has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The purpose of this SER is to issue the staff's evaluation of the acceptability of the design modification for both installation and full-power operation in the D2/D3 steam generators based on the Design Review Panel Report of January 1983

  8. Aging of tubes in the Krško nuclear power plant's steam generators

    OpenAIRE

    Cizelj, Leon; Androjna, Ferdo

    2015-01-01

    The paper reviews the domestic efforts devoted to the safe and reliable operation of the Krško nuclear power plant (NPP) at full power, close to the design limit of the steam generators (18% of plugged tubes) for a full decade. This includes an overview of the recent status and history of the degradation processes, discussion of repair criteria, defining the acceptable size of defects and selected results from safety analyses supporting the operation of degraded steam generator (SG) tubes. It...

  9. 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)

  10. The CANTEACH project: preserving CANDU technical knowledge

    International Nuclear Information System (INIS)

    Almost sixty years have passed since the nuclear energy venture began in Canada. Fifty years have passed since the founding of AECL. Tens of thousands of dedicated people have forged a new and successful primary energy supply. CANDU technology is well into its second century. This specialty within the world's fission technology community is quite unique, first because it was established as a separate effort very early in the history of world fission energy, and second because it grew in an isolated environment, with tight security requirements, in its early years. Commercial security rules later sustained a considerable degree of isolation. The pioneers of CANDU development have finished their work. Most of the second generation also has moved on. As yet, we cannot point to a consistent and complete record of this remarkable achievement. We, as a nuclear enterprise, have not captured the design legacy in a form that is readily accessible to the current and future generation of professionals involved with CANDU reactors, be they students, designers, operations staff, regulators, consultants or clients. This is a serious failure. Young people entering our field of study must make do with one or two textbooks and a huge collection of diverse technical papers augmented by limited-scope education and training materials. Those employed in the various parts of the nuclear industry rely mostly on a smaller set of CANDU- related documents available within their own organization; documents that sometimes are rather limited in scope. University professors often have even more limited access to in-depth and up to date information. In fact, they often depend on literature published in other countries when preparing lectures, enhanced by guest lecturers from various parts of the industry. Because CANDU was developed mostly inside Canada, few of these text materials contain useful data describing processes important to the CANDU system. For many years it has been recognized that

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

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

    International Nuclear Information System (INIS)

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

  14. 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)

    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

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

    Directory of Open Access Journals (Sweden)

    Vladimir Melikhov

    2011-01-01

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

  16. Exergy and exergoeconomic analysis of sustainable direct steam generation solar power plants

    International Nuclear Information System (INIS)

    Highlights: • Exergy and exergoeconomic analyses are presented for direct steam generation plant. • Both non-reheating and reheating by steam–steam heat exchanger are considered. • The contribution of each component to the total exergy destruction is determined. • The cost associated with exergy destruction and production cost are evaluated. • The effect of degree of reheating on the performance is presented. - Abstract: Solar direct steam generation is considered as a promising technology for steam production in thermal power generation due to high temperature levels that can be achieved compared to other technologies that use indirect steam generation. This paper demonstrates exergy and exergoeconomic analysis of commercial-size direct steam generation parabolic trough solar thermal power plant. For steam power cycles, reheating might be necessary to avoid great wetness of steam which shortens the lifetime of the turbines. Therefore, two configurations have been considered in this study; the non-reheating configuration as well as reheating by steam–steam heat exchanger. For each component, exergy and exergy-costing balance equations have been formulated based on a proper definition of fuel–product–loss. Exergy results show that particular attention should be paid to solar field, condenser, low pressure turbine and high pressure turbine (in a descendant order) as they constitute the major sources of exergy destruction. Results from exergoeconomic analysis, however, show that the condenser should be the fourth component in the order of importance after the solar field and low/high pressure turbines. Increasing the temperature at the inlet of the low pressure turbine by 100 K using steam–steam reheating is shown to result in 9.1% increase in the vapor fraction at the exit of turbine. This increase in steam quality, however, would be achieved by drop less than 1.5% in thermal and exergetic efficiencies, and about 2% increase in cost of electricity

  17. CANDU, building the future

    Energy Technology Data Exchange (ETDEWEB)

    Stern, F. [Stern Laboratories (Canada)

    1997-07-01

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

  18. CANDU market prospects

    International Nuclear Information System (INIS)

    This 1994 survey of prospective markets for CANDU reactors discusses prospects in Turkey, Thailand, the Philippines, Korea, Indonesia, China and Egypt, and other opportunities, such as in fuel cycles and nuclear safety. It was concluded that foreign partners would be needed to help with financing

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  20. 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.

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

    Science.gov (United States)

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

    1983-05-01

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

  2. CANDU safety analysis system establishment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo; Rhee, B. W.; Park, J. H.; Kim, H. T.; Choi, H. B.; Shim, J. I.; Yoon, C.; Yang, M. K

    2002-03-01

    To develop CANDU safety analysis system, methodology, and assessment technology, GAIs from CNSC and GSIs drived by IAEA are summarized. Furthermore, the following safety items are investigated in the present study. - It is intended to secure credibility of the void reactivity in the stage of nuclear design and analysis. The measurement data concerned with the void reactivity were reviewed and used to assess the physics code such as POWDERPUFS-V/RFSP, and the lattice code such as WIMS-AECL and MCNP-4B. - Reviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc. were examined. - The development of 3D CFD transient analysis model has been performed to predict local subcooling of the moderator in the vicinity of Calandria tubes in a CANDU-6 reactor in the case of Large LOCA transient. - The trip coverage analysis methodology based on CATHENA code is developed. The simulation of real plant transient showed good agreement. The trip coverage map was generated successfully for two typical depressurization and pressurization event. - The multi-dimensional analysis methodology for hydrogen distribution and hydrogen burning phenomena in PHWR containment is developed using GOTHIC code. The multi-dimensional analysis predicts the local hydrogen behaviour compared to the lumped parameter model.

  3. Development of a nuclear steam generator system for gas-cooled reactors for application in oil sands extraction

    International Nuclear Information System (INIS)

    Canada has vast energy reserves in the Oil Sands regions of Alberta and Saskatchewan. Present extraction technologies, such as strip mining, where oil deposits are close to the surface, and Steam Assisted Gravity Drainage (SAGD) technologies for deeper deposits consume significant amounts of energy to produce the bitumen and upgraded synthetic crude oil. Studies have been performed to assess the feasibility of using nuclear reactors as primary energy sources to produce, in particular the steam required for the SAGD deeper deposit extraction process. Presently available reactors fall short of meeting the requirements, in two areas: the steam produced in a 'standard' reactor is too low in pressure and temperature for the SAGD process. Requirements can be for steam as high as 12MPa pressure with superheat; and, 'standard' reactors are too large in total output. Ideally, reactors of output in the range of 400 to 500 MWth, in modules are better suited to Oil Sands applications. The above two requirements can be met using gas-cooled reactors. Generally, newer generation gas-cooled reactors have been designed for power generation, using Brayton Cycle gas turbines run directly from the heated reactor coolant (helium). Where secondary steam is required, heat recovery steam generators have been used. In this paper, a steam generating system is described which uses the high temperature helium from the reactor directly for steam generation purposes, with sufficient quantities of steam produced to allow for SAGD steam injection, power generation using a steam turbine-generator, and with potential secondary energy supply for other purposes such as hydrogen production for upgrading, and environmental remediation processes. It is assumed that the reactors will be in one central location, run by a utility type organization, providing process steam and electricity to surrounding Oil Sands projects, so steam produced is at very high pressure (12 MPa), with superheat, in order to

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

    Institute of Scientific and Technical Information of China (English)

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

    2005-01-01

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

  5. Ultrasonic wall thickness gauging for ferritic steam generator tubing as an in-service inspection tool

    International Nuclear Information System (INIS)

    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

  6. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  7. Parameter optimization of heat recovery steam generation for hyndai engine h25/33

    OpenAIRE

    MARCHENKO ANDRII PETROVYCH; ALI ADEL HAMZAH; OMAR ADEL HAMZAH

    2016-01-01

    Conducted experimental studies of thermodynamic parameters changes in working environments in Hyundai engine H25/33 when the engine is operating at different times of the year. Obtained regressional dependence to calculate the parameters of working environment in the range of ambient temperature changes from 0 to 40 °C. Based the possibility of use of ICE cooling water in the heat recovery steam generator in its appropriate treatment. Formed mathematical model of the heat recovery steam gener...

  8. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  11. 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.

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

    International Nuclear Information System (INIS)

    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

  13. Proceedings of the NEA/CSNI-UNIPEDE Specialist Meeting on Operating Experience with Steam Generators

    International Nuclear Information System (INIS)

    The long history of operating experience with pressurized water reactors has indicated that the steam generators are of primary importance in nuclear power plant design and operation; this is furthermore confirmed by analyzing the data of the Incident Reporting System (IRS). It is for this reason that the OECD/NEA Committee on the Safety of Nuclear Installations organizes, in cooperation with UNIPEDE, a Specialist Meeting on 'Operating Experience with Steam Generators'. This Specialist Meeting, held in Brussels, Belgium, in September 1991, is hosted by the Belgian Government and AIB-Vincotte Nuclear. In addition to being a follow-up to the October 1984 meeting (organized by the CSNI and UNIPEDE in Stockholm, Sweden), this Meeting reviews the current state-of-the-art of steam generator technology thus providing a forum for the exchange of related experience in operation, inspection, maintenance, repair, modifications, replacement, and licensing requirements pertaining to steam generators. Forty-seven papers are presented in eight sessions entitled: Operating Experience (two sessions), Structural Integrity and Licensing Issues, Analysis and Prediction of Degradation Mechanisms, Inservice Inspection Methods, Preventive and Corrective Actions (two sessions) and Replacement of Steam Generators. There are furthermore two panel sessions entitled 'Observed Degradation Mechanisms and Licensing Positions', and 'Inspection, Repair and Replacement Strategies'. These proceedings consist of a compilation of the papers presented at the Meeting, which is attended by more than one hundred and fifty participants from fifteen countries and several international organisations

  14. Numerical analysis of APR1400 Steam Generator by CUPID/MARS heat structure coupling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Ryong Lee; Lee, Seung Jun; Pakr, Ik Kyu; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of); Cho, Hyoung Kyu [Seoul National University, Seoul (Korea, Republic of)

    2015-05-15

    To design and analyze steam generators, many computer codes have been developed and used around the world. In this study, the coupled CUPID and MARS code was used for the simulation of boiler side of the PWR steam generator. This paper presents the description of the coupling method, validation for porous media approach against the rod bundle experiment and the preliminary simulation results of PWR steam generator using the coupled code. In the present study, the multi-scale thermal-hydraulic analysis method using the coupled CUPID/MARS code was applied for the simulation of the steam generator. The primary side of the steam generator and other RCS was simulated by MARS and the secondary side was calculated by CUPID with porous media approach. For coupled simulation, the porous medium was applied in order to take into account the effect of the U-tube bundle and other supporting structure which play a role to be a flow resistance. More realistic physical model such as moisture separator slug behavior should be developed for the near future. The application of the coupled simulation should be extended to the accident scenario.

  15. Design and fabrication of steam generators (superheaters) for the prototype fast breeder reactor 'MONJU'

    International Nuclear Information System (INIS)

    In liquid metal-cooled fast breeder reactors, steam generators are one of the important equipments, and emphasis has been placed on their development in various countries in the world. Also in Japan, centering around the Power Reactor and Nuclear Fuel Development Corp., the research and development in the wide range from the fundamentals on heat transfer and flow, materials and strength for steam generators to the manufacture, operation and various tests of large mock-ups including a 50 MW steam generator have been carried out. Further, as for the manufacture and inspection, the improvement of the method of welding tubes and tube plates, the adoption of a fine focus X-ray inspection apparatus and others were carried out. Moreover, as the maintenance technique, the ultrasonic flaw detection probes for the heating tubes were developed. The steam generators (superheaters) for the FBR 'Monju' power station are the heat exchangers of helical coil tube-shell type using SUS 321 steel as the heating tube material. Based on the results of these research and development, the design and manufacture of these superheaters and their installation in the reactor auxiliary building of the FBR 'Monju' power station were completed. The outline of the design, the research and development and the manufacture of the steam generators (superheaters) are reported. (K.I.)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  18. Wasteless combined aggregate-coal-fired steam-generator/melting-converter.

    Science.gov (United States)

    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 concrete and as additives in the production of cement, bricks and other building materials. PMID:12781221

  19. Assessment of autogenous Type 410S stainless steel welds in replacement steam generator tube support structures

    International Nuclear Information System (INIS)

    To eliminate fretting wear caused by flow-induced vibration in recirculating steam generators, tubes are separated from each other by tube support lattice bars. In the U-bend portion of the tube bundle, rows of tubes are separated by fan bars that radiate from collector bars located in the straight-leg portion of the steam generator. The replacement steam generators constructed by Babcock ampersand Wilcox International use Type 410S stainless steel with a specified maximum hardness of Rb 95 for tube support lattice bars, collector bars, and fan bars. An autogenous weld is used to join the fan bar to the collector bar. Corrosion tests were conducted to assess the stress corrosion cracking (SCC) susceptibility of welded type 410S stainless steel. These tests included constant-extension-rate (CERT) tests and long-term immersion tests on 410S in various welded and heat-treated conditions. The results of this test program demonstrate that, when highly stressed, the as-welded 410S weld joints are susceptible to SCC in steam generator environments. However, highly stressed 410S autogenous welds given a post-weld heat treatment were not susceptible to SCC even under faulted steam generator operating conditions

  20. Advances in production of realistic cracks to NDT development and qualification purposes of steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Virkkunen, I.; Kemppainen, M. [Truflaw Ltd., Espoo (Finland); Tchilian, J.-M. [AREVA Nuclear Power Plant Sector, Saskatoon, Saskatchewan (Canada); Martens, J. [AREVA NP Intercontrole (France)

    2009-07-01

    Realistic defects are needed for steam generator tube inspections when developing new NDT methods or assessing the performance and reliability of methods and procedures used. Furthermore, realistic defects give the most reliable results in assessing service-related reliability of steam generator tubes by, for example, burst or leak tests. It is crucial to have representative defects as the defect characteristics has marked effect on the results both in NDE, burst and leak tests. Representativeness should be to the actual service-induced defects, and the evaluation should be based on the essential defect characteristics. In this paper real world application cases are presented about crack production to steam generator tubes. Crack production technique used is based on controlled thermal fatigue process creating natural cracks. Such cracks have been produced in Alloy 690 and austenitic stainless steel steam generator tubes. These cracks have been used, for example, for advanced NDT qualification purposes of a new build nuclear power plant. Paper presents results of the destructive tests performed after validation tests of the crack manufacturing in the Alloy 690 and austenitic stainless steel. These results are shown for both of the materials with measured essential crack characteristics. In addition to metallographic analysis, the paper presents the results of performed NDT inspections for the Alloy 690. Results have been obtained with an advanced inspection technique developed and used for today's inspections of steam generator tubes in nuclear power plants. (author)

  1. Non-intrusive downcomer flow measurements: a means of monitoring steam generator performance

    International Nuclear Information System (INIS)

    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)

  2. Particle Swarm Optimization to the U-tube steam generator in the nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, Wesam Zakaria, E-mail: mimi9_m@yahoo.com

    2014-12-15

    Highlights: • We establish stability mathematical model of steam generator and reactor core. • We propose a new Particle Swarm Optimization algorithm. • The algorithm can overcome premature phenomenon and has a high search precision. • Optimal weight of steam generator is 15.1% less than the original. • Sensitivity analysis and optimal design provide reference for steam generator design. - Abstract: This paper, proposed an improved Particle Swarm Optimization approach for optimize a U-tube steam generator mathematical model. The UTSG is one of the most important component related to safety of most of the pressurized water reactor. The purpose of this article is to present an approach to optimization in which every target is considered as a separate objective to be optimized. Multi-objective optimization is a powerful tool for resolving conflicting objectives in engineering design and numerous other fields. One approach to solve multi-objective optimization problems is the non-dominated sorting Particle Swarm Optimization. PSO was applied in regarding the choice of the time intervals for the periodic testing of the model of the steam generator.

  3. Assessment of steam generator slagging and slag removal with water guns

    Energy Technology Data Exchange (ETDEWEB)

    Bude, F.; Schettler, H.; Weidlich. H.G.

    1983-09-01

    This paper discusses combustion parameters and slag buildup on heating surfaces in brown coal fired steam generators. At the Berlin steam generator plant (GDR), the influence of slagging on heat transfer in a combustion chamber is calculated according to the method of A.M. Gurwitsch (1950). Combustion properties are further assessed with the approximation method according to B. Weiser (1976), for which nomograms have been developed. For the prediction of brown coal slagging behavior, four ash and slag analyses are described and compared. Slag removal from heating surfaces in various 815 t/h steam generators is carried out by 36 back-acting blowers, distributed over all 4 sides of the combustion chamber. These blowers, however, are not capable of complete cleaning of the heating surfaces. An automated and electrically operated water jet gun (type EWL 1) was, therefore, developed. It can be installed in all steam generator sizes including those with very large combustion chambers. The performance of the EWL 1 is evaluated with graphs. Four water guns are sufficient for periodic combustion chamber cleaning of 1,000 t/h steam generators. (9 refs.)

  4. Morpholine decomposition products in the secondary cycle of CANDU-PHWR plants

    International Nuclear Information System (INIS)

    Trace amounts of organic compounds resulting from the decomposition of morpholine additive used for erosin-corrosion control were determined in CANDU-PHWR steam-condensate cycles. Most of the morpholine breakdown products (2-(2-aminoethoxy) ethanol, ethanolamine, ammonia, methylamine, ethylamine, ethylene glycol, glycolic and acetic acids) identified during thermal-decomposition tests in the laboratory were detected in the steam-condensate cycles investigated, thus confirming the proposed morpholine reaction scheme. Their relative concentration in cycle components is affected by the use of condensate polishing, the presence of contaminants in the feeding morpholine solutions, the presence of non-ionic or weakly-ionized organic matter in the makeup water, and the organic contaminants introduced into the cycle by condenser leaks. Comparison of the analytical results before and after feeding the morpholine into the cycle of one of the plants investigated confirms that the thermal decomposition of this additive contributes significantly to the formation of glycolic and acetic acids, reported to be responsible for a cation conductivity increase of about 0.009 and 0.0675 mS/m in steam-generator blowdowns and moisture separator/reheater drains, respectively. Finally, an important fraction of these breakdown products is removed by the blowdown of the steam generator, the deaerator and the condensate polisher. (orig.)

  5. CANDU at the crossroads

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-11-01

    ''Ready for the challenge of the 90s'' was the theme of this year's gathering of the Canadian Nuclear Association held in Toronto, 3-6 June. What that challenge really entails is whether the CANDU system will survive as the last remaining alternative to the light water reactor in the world reactor market, or whether it will decline into oblivion along with the Advanced Gas Cooled reactor and so many other technically excellent systems which have fallen along the way. The fate of the CANDU system will not be determined by its technical merits, nor by its impeccable safety record. It will be determined by public perceptions and by the deliberations of an Environmental Assessment Panel established by the Government of Ontario. The debate at the Association meeting is reported. (author).

  6. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-05-01

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

  8. Exergetic Optimization of the Heat Recovery Steam Generators by Imposing the Total Heat Transfer Area

    Directory of Open Access Journals (Sweden)

    Michel Feidt

    2004-09-01

    Full Text Available The paper presents an original and fast method for the heat recovery steam generator (HRSG exergetic optimization. The objective is maximizing the exergy transfer to the water / steam circuit. The proposed approach, different from the classical method that fixes the pinch point, is essentially thermodynamic but it considers also the economics by imposing the total heat transfer area of HRSG. The HRSG may have one or two steam pressures, without reheat. The input data from the gas turbine are: the mass flow rate, the temperature and the molar composition of flue gases. The results are the optimum pressures of the superheated steam. The numerical computations were realized in Delphi programming utility. The obtained results are in agreement with the recent literature.

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

    OpenAIRE

    Puchalski Bartosz; Duzinkiewicz Kazimierz; Rutkowski Tomasz

    2015-01-01

    In the paper, analysis of multi-region fuzzy logic controller with local PID controllers for steam generator of pressurized water reactor (PWR) working in wide range of thermal power changes is presented. The U-tube steam generator has a nonlinear dynamics depending on thermal power transferred from coolant of the primary loop of the PWR plant. Control of water level in the steam generator conducted by a traditional PID controller which is designed for nominal power level of the nuclear react...

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

    OpenAIRE

    Laila Khalilzadeh Ganjali-khani; Farid Sheikholeslam; Homayoun Mahdavi-Nasab

    2013-01-01

    One of the most effective strategies for steam generator efficiency enhancement is to improve the control system. For such an improvement, it is essential to have an accurate model for the steam generator of power plant. In this paper, an industrial steam generator is considered as a nonlinear multivariable system for identification. An important step in nonlinear system identification is the development of a nonlinear model. In recent years, artificial neural networks have been successfully ...

  11. 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.

  12. 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.

  13. Study of the vibrations induced by two-phase flow in steam generator: measurement of void fraction in a two-phase flow

    International Nuclear Information System (INIS)

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

  14. Pressure and temperature straining of liquid sodium heated sectional steam generators in accident situations

    International Nuclear Information System (INIS)

    The results are presented and discussed of the calculation of the pressure and temperature strain on a sectional steam generator fuelled by liquid sodium and having an output of 30 MW(th). In all sections of the steam generator is the ratio of section length to diameter much greater than 1. In the superheater sections this ratio is 63, in the evaporator and economizer it is 45. The flow rate, pressure, temperature and specific weight of the sodium are taken as constant. For a major leak of water into the sodium in case of a total failure of the heat transfer pipe a typical course is given of the temperature of water/sodium reaction products, the course of wall temperature along the section length filled with reaction products, and the course of the pressure load on the sections and on the steam generator at given parameters (sodium: inlet temperature 445 degC, outlet temperature 292 degC, mean pressure 0.2 MPa; water, steam: inlet temperature 190 degC, outlet temperature 435 degC, mean pressure 8.5 MPa) where the maximum weight flow of water into sodium is msub(max)=3.2 kg/s. An evaluation is made with regard to the project design and the design of the steam generator. (B.S.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Leak locating in Bruce NGS-A steam generators using gas tracer techniques

    International Nuclear Information System (INIS)

    In 1981, Ontario Hydro requested development of a leak locating technique capable of locating a 0.5 kg.h-1 heavy water leak within 72 h of access to the steam generator head. A gas tracer technique has been developed to the point where it can now be used in the station to locate such leaks. The technique consists of pressurizing the shell side to 450 kPa with a sulfur hexafluoride air mixture and sampling on the tube side. To speed up the search, a multi-tube sampler is used to sniff a number of tubes simultaneously. The technique as proposed requires a man to enter the steam generator head, but can be adapted for use from outside the steam generator head. The development equipment and procedures required to complete a search are described

  18. Corrosion deposits removal from Kozloduy NPP VVER-440 steam generator tubing by chemical cleaning

    International Nuclear Information System (INIS)

    A strict control of primary and secondary circuits metal equipment corrosion of VVER-440 Kozloduy NPP units has been performed for the whole period of operation. This is carried out following a specific program including visual inspection and chemical analysis of equipment corrosion deposits. During their migration, the corrosion products deposit on the metal surface in the so-called standstill zones. One of these is the steam generator. The process results in: deterioration of thermal exchange; deterioration of corrosion conditions under deposits corrosion, pitting corrosion, etc. Using quantity deposits data and deposits chemical consistence, chemical cleaning of steam generator surfaces is performed. Decision for such chemical treatment of secondary circuit equipment is taken when the amount of deposits on the steam generator tubing is greater than 150 g/m2. This limit is based on operational experience and manufacturer requirements. (R.P.)

  19. Development of a high pressure water jet nozzle for steam generator lancing system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, N. H.; Jeong, W. T.; Son, S. Y.; Choi, Y. S. [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Nho, B. J. [Chonbuk Univ., Jeonju (Korea, Republic of); Park, Y. S. [Hanboi ENG, Taejon (Korea, Republic of)

    2003-10-01

    Metal-oxide sludge accumulates on the tube sheet of nuclear steam generators as time passes. To prevent degradation of thermal efficiency of nuclear steam generators, it is recommended to clean the tube sheet and the tubes. It is important that efficiency of lancing of steam generators in nuclear power plants depends on nozzle performance. The aspect ratio, among many factors affecting the performance of a nozzle, plays a major role in determining the outer flow pattern and nozzle performance. So in this study, some flow characteristics with the variation of nozzle aspect ratios have been experimentally investigated. By this experiments, the increase of aspect ratio causes decrease of water jet energy. As a result, it was obviously concluded that the nozzle performance depends on the aspect ratio of nozzle.

  20. Dynamical instabilities in steam generators. Numerical simulation, experimental analysis and theoretical interpretation

    International Nuclear Information System (INIS)

    The dynamical instability phenomena, which may have an influence on the most important steam generator flow parameters, are studied. The instabilities involved in two-phase flows are underlined. The oscillation phenomena conducting to dynamical instabilities in steam generators are considered. A model for the numerical resolution of four one-dimensional equations is presented. In the model, thermal and mechanical non-equilibrium effects are taken into account. Moreover, the model is integrated to the SICLE thermohydraulic software. The results of the experiments carried out on two steam generator prototypes are analyzed. A theoretical model, which involves a small number of parameters, relating the dynamics of the boiling front and the inlet velocity is obtained

  1. Study on temperature conditions of steam generating surface operation on burnout

    International Nuclear Information System (INIS)

    Insufficient heat transfer (second-order burnout) in a direct-flow steam generator has been studied. Vapor contents for which the insufficient heat exchange arises, wall temperature oscillations and heat exchange in the hypercrisis range have been determined. The experimental data have been analyzed for pressures from 7 to 18 MPa, water flow rates from 350 to 1000 kg/cmxs, heat fluxes from 0.2 to 0.6 MW/m2. The steam generating wall temperature oscillations have been investigated on a direct-pipe model of a direct-flow steam generator of the ''pipe-in-pipe'' type heated by sodium. No unique dependence of the oscillation intensity on the specific heat flux has been established. A simple formula to calculate hypercrisis heat transfer coefficients with an accuracy of 25% has been obtained

  2. Intermediate leak protection/automatic shutdown for B and W helical coil steam generator

    International Nuclear Information System (INIS)

    The report summarizes a follow-on study to the multi-tiered Intermediate Leak/Automatic Shutdown System report. It makes the automatic shutdown system specific to the Babcock and Wilcox (B and W) helical coil steam generator and to the Large Development LMFBR Plant. Threshold leak criteria specific to this steam generator design are developed, and performance predictions are presented for a multi-tier intermediate leak, automatic shutdown system applied to this unit. Preliminary performance predictions for application to the helical coil steam generator were given in the referenced report; for the most part, these predictions have been confirmed. The importance of including a cover gas hydrogen meter in this unit is demonstrated by calculation of a response time one-fifth that of an in-sodium meter at hot standby and refueling conditions

  3. 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.

  4. Laser welded sleeving - a proven technology for steam generator life enhancement

    International Nuclear Information System (INIS)

    Laser welded sleeving was performed for the first time in the United States in April 1992 at the J.M. Farley Nuclear Plant Unit 2. In all, 68 tube support plate sleeves and 30 tubesheet sleeves were installed in two steam generators. This was followed by a larger sleeving campaign in the plant's Unit 1 steam generators in October 1992 when 148 tube support plate sleeves and 46 tubesheet sleeves were installed in three steam generators. The successful implementation of this new technology at Farley provides the industry with a field proven and effective option to repair steam generator tubes and maintain operating plant performance. The laser welding was performed using a fiber optic delivery system to transmit light energy from a pulsed solid state laser located outside containment to the weld head which could be positioned remotely in the tubesheet and as high as the sixth support plate in the steam generator. The sleeve material was thermally treated Alloy 690 (UNS 06690). The joint design was a partial penetration, autogenous weld. All free-span welds, namely the support plate sleeve welds and the tubesheet sleeve upper weld, were thermally stress relieved to enhance stress corrosion life. Those welds were also required to pass a stringent ultrasonic test examination. All the processes for laser welded sleeving were performed remotely using the Westinghouse steam generator service robot, ROSA III. The Farley campaigns showed that laser welded sleeving offers a high degree of process control not found with other methods and produces welds that can be fully inspected by ultrasonic examination. They also demonstrated the field hardiness of the sophisticated laser welding system. 5 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  7. Clinch River Breeder Reactor Plant steam generator: FEW tube test model post test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) is part of an extensive testing program being carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. The problems encountered with the mechanical features of the FTT model design which led to premature test termination and the results of the post-test examination are described

  8. An integrated evaluation of the performance effects of steam generator tube plugging

    International Nuclear Information System (INIS)

    The integrity of the walls of a small number of steam generator tubes can degrade, with time, during normal operation of a Pressurized Water Reactor. In order to avoid the potential for unacceptable primary-to-secondary leakage, these steam generator tubes must be plugged. This paper presents an integrated evaluation of the impact of tube plugging on plant performance. Such an evaluation is recommended for determining the number of tubes that can be plugged without a significant adverse impact on plant steady state design performance and a large reduction in plant safety analyses margins

  9. Evaluation of PWR steam generator water hammer. Final technical report, June 1, 1976--December 31, 1976

    International Nuclear Information System (INIS)

    An investigation of waterhammer in the main feedwater piping of PWR steam generators due to water slugs formed in the steam generator feedring is reported. The relevant evidence from PWR operation and testing is compiled and summarized. The state-of-the-art of analysis of related phenomena is reviewed. Original exploratory modeling experiments at 1/10 and 1/4 scale are reported. Bounding analyses of the behavior are performed and several key phenomena have been identified for the first time. Recommendations to the Nuclear Regulatory Commission are made

  10. 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

  11. Investigations on steam generator tube samples of Inconel 600 after long term service in the FDR

    Energy Technology Data Exchange (ETDEWEB)

    Milferstaedt, D.; Schmelzer, F.; Schmitt, F.J.

    1973-12-31

    Before the startup of the Otto Hahn Reactor, pipe samples of steam generator material were attached to the steam generator. These pipe samples werc tested for corrosion behavior at given time intervals. After the first core change the samples were dismantled and investigated. From these investigations it was established that no damage or an anomalous corrosion behavior was detected. They are coated with a very thin blue--green solidly adhering oxide layer. A sharp decrease of the stability value was not established. (JSR)

  12. Motivation of parametric studies. French recommendations concerning surveillance in exploitation. Program of steam generator inspection

    International Nuclear Information System (INIS)

    The PISC/2 program deals with parametric studies; the different parameters that may have an influence on the defect detection and sizing are the following ones: effect of the defect characteristic and sitting, effect of the measuring system characteristics, and possible utilization of electro-magnetic techniques. A second part of this report concerns the French recommendations concerning surveillance while the power plant is operating. Finally the PISC 3 program is presented; it will deal with steam generator control: experimental evaluation of the performance of the tests applied to the nuclear power plant steam generator tubes

  13. Treatment and disposal of steam generator and heat exchanger chemical cleaning wastes

    International Nuclear Information System (INIS)

    Wet air oxidation was effective in reducing the organic loading of Ontario Hydro's EDTA-based steam generator cleaning wastes and the organic acid formulation used for heat exchanger chemical cleaning. Destruction of the complexing agents resulted in direct precipitation of iron from the waste steam generator magnetite solvent and from the heat exchanger cleaning waste. The oxidized liquors contain lower molecular weight organic acids, ammonia and amines, suitable for secondary biological treatment. The oxidized copper waste requires further treatment to reduce dissolved copper levels prior to biological digestion. A preliminary evaluation of UV and ozone degradation of these wastes showed less promise than wet air oxidation. 24 refs., 1 fig., 4 tabs

  14. Review of tube support plate analysis for steam generators of Millstone Unit II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Magnetite growth in steam generator tube support plates was observed in the Millstone Unit II Nuclear Power Plant. If growth is allowed to continue, the tube may eventually fail resulting from plate shifting and the squeezing action of the growing magnetite. The corrective actions undertaken by the Northeast Nuclear Energy Company (NNECO) for this effect have been summarized in a report submitted to the U.S. Nuclear Regulatory Commission (NRC) entitled, Millstone Unit No. II Steam Generator Repairs and Corrective Actions, Docket No. 50-336. The analytical study part of this report is reviewed here, and conclusions and recommendations for further research are given

  15. Digital simulation test system for steam generator multi-purpose thermalhydraulic test facility

    International Nuclear Information System (INIS)

    Taking advantage of NPA2000's software and hardware environment. This project was to develop a full scope digital simulation system of Steam Generator Thermalhydraulic Test Facility. The research work on steam generator was done, including verifying steady state parameters and transient thermalhydraulic processes in advance, predicting the operation scenarios under the incidental conditions and studying the emergency acts, etc. On the other hand, the real test recorded data could greatly help modify and improve the simulation model to make it more practical and to achieve higher fidelity. The enhanced simulation system would be a good supplement to the real facility with those accident or malfunction conditions and destructive conditions

  16. Design of an adaptive pole assignment controller for the water level of steam generators

    International Nuclear Information System (INIS)

    In this thesis an adaptive observer is designed and a pole assignment technique is applied in order to accomplish a satisfactory automatic control of steam generators from zero to full power. The change of the water level of a steam generator is caused by three effects; mass capacity, swelling and shrinking, and mechanical oscillations. The knowledge of the water level contributions caused by these effects will be helpful in controlling the water level. The state observer is designed in order to use state feedback. The obvious result of introducing state feedback is to change the undesired open-loop system into the desired overall closed-loop system. Since a nuclear steam generator is controllable and observable, it is possible to design a state observer. Also, since a steam generator is subject to parameter variation according to the change of operating conditions, an adaptive observer must be used. The adaptive observer estimates the parameters and states of the steam generator simultaneously. A fourth-order linear model is presented and, on the basis of this model, an adaptive observer is designed. Since an implicit-type adaptive observer is applied, a state reconstruction process and a parameter adaptation one are separated and system inputs and outputs are unnecessary to be bounded for the stability. The time-varying problem of the steam generator is resolved by estimating at every time step the parameters which change according to the operating conditions. A pole assignment controller is derived on the basis of the adaptive observer. The characteristics of the overall closed-loop control system can be expressed in terms of its assigned poles. The troublesome tuning procedure of the conventional P-I controller is reduced to the determination of the desired poles only. The proposed algorithm is compared with the P-I controller through numerical simulations. Also, the adaptive pole assignment controller is studied experimentally by implementing it to the mock

  17. Ultrasonic inspection of liquid-metal fast breeder reactor steam generator duplex tubing

    International Nuclear Information System (INIS)

    Two ultrasonic inspections of the Experimental Breeder Reactor II steam generator duplex tubing have been completed. Inspections performed on one evaporator in 1976 provided baseline data, and a subsequent inspection in 1978 revealed no change in tube condition. With the completion of the 1978 inspection, all available tubes in one evaporator have been inspected. The steam generator contains duplex tubes fabricated from 2 1/4 Cr-1 Mo ferritic steel. Access to the bore (water) side of the tubes was gained through the steam outlet piping. The inspection included a complete volumertic (100% of the tube material) examination, measurement of wall thickness, and evaluation of the condition of the braze bonding the two walls of the tube together. The test equipment was routinely calibrated against a standard containing artificial flaws. Artificial flaws as small as 1.6 mm long x 0.25 mm deep were readily detected

  18. High-definition radiography of tube-to-tubesheet welds of steam generator of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    In the steam generator of the Prototype Fast Breeder Reactor (PFBR), steam is generated by the transfer of heat from secondary sodium to water. Due to the inherent dangers of sodium-water reaction, the integrity of weld joints separating sodium and water/steam is of paramount importance. This is particularly true and very important for the tube-to-tubesheet joints. This paper discusses the use of projective magnification technique by microfocal radiography for the quality evaluation and optimisation of the welding parameters of such small tube-to-tubesheet welds of the steam generator of PFBR. (author)

  19. Assessment of LOCA with loss of class IV power for CANDU-6 reactors using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Recently, there is an effort to improve the accuracy and reality in the transient simulation of nuclear power plants. In the prediction of the system transient, the system code simulates the system transient using the power transient curve predicted from the reactor core physics code. However, the pre-calculated power curve could not adequately predict the behavior of power distribution during transient since the coolant density change has influence on the power shape due to the change of the void reactivity. Therefore, the consolidation between the reactor core physics code and the system thermal-hydraulic code takes into consideration to predict more accurate and realistic for the transient simulation. In this regard, there are two codes are developed to assess the safety of CANDU reactor. RELAP-CANDU is a thermal-hydraulic system code for CANDU reactors developed on the basis of RELAP5/MOD3 in such a way to modify inside model for simulating the thermal-hydraulic characteristics of horizontal type reactors. SCAN (SNU CANDU-PHWR Neutronics) is a three dimensional neutronics nodal code to simulate the core physics characteristics for CANDU reactors. To couple SCAN code with RELAP-CANDU code, SCAN code was improved as a spatial kinetics calculation module in such a way to generate a SCAN DLL (dynamic linked library version of SCAN). The coupled code system, RELAP-CANDU/SCAN, enables real-time feedback calculations between thermal-hydraulic variables of RELAP-CANDU and reactor powers of SCAN. To verify the reliability of RELAP-CANDU/SCAN coupled code system, an assessment of 40% reactor inlet header (RIH) break loss of coolant accident (LOCA) with loss of Class IV power (LOP) for Wolsong Unit 2 conducted using RELAP/CANDU-SCAN coupled system. The LOCA with LOP is one of GAI (Generic Action Items) for CANDU reactors issued by CNSC (Canadian Nuclear Safety Commission) and IAEA (International Atomic Energy Agency)

  20. Asbestos exposure in a steam-electric generating plant

    Energy Technology Data Exchange (ETDEWEB)

    Scansetti, G.; Pira, E.; Botta, G.C.; Turbiglio, M.; Piolatto, G. (Turin Univ. (Italy). Inst. of Occupational Health)

    1993-12-01

    A study on asbestos risk in an old multi-fuel-fired steam-electric power station was carried out. In spite of the presence of large amounts of asbestos-containing materials (20 km of asbestos insulated pipes), the mean airborne concentration of asbestos was as low as 1.55 fibres 1.[sup -1](SD 2.05) under normal operating conditions. Much higher concentrations may obviously occur during maintenance or renovation operations. Man-made mineral fibres (MMMF) were detected only occasionally in some samples. Three non-consecutive sputum samples were collected for all the 521 workers included in the study: 3.1% had asbestos bodies (AB), but in no case were there more than four AB per gramme sputum. Small opacities, in most cases irregular of mixed type, were presented in 15 out of 470 radiograms of acceptable quality (3.2%). No AB were found in these cases. Pleural changes were less common: two out of five bilateral cases had AB in the sputum. It is concluded that repeated AB counts in the sputum turned out to be more useful than the search of pleural abnormalities by traditional postero-anterior (PA) view in detecting the signs of low asbestos exposure. (Author)

  1. Industrial process heat from CANDU reactors

    International Nuclear Information System (INIS)

    It has been demonstrated on a large scale that CANDU reactors can produce industrial process steam as well as electricity, reliably and economically. The advantages of cogeneration have led to the concept of an Industrial Energy Park adjacent to the Bruce Nuclear Power Development in the province of Ontario. For steam demands between 300,000 and 500,00 lb/h (38-63 kg/s) and an annual load factor of 80%, the estimated cost of nuclear steam at the Bruce site boundary is $3.21/MBtu ($3.04GJ), which is at least 30% cheaper than oil-fired steam at the same site. The most promising near term application of nuclear heat is likely to be found within the energy-intensive chemical industry. Nuclear energy can substitute for imported oil and coal in the eastern provinces if the price remains competitive, but low cost coal and gas in the western provinces may induce energy-intensive industries to locate near those sources of energy. In the long term it may be feasible to use nuclear heat for the mining and extraction of oil from the Alberta tar sands. (auth)

  2. Frequency and distribution of leakages in steam generators of gas-cooled reactors

    International Nuclear Information System (INIS)

    In gas cooled reactors with graphitic primary circuit structures - such as HTR, AGR or Magnox - the water ingress is an event of great safety concern. Water or steam entering the primary circuit react with the hot graphite and carbon-oxide and hydrogen are produced. As the most important initiating event a leak in a steam generator must be taken into account. From the safety point of view as well as for availability reasons it is necessary to construct reliable boilers. Thus the occurrence of a boiler leak should be a rare event. In the context of a probabilistic safety study for an HTR-Project much effort was invested to get information about the frequency and the size distribution of tube failures in steam generators of gas cooled reactors. The main data base was the boiler tube failure statistics of United Kingdom gas cooled reactors. The data were selected and applied to a modern HTR steam generator design. A review of the data showed that the failure frequency is not connected with the load level (pressures, temperatures) or with the geometric size of the heating surface of the boiler. Design, construction, fabrication, examination and operation conditions have the greatest influence an the failure frequency but they are practically not to be quantified. The typical leak develops from smallest size. By erosion effects of the entering water or steam it is enlarged to perhaps some mm2, then usually it is detected by moisture monitors. Sudden tube breaks were not reported in the investigated period. As a rule boiler leaks in gas cooled reactors are much more, rare then leaks in steam generators of light water reactors and fossil fired boilers. (author)

  3. The development and application of overheating failure model of FBR steam generator tubes

    International Nuclear Information System (INIS)

    The following items have been studied to evaluate overheating failure of FBR steam generator heat transfer tubes: 1) To establish a structural integrity analysis method, 2) To improve and validate blow down analytical method, 3) To quantitatively validate the entire overheating analysis model by sodium water reaction data. Based on the above studies, the analytical method was applied to PFR superheater leak event and the Monju steam generator accidental analysis. The followings were quantitatively shown through the analysis: 1. The most important cause that multi-tube failure occurred in the 1987 PFR superheater-2 leak is that the superheater did not equip a fast steam dump system at the time of the leak event. 2. Overheating failure will not occur under any operational conditions of Monju in both steady state and transient phases such as water/steam blow-down. 3. Although safety margin becomes small when the water/steam flow rate becomes small during the blow-down, the modification of the plant such as hastening blow-down by equipping more relief valves will drastically improve the safety margin. (J.P.N.)

  4. Theoretical and experimental study of a reactive steam jet in molten sodium. Application to the wastage of steam generators of FBR power plants

    International Nuclear Information System (INIS)

    This study aims to analyze and explain the structure of a reactive jet of water steam in liquid sodium, as from a ligh pressure tank and an orifice of very small section. The prior understanding of this reactive jet makes it possible to explain certain results of erosion-corrosion (Wastage) that can occur in the steam generators of breader reactor power stations. This study gave rise to an experimental simulation (plane jet of water steam on a bed of sodium), as well as to suggesting a reactive jet model according to the principle of an ''immersed Na-H2O diffusion flame''

  5. Clean energy for a new generation. Steam generator life cycle management and Bruce restart

    International Nuclear Information System (INIS)

    In the mid to late 1990s, Ontario Hydro decided to lay-up and write-down the Bruce A Nuclear Reactors. Upon transition to Bruce Power L.P., Canada's first and only private nuclear operator, new life and prospects were injected into the site, local economy and the provincial energy portfolio. The first step in this provincial power recovery initiative involved restart of Bruce Units 3 and 4 in the 2003/04 time-frame. Units 3 and 4 have performed beyond expectation during the last five-year operating interval. A combination of steam generator and fuel channel issues precluded a similar restart of Units 1 and 2. Enter the refurbishment of Bruce Units 1 and 2. This first-of-a-kind undertaking within the Canadian nuclear power industry is testament to the demonstrated industry leadership by Bruce Power L.P., their investors and the significant vendor community contribution that is supporting this major power infrastructure enhancement. Initiated as a 'turn-key' project solution separated from the operating units, this major refurbishment project has evolved to a fully managed in-house refurbishment project with the continued support from the broader vendor community. As part of this first-of-kind undertaking, Bruce Power L.P. is in the process of accomplishing such initiatives as a complete fuel channel re-tube (i.e. full core calandria and pressure tube replacement), replacement of all boilers (i.e. 16 in total) and the majority of feeder pipe replacement. Complimentary major upgrades and replacement of the remainder of plant equipment including both nuclear and non-nuclear valves, heat exchangers, electrical infrastructure, service water systems and components, all while meeting a parallel evolving/maturing regulatory environment related to achieving compliance with IAEA derived modern codes and standards. Returning to ground level, boiler replacement is a key part of the refurbishment undertaking and this further reflected a meeting of the 'old' and the 'new'. Pre

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

    Institute of Scientific and Technical Information of China (English)

    李文红; 施天裕

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-12-15

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

  8. Seismic design, analysis and testing of the HTGR MK-IVA steam generator

    International Nuclear Information System (INIS)

    The HTGR MK-IVA Steam Generator is a once-through design consisting of two heat transfer bundles, a helical economizer-evaporator-superheater (EES), and a straight tube finishing superheater (STSH) which is located at the longitudinal center of the helical EES. The helically coiled EES bundle support structural design, dynamic analysis, and seismic test program are described

  9. Review on Dutch-German steam generator safety activities in the field of SNR-development

    International Nuclear Information System (INIS)

    This paper reviews the results of the experimental and theoretical safety related work for SNR steam generators in brief. A review of the operating experience with SNR-300 sodium-water heat-exchangers and protection philosophy used in SNR-2 is also given

  10. Steam generators and heat exchangers for gas-cooled reactors. Background and status in Switzerland

    International Nuclear Information System (INIS)

    The Swiss company Sulzer Brothers Ltd. built its first nuclear steam generator in 1961 for a CO2-cooled prototype reactor. Since then the Company has been involved in the planning, development and manufacture of steam generators for gas-cooled reactors, in particular for the French Magnox reactor program. In 1980 Sulzer delivered the 6-module steam generator for the German High Temperature Reactor Prototype THTR-300. The production of hardware was continuously accompanied and supported by extensive research and development activities. Experimental programs comprised thermohydraulic investigations related to the primary gas-side as well as to the secondary side and its two-phase-flow stability. In the area of high temperature materials thermal cycling tests were performed to analyse the fatigue of bimetallic welds under severe transients. Low cycle creep fatigue damage in tube bends and the wear and fretting characteristics of protective coatings on the helium side of hot tubes were investigated. Fabrication experiments for large helical heat exchangers served to extrapolate known manufacturing technology to commercial size HTGR units. In the frame of international GCR programs Switzerland participated in the Gas Breeder Reactor Association and the High Temperature Helium Turbine Project. For these projects Sulzer designed and developed steam generators, recuperators and primary coolers

  11. Improving electron beam weldability of heavy steel plates for PWR-steam generator

    International Nuclear Information System (INIS)

    Installation and replacement of many PWR-steam generators are planned inside and outside Japan. The steel plates for steam generators are heavy in thickness, and increase the number of welding passes and prolong the welding time. Electron beam welding (EBW) can greatly reduce the welding period compared with conventional welding methods (narrow-gap gas metal arc welding (GMAW) and submerged arc welding (SAW)). The problems in applying EBW are to prevent weld defects and to improve the toughness of the weld metal. Defect-free welding procedures were successfully established even in thick steel plates. The factors that deteriorate weld-metal (WM) toughness of EBW were investigated. The manufacturing process, which utilizes a new secondary refining process at steelmaking and a high-torque mill at plate mill in actual mass-production, were established. EBW base metal and WM have better properties including fracture toughness than those of conventional welding processes. As a result, an application of EBW to the fabrication of PWR-steam generators has become possible. Large amounts of ASTM A533 Gr B Cl 2 (JIS SQV2B) steel plates in actual PWR-steam generators have come to be produced (more than 1,500 ton) by applying EBW. (author)

  12. Justification procedure of hydrogen safety of RU BN compartments as exemplified by steam generator box

    International Nuclear Information System (INIS)

    Basic proposals on completing calculational procedures of sodium cooled reactors hydrogen safety justification, particularly steam generator (SG) box, are presented. It is pointed out that hydrogen appearance in SG box is possible only in accidents with multiple failures. The data presented shows that hydrogen safety and explosion-proofness of fast reactor facility SG box are provided without any special complex and expensive means

  13. Analysis of cold crack of AP1000 steam generator tube sheet cladding

    International Nuclear Information System (INIS)

    This paper discusses the causes of AP1000 Steam Generator (SG) tube sheet underclad cracking. Base metal weldability, hydrogen influence, welding techniques and weld residue stresses are all discussed in details in contributing to the underclad cracking problems. Feasible and realistic improvement plans are proposed for the AP1000 SG tube sheet cladding, including the controlling forging procurement, welding process and cladding techniques. (authors)

  14. Opinion on serviceability of Bugey 3 reactor steam generators until their replacement foreseen in September 2010

    International Nuclear Information System (INIS)

    This document briefly reports the damage characterization of tubular bundles in steam generators of the Bugey 3 reactor, discusses the actions which are foreseen to prevent a tube failure risk, and discusses the risk of leakage during operation. Recommendations are formulated about investigation on the corrosion, and about prediction computation to be performed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

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

  16. EVALUATION OF STATIONARY SOURCE PARTICULATE MEASUREMENT METHODS. VOLUME II. OIL-FIRED STEAM GENERATORS

    Science.gov (United States)

    An experimental study was conducted to determine the reliability of the Method 5 procedure for providing particulate emission data from an oil-fired steam generator. The study was concerned with determining whether any 'false' particulate resulted from the collection process of f...

  17. Monte Carlo simulation of single and two-dosimeter approaches in a steam generator channel head.

    Science.gov (United States)

    Kim, C H; Reece, W D

    2002-08-01

    In a steam generator channel head, it was not unusual to see radiation workers wearing as many as twelve dosimeters over the surface of the body to avoid a possible underestimation of effective dose equivalent (H(E)) or effective dose (E). This study shows that only one or two dosimeters can be used to estimate H(E) and E without a significant underestimation. MCNP and a point-kernel approach were used to model various exposure situations in a steam generator channel head. The single-dosimeter approach (on the chest) was found to underestimate H(E) and E significantly for a few exposure situations, i.e., when the major portion of radiation source is located in the backside of a radiation worker. In this case, the photons from the source pass through the body and are attenuated before reaching the dosimeter on the chest. To assure that a single dosimeter provides a good estimate of worker dose, these few exposure situations cannot dominate a worker's exposure. On the other hand, the two-dosimeter approach (on the chest and back) predicts H(E) and E very well, hardly ever underestimating these quantities by more than 4% considering all worker positions and contamination situations in a steam generator channel head. This study shows that two dosimeters are adequate for an accurate estimation of H(E) and E in a steam generator channel head. PMID:12132712

  18. Devices for the contamination containment employees in the steam generator inspection

    International Nuclear Information System (INIS)

    The process of induced current inspection of the tubes of the steam generator is a typical programmed inspections at each refueling outages of pressurized water in nuclear power plants. components inspection being quite active, interested in the program of continuous improvement, further optimize the inspection system.

  19. Evaluation of hideout return data from U.S. PWR steam generators

    International Nuclear Information System (INIS)

    Since the middle to late 1970's, dramatic reductions in the quantities of impurities in the bulkwater of PWR steam generators have been made by U.S. utilities. Today most utilities operate at full power with impurity concentrations in the steam generator blowdown in the low ppb range, well within existing industry guideline control limits. Despite these efforts, some of these same utilities have subsequently encountered secondary side stress corrosion cracking (SCC) and intergranular attack (IGA) of steam generator tubing within deep tubesheet crevices and more recently at tube support intersections. It must, therefore, be concluded that either continuous low level input of contaminants within existing guideline limits, or intermittent short duration input, undetected by either current sampling and analysis techniques or procedures, are permitting ingress of corrosive impurity species which subsequently concentrate in flow-occluded regions to produce localized tube corrosion. To better understand both the quantity and composition of accumulated impurity species, more and more utilities, even those who have not experienced any steam generator corrosion, have begun to perform rigorous sampling and analysis evaluations of returning chemical contaminants each time the units are brought off-line. This paper will show examples of how these data are being used by U.S. industry to gain valuable information about accumulated contaminant inventories, to make cycle-to-cycle and plant-to-plant comparisons, and to develop plant specific actions to promote maximum contaminant removal. (author)

  20. Experience from in-service inspections and repairs of steam generators and pressurizers

    International Nuclear Information System (INIS)

    Experience is described concerning repairs of sealing nodes of the primary and secondary circuit of steam generators, defects of thread nests, and replacement of protective jackets of primary collector welds. Problems are discussed of in-service inspections and repairs of the austenitic overlay of the pressurizer and of testing the heat transfer tubes by the eddy current method. (E.J.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-15

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

  2. Extraction: a system for automatic eddy current diagnosis of steam generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Improving speed and quality of Eddy Current non-destructive testing of steam generator tubes leads to automatize all processes that contribute to diagnosis. This paper describes how we use signal processing, pattern recognition and artificial intelligence to build a software package that is able to automatically provide an efficient diagnosis. (authors). 2 figs., 5 refs

  3. SID seeks sludge and foreign bodies in steam generator upper heads while CECIL robotically removes them

    International Nuclear Information System (INIS)

    The CECIL [Consolidated Edison Combined Inspection and Lancing] system provides a range of maintenance tools for cleaning the region between the tubesheet and the first support plate in steam generators. SID [Secondary Inspection Device] extends access to and inspection of the upper bundle region. Work is being done to combine CECIL and SID technology to inspect and hydraulically clean the upper bundle. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Strohmer, F.

    2013-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Susyadi

    2007-07-01

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

  6. Stress corrosion cracking susceptibility of steam generator tubing on secondary side in restricted flow areas

    International Nuclear Information System (INIS)

    Nuclear steam generator tubes operate in high temperature water and on the secondary side in restricted flow areas many nonvolatile impurities accidentally introduced into circuit tend to concentrate. The concentration process leads to the formation of highly aggressive alkaline or acid solutions in crevices, and these solutions can cause stress corrosion cracking (SCC) on stressed tube materials. Even though alloy 800 has shown to be highly resistant to general corrosion in high temperature water, it has been found that the steam generator tubes may crack during service from the primary and/or secondary side. Stress corrosion cracking is still a serious problem occurring on outside tubes in operating steam generators. The purpose of this study was to evaluate the environmental factors affecting the stress corrosion cracking of steam generators tubing. The main test method was the exposure for 1000 hours into static autoclaves of plastically stressed C-rings of Incoloy 800 in caustic solutions (10% NaOH) and acidic chloride solutions because such environments may sometimes form accidentally in crevices on secondary side of tubes. Because the kinetics of corrosion of metals is indicated by anodic polarization curves, in this study, some stressed specimens were anodically polarized in caustic solutions in electrochemical cell, and other in chloride acidic solutions. The results presented as micrographs, potentiokinetic curves, and electrochemical parameters have been compared to establish the SCC behavior of Incoloy 800 in such concentrated environments. (authors)

  7. Draft environmental statement related to steam generator repair at H.B. Robinson Steam Electric Plant Unit No. 2, (Docket No. 50-261)

    International Nuclear Information System (INIS)

    The staff has considered the environmental impacts and economic costs of the proposed steam generator repair at the H.B. Robinson Steam Electric Plant Unit No. 2 along with reasonable alternatives to the proposed action. The staff has concluded that the proposed repair will not significantly affect the quality of the human environment and that there are no preferable alternatives to the proposed action. Furthermore, any impacts from the repair program are outweighted by its benefits

  8. Evolution of the CANDU ICS-90+ control room design

    International Nuclear Information System (INIS)

    The design of the CANDU Control Room and the associated design process has evolved considerably over several generations of plants, from the first commercial scale demonstration CANDU at Douglas Point through to the large scale CANDUs at Darlington, and beyond, for the next generation of CANDU plant, ICS-90+, represented by new designs like CANDU 3. In the early plants, the control room configuration was based on designers' projections of control interface requirements. With succeeding generations, of designs, there has been an evolution towards: increasing attention to formal requirements definition, incorporation into the Human Machine Interface (HMI) of a larger base of operational experience, more systematic consideration of Human Factors (HF) aspects of the design and the application of a more powerful computer based HMI. For the newest plant, the CANDU 3, a Human Factors Engineering Program Plan (HFEPP) defines the overall HF engineering process, the associated requirements and HF engineering standards to be followed in each stage, and for all HMI aspects of the control room and plant design. The CANDU 3 control room also incorporates several new design innovations that will facilitate operating crew performance improvements. These are based on past experience with operating CANDU plants, incorporated with the use of formal design and validation methods plus results from Canadian research program to support control centre design and operation. For example, there are design improvements to facilitate: operator tracking of plant state, problem solving, alarm filtering, annunciation system interrogation, special safety system testing features, etc. The present paper will expand and elaborate on each of the above topics. (author). 7 refs, 2 figs, 1 tab

  9. Mathematical model of processes in cut-off leaking steam generator as a part of operating ship reactor

    International Nuclear Information System (INIS)

    A mathematical model of processes in cut-off leaking steam generator of an operating ship reactor was presented. The model describes the coolant pressure increasing when filling the heat pipes system of cut-off leaking direct-flow steam generator, and allows estimating the time lag before putting restrictions on power maneuvering of the ship reactor

  10. Predicted wear on the tube outside surface due to foreign object in the secondary side of steam generator

    International Nuclear Information System (INIS)

    It is necessary to evaluate the effects of foreign objects on steam generator tubes and to use this information to take appropriate safety precautions to prevent nuclear accidents. Foreign objects may include loose parts from the feed water system and items lost by workers during o/h, and may flow into the secondary side of steam generators during operation. A foreign object could damage steam generator tube walls if there is relative motion between the tube and the foreign object. This is especially true for foreign objects that land on the tube sheet because the velocity of cross flow, which creates a contact force between the tube and foreign object, is relatively high there. During steam generator overhauls, foreign objects are detected by non destructive methods such as the visual test and/or the eddy current test. Confirmed foreign objects should be removed for nuclear safety. The Foreign Object Search and Retrieval System (FOSAR) can be used to remove foreign objects from the steam generators with a square tube array. However, the FOSAR cannot be used (or can be used in only a very restricted area, such as the outside of the tube bundle) in the steam generators with a triangular tube array. In order to continue nuclear power plant operations without removing foreign objects, the integrity of the steam generator tube must be verified. This paper introduces a practical method developed to evaluate the effects of foreign objects detected on tube sheets in the secondary sides of steam generators

  11. CFD simulation of particle entrapment of steam generator sludge collector/loose parts weir

    International Nuclear Information System (INIS)

    A computational fluid dynamics (CFD) study was performed to explore the performance and interaction of two components designed for Pressurized Water Reactor (PWR) recirculating steam generators: the sludge collector and the loose parts weir. The sludge collector is a passive device located in the upper internals region of a PWR steam generator. The sludge collector's function is to trap sludge, particulates suspended in the secondary side recirculating flow and minimize its deposition on the tube bundle, where it increases susceptibility to tube degradation. The loose parts weir is a separate passive device placed around the sludge collector, which is expected to act as a barrier to prevent any loose parts in the upper internals region from reaching the tube bundle. Loose parts in the steam generator, if not captured, can reach the tube bundle and cause tube wall damage. The loose parts weir is considered for installation in combination with the sludge collector for both new and existing steam generators. Previously, the configuration of the sludge collector was determined based on testing and analysis. Inclusion of the loose parts weir significantly alters the flow field of the sludge collector and thus its overall performance. The purpose of this investigation is to verify the performance of the sludge collector and loose parts weir. A CFD study was performed to evaluate the interaction between the sludge collector and a loose parts weir within the steam generator, which provides insight into the flow redistribution. Models were developed for the sludge collector with and without the loose parts weir using commercial CFD software. Results including the velocity and pressure profile in the steam generator upper internals, as well as fluid mass flow passing through the sludge collector, are presented and discussed for various configurations of the sludge collector and loose parts weir. Sludge particle collection rates are estimated using an empirical correlation

  12. Examination of failed studs from No. 2 steam generator at the Maine Yankee Nuclear Power Station

    International Nuclear Information System (INIS)

    Three studs removed from service on the primary manway cover from steam generator No. 2 of the Maine Yankee station were sent to Brookhaven National Laboratory (BNL) for examination. The examination consisted of visual/dye penetrant examination, optical metallography and Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS) evaluation. One bolt was through cracked and its fracture face was generally transgranular in nature with numerous secondary intergranular cracks. The report concludes that the environmenally assisted cracking of the stud was due to the interaction of the various lubricants used with steam leaks associated with this manway cover

  13. Development of helium leak testing system and procedure for testing welds of steam generator

    International Nuclear Information System (INIS)

    FBR steam generator (SG) is a vertical shell and tube type heat exchanger with sodium on shell side and water/steam on tube side. As sodium and water are highly reactive, the boundary separating these two fluids, is designed for highest integrity. All welds in these SGs are hence, very critical and require a very high degree of reliability and integrity, especially the thin joints viz. tube to tube sheet joints. Failure of these joints will cause cascading effect. Final integrity of weld joints is required to be ascertained by performing helium leak testing. The paper describes the methodology developed and implementation for leak testing

  14. Impact of lattice geometry distortion due to ageing on selected physics parameters of a CANDU reactor

    International Nuclear Information System (INIS)

    In this paper, results related to a limited scope assessment of the geometry-distortion-induced effects on key reactor physics parameters of a CANDU reactor are discussed. These results were generated by simulations using refined analytical methods and detailed modeling of CANDU reactor core with aged lattice cell geometry. (authors)

  15. Equipment for monitoring steam or water leakage into coolant, especially steam generator sodium

    International Nuclear Information System (INIS)

    A system of exciting coils is wound on the external tube through which the coolant flows. The coils are excited with a power generator. This alternating field reaches into the liquid metal. Two measuring coils are also wound on the external surface of the said tube. The measuring coils are connected across one another and the resulting voltage is applied to the input of a differential amplifier with indicator. In a non-accident state the indicator shows the flow rate of the liquid metal. In case of a leakage the chemical reaction products will change the flow rate of sodium, which is recorded by the indicator connected both to the measuring and accident circuit. (E.F.)

  16. Analysis and potential of once-through steam generators in line focus systems - Final results of the DUKE project

    Science.gov (United States)

    Feldhoff, Jan Fabian; Hirsch, Tobias; Pitz-Paal, Robert; Valenzuela, Loreto

    2016-05-01

    The direct steam generation in line focus systems such as parabolic troughs and linear Fresnel collectors is one option for providing `solar steam' or heat. Commercial power plants use the recirculation concept, in which the steam generation is separated from the superheating by a steam drum. This paper analyzes the once-through mode as an advanced solar field concept. It summarizes the results of the DUKE project on loop design, a new temperature control strategy, thermo-mechanical stress analysis, and an overall cost analysis. Experimental results of the temperature control concept at the DISS test facility at Plataforma Solar de Almería are presented.

  17. Analysis method of deposit on steam generator tubes using eddy current

    International Nuclear Information System (INIS)

    The steam generator tubes in operating nuclear power plants have an important problem during their operation time, an accumulation of corrosion products. Such corrosion products form from the secondary side of the plant system, such as carbon steel pipelines, heat exchange shell, and turbine. The accumulation position of corrosion product is mainly on the top of the tubesheet, the tube support structures of the steam generator. It is extinguished as sludge, and as a deposit of the corrosion product. The volume increase and hardening of the sludge eventually cause the tube deformation, and a blockage of the coolant flow of the tube supports. A deposit outside the tube results in reducing the heat transfer through the tube wall, and distorts the eddy current signals during an in service inspection. For the management of steam generator tubes from the problems mentioned previously, several methods were performed. The corrosion products could be reduced by chemical cleaning and sludge lancing. The monitoring of the quantity of sludge and deposit is very important data in the management of steam generator tubes. An eddy current testing (ECT) method is very useful to detect flaws and defects in the steam generator (SG) tubes of nuclear power plants (NPPs) during an in service inspection. Recently, it was reported that deposit loading can be measured using eddy current test data, especially of a bobbin probe. For a precision measurement using a non destructive method, a calibration technique is required using the simulated deposit and signal characteristics from the deposit standard. In this study, a personal computer based device was developed for an analysis of ECT signals. The soft wave program can convert the commercial eddy current data and measure the deposit amount from the calibration data

  18. Current forgings and their properties for steam generator of nuclear plant

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Hisashi; Suzuki, Komei; Kusuhashi, Mikio; Sato, Ikuo [Japan Steel Works Ltd., Muroran (Japan)

    1997-12-31

    Current steel forgings for steam generator (SG) of PWR plant are reviewed in the aspect of design and material improvement. The following three items are introduced. The use of integral type steel forgings for the fabrication of steam generator enhances the structural integrity and makes easier fabrication and inspection including in-service inspection. The following examples of current integral type forgings developed by the Japan Steel Works, Ltd. (JSW) are introduced: (1) primary head integrated with nozzles, manways and supports; (2) steam drum head integrated with nozzle and handhole; (3) conical shell integrated with cylindrical sections and handholes. In order to decrease the weight of steam generator, the high strength materials such as SA508, Cl.3a steel have been adopted in some cases. The properties of this steel are introduced and the chemistry and heat treatment condition are discussed. As one of the methods to minimize the macro- and micro-segregations, the use of vacuum carbon deoxidation (VCD), i.e. deoxidization of steel by gaseous CO reaction, with addition of Al for grain refining was investigated. The properties of SA508, Cl.3 steels with Low Si content are compared with those of conventional one.

  19. The evolution of the CANDU energy system - ready for Europe's energy future

    International Nuclear Information System (INIS)

    As air quality and climate change issues receive increasing attention, the opportunity for nuclear to play a larger role in the coming decades also increases. The good performance of the current fleet of nuclear plants is crucial evidence of nuclear's potential. The excellent record of Cernavoda-1 is an important part of this, and demonstrates the maturity of the Romanian program and of the CANDU design approach. However, the emerging energy market also presents a stringent economic challenge. Current NPP designs, while established as reliable electricity producers, are seen as limited by high capital costs. In some cases, the response to the economic challenge is to consider radical changes to new design concepts, with attendant development risks from lack of provenness. Because of the flexibility of the CANDU system, it is possible to significantly extend the mid-size CANDU design, creating a Next Generation product, without sacrificing the extensive design, delivery and operations information base for CANDU. This enables a design with superior safety characteristics while at the same time meeting the economic challenge of emerging markets. The Romanian nuclear program has progressed successfully forward, leading to the successful operation of Cernavoda-1, and the project to bring Cernavoda-2 to commercial operation. The Romanian nuclear industry has become a full-fledged member of the CANDU community, with all areas of nuclear technology well established and benefiting from international cooperation with other CANDU organizations. AECL is an active partner with Romanian nuclear organizations, both through cooperative development programs, commercial contracts, and also through the activities of the CANDU owners' Group (COG). The Cernavoda project is part of the CANDU 6 family of nuclear power plants developed by AECL. The modular fuel channel reactor concept can be modified extensively, through a series of incremental changes, to improve economics, safety

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

    Institute of Scientific and Technical Information of China (English)

    Shujuan LI; Yuzhen WANG

    2006-01-01

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

  1. High-speed Steam Turbine Systems for Distributed Generation Applications

    OpenAIRE

    Petrov, Miroslav; Fridh, Jens; Göransson, Åke; Fransson, Torsten

    2012-01-01

    The efficiency of utilization of low-grade solid fuels of either renewable or fossil origin such as biomass, municipal or agricultural wastes, peat, lignite, etc. for distributed generation applications and combined heat and power (CHP) production at small scales can be improved by a simple technology shift. This study evaluates the technical feasibility of a compact power generation package comprising a small steam turbine directly coupled to a high-speed alternator delivering around 2 MW of...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  3. Independence and diversity in CANDU shutdown systems

    International Nuclear Information System (INIS)

    Atomic Energy Control Board regulations state that Canadian CANDU reactors shall have two fully effective, independent and diverse shutdown systems. The Darlington Nuclear Generating Station is the first power plant operated by Ontario Hydro to make use of software-based computer control in its shutdown systems. By virtue of the reliance placed on these systems to prevent exposure of the public to harmful radioactivity in the event of an accident, the shutdown system software has been categorized as safety critical software. An important issue that was considered in the design of the Darlington shutdown systems was how the software should be designed and incorporated into the systems to comply with the independence and diversity requirement. This paper describes how the independence and diversity requirement was complied with in previous CANDU shutdown system designs utilizing hardware components. The difference between systems utilizing hardware alone, and those utilizing both hardware and software are discussed. The results of a literature search into the issue of software diversity, the behaviour of multi-version software systems, and experience in other industries utilizing safety critical software are referred to. This paper advocates a systems approach to designing independent shutdown systems utilizing software. Opportunities exist at the system level for design decisions that can enhance software diversity and can reduce the likelihood of common mode faults in the systems. In the light of recent experience in implementing diverse safety critical software, potential improvements to the design process for CANDU shutdown systems are identified. (Author) 19 refs

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

    Directory of Open Access Journals (Sweden)

    Gustavo Boroni

    2011-01-01

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

  5. Steam turbine generators -from NC manufacturing to CAD/CAM

    Energy Technology Data Exchange (ETDEWEB)

    Searle, D.R.; King, F.E.; Kiniskern, J.M.

    1985-04-01

    A program has been designed to integrate engineering, manufacturing, and business systems using a common data base. There has been a significant increase in benefits obtained by extending the automation of the design/drafting function to include manufacturing operations. This extension would have been difficult without the existence of highly developed family-of-parts NC programs. The integration concept has also been applied to turbine buckets and is being extended to other turning-generator components.

  6. The AREVA customized chemical cleaning C3-concept as part of the steam generator asset management

    International Nuclear Information System (INIS)

    In pressurized water reactors corrosion products and impurities are transported into the steam generators by feed water. Corrosion products and impurities are accumulated in the SGs as deposits and scales on the tubes, the tube support structures and the tube sheet. Depending on the location, the composition and the morphology such deposits may negatively affect the performance of the steam generators by reducing the thermal performance, changing the flow patterns and producing localized corrosion promoting conditions. Accordingly removal of deposits or deposit minimization strategies are an essential part of the asset management program of the steam generators in Nuclear Power Plants. It is evident that such a program is plant specific, depending on the individual condition prevailing. Parameters to be considered are for example: - Steam generator and balance of plant design; - Secondary side water chemistry treatment; - Deposit amount and constitution; - Deposit distribution in the steam generator; - Existing or expected corrosion problems. After evaluation of the steam generator condition a strategy for deposit minimization has to be developed. Depending on the individual situation such strategies may span from curative full scale cleanings which are capable of removing the entire sludge inventory in the range of several 1000 kg per SG to preventive cleanings that remove only a portion of the deposits in the range of several 100 kg per SG. But also other goals depending on the specific plant situation, like tube sheet sludge piles or hard scale removal, may be considered. Beside the chemical cleaning process itself also the integration of the process into the outage schedule and considerations about its impact on other maintenance activities is of great importance. It is obvious that all these requirements cannot be met easily by a standardized cleaning method, thus a customisable chemical cleaning technology is required. Based on its comprehensive experience

  7. Research program of natural circulation steam generator design of national 1000 MWe PWR

    International Nuclear Information System (INIS)

    A concept design of natural circulation steam generator for the national 1000MWe PWR of the Chinese National Nuclear Program has been proposed and a relevant research program to validate the efficiency and/or effectiveness of some new design assemblies and/or components has been completed. There are three salient features in the steam generator design. Firstly, steam separation equipment was improved and carryover moisture was further reduced to below 0.1%. Secondly, the water level at the secondary side being elevated, secondary side water volume was expanded to satisfy the EPRI-URD requirements of LFB20. Finally, an inactive device, sludge collector was incorporated to enhance the secondary sludge control. The validation research program consists of two parts; cold state screening test, hot state validation test and corresponding computational analysis, and cold state test and corresponding computational analysis of sludge collector design. The validation tests were completed in 2001. Three sets of steam separator were selected from cold screening tests for hot validation. The hot test showed the most important parameters, outlet carryover, of all three sets were under 0.1%. The best case was only 0.0018%. The sludge collector showed a collection efficiency of over 50%

  8. The Development of a Small High Speed Steam Microturbine Generator System

    Science.gov (United States)

    Alford, Adrian; Nichol, Philip; Frisby, Ben

    2015-08-01

    The efficient use of energy is paramount in every kind of business today. Steam is a widely used energy source. In many situations steam is generated at high pressures and then reduced in pressure through control valves before reaching point of use. An opportunity was identified to convert some of the energy at the point of pressure reduction into electricity. This can be accomplished using steam turbines driving alternators on large scale systems. To take advantage of a market identified for small scale systems, a microturbine generator was designed based on a small high speed turbo machine. This gave rise to a number of challenges which are described with the solutions adopted. The challenges included aerodynamic design of high efficiency impellers, sealing of a high speed shaft, thrust control and material selection to avoid steam erosion. The machine was packaged with a sophisticated control system to allow connection to the electricity grid. Some of the challenges in packaging the machine are also described. The Spirax Sarco TurboPower has now concluded performance and initial endurance tests which are described with a summary of the results.

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

    Science.gov (United States)

    Bovsunovsky, A. P.

    2015-07-01

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

  10. Design and performance verification of fuel assembly and steam generator simulators for SMART reactor

    International Nuclear Information System (INIS)

    The SMART reactor has been developed at KAERI, for the generation of electric power and also for seawater desalination. In order to verify the performance of the SMART design with respect to flow and pressure distribution, an experimental test facility named SCOP has been developed. For the purpose of preserving the flow distribution characteristics, SCOP is linearly reduced with a scaling ratio of 1/5. A CFD analysis was carried out to draw basic design parameters of the venturi tube and the perforated plates in a fuel assembly simulator. A CALIP, which is a flow and pressure drop calibration test facility, has been constructed to evaluate the pressure drop characteristic of fuel assembly and steam generator simulators. This paper shows the results of the actual performance verification and evaluation of fuel assembly and steam generator simulator, were evaluated using a CALIP. (author)

  11. Experience with vacuum distillation cleaning of a full-size steam generator

    International Nuclear Information System (INIS)

    In the 50 MW Sodium Component Test Facility at Hengelo tests are conducted on several types of full size prototype steam generators and an intermediate heat exchanger. The necessary post-test examination of these prototype components requires a complete removal of all sodium. Since in some cases the endurance test has to be continued after internal inspection, the cleaning-method should be such that no damage occurs to the component. After partial disassembly and internal inspection the component will be reassembled and must be acceptable for further use. The qualification tests of the Neratoom straight tube steam generator were concluded in June 1974. The evaporator module was decided to be partially disassembled in order to meet the requirement of a thorough examination before fabrication of the SNR-generators is started. In preparation for the most suitable cleaning procedure, several methods of sodium removal were considered

  12. Design and performance of BWC replacement steam generators for PWR systems

    International Nuclear Information System (INIS)

    In recent years, Babcock and Wilcox Canada (BWC) has provided a number of PWR Replacement Steam Generators (RSGS) to replace units that had experienced extensive Alloy 600 tube degradation. BWC RSG units are in operation at Northeast Utilities' Millstone Unit 2, Rochester Gas and Electric's Ginna Station, Duke Energy's Catawba Unit 1, McGuire Unit 1 and 2, Florida Power and Light's St. Lucie Unit 1 and Commonwealth Edison's Byron 1 Station. Extensive start-up performance characteristics have been obtained for Millstone 2, Ginna, McGuire 1, and Catawba 1 RSGS. The Millstone 2, Ginna and Catawba 1 RSGs have also undergone extensive inspections following their first cycle of operation. The design and start-up performance characteristics of these RSGs are presented. The BWC Replacement Steam generators were designed to fit the existing envelope of pressure boundary dimensions to ensure licensability and integration into the Nuclear Steam Supply System. The RSGs were provided with a tube bundle of Alloy 690TT tubing, sized to match or exceed the original steam generator (OSG) thermal performance including provision for the reduced thermal conductivity of Alloy 690 relative to Alloy 600. The RSG tube bundle configurations provide a higher circulation design relative to the OSG, and feature corrosion resistant lattice grid and U-bend tube supports which provide effective anti-vibration support. The tube bundle supports accommodate relatively unobstructed flow and allow unrestrained structural interactions during thermal transients. Efficient steam separators assure low moisture carryover as well as high circulation. Performance measurements obtained during start-up verify that the BWC RSGs meet or exceed the specified thermal and moisture carryover performance requirements. RSG water level stability results at nor-mal operation and during plant transients have been excellent. Visual and ECT inspections have confirmed minimal deposition and 100% tube integrity following

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

    International Nuclear Information System (INIS)

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

  14. Wear on Plugged Tube due to the Foreign Objects on the Secondary Side of Steam Generator

    International Nuclear Information System (INIS)

    In this paper, the changes of the tube frequency and amplitude are introduced before and after plugging. The amplitude of the bottom span for the steam generator tube is not much changed after tube plugging. Moreover, the contact force between the plugged tube and the foreign object is the same as that of intact tube and the foreign object. However, the frequencies of plugged tubes are about 9∼12% higher than those of intact tubes. That means the wear due to the foreign object would be accelerated after the tube plugging. Therefore, the tube stabilizer should be installed when the tube is plugged due to the foreign object wear. The tube wall of steam generator is a pressure boundary between the coolant of the primary system and the feedwater of the secondary system. It is very important to insure the structural integrity of the tubes because the radioactive coolant is flow into the feedwater due to the pressure difference as the result of tube failure. The degradations of steam generator tubes are corrosion, wear, fatigue and foreign object wear, etc. The foreign object wear is one of mechanical degradation due to materials flew into the secondary side of steam generator. The steam generator tubes, estimated not to insure structural integrity from the results of the nondestructive evaluation such as eddy current test and visual inspection, are excluded from the service with plugging. However, the tube wear is still being progressed after the plugging because the relative motion between the tube and structure is still existed due to the secondary side flow in the steam generator. If the tube is completely cut because of the degradation, the tube can be a stress or of failure of tubes around the plugged tube. The contact force between the structure and tube is lowered as the wear is progressed. However, the contact force between the foreign object and tube is not changed as the wear is progressed. Therefore, the structural integrity of tubes around the foreign

  15. Evolution of the CANDU control centre design process

    International Nuclear Information System (INIS)

    The design of the CANDU NPP control centre and the associated control centre design process has evolved considerably over several generations of plants, from Douglas Point through Darlington, and beyond, to new designs like CANDU 3. In the early plants, the control centre configuration had to be based on designers' projections of control interface requirements. With succeeding generations of designs, along with the introduction of advancing computer control technology, a larger based of operational experience has been factored into the control interface design, and increasing attention has been given to more formal requirements definition, and more systematic consideration of human factors aspects of the design

  16. Technology transfer in CANDU marketing

    International Nuclear Information System (INIS)

    The author discusses how the CANDU system lends itself to technology transfer, the scope of CANDU technology transfer, and the benefits and problems associated with technology transfer. The establishment of joint ventures between supplier and client nations offers benefits to both parties. Canada can offer varying technology transfer packages, each tailored to a client nation's needs and capabilities. Such a package could include all the hardware and software necessary to develop a self-sufficient nuclear infrastructure in the client nation

  17. The CANDU experience in Romania

    International Nuclear Information System (INIS)

    The CANDU program in Romania is now well established. The Cernavoda Nuclear Station presently under construction will consist of 5-CANDU 600 MWE Units and another similar size station is planned to be in operation in the next decade. Progress on the multi-unit station at Cernavoda was stalled for 18 months in 1982/83 as the Canadian Export Development Corporation had suspended their loan disbursements while the Romanian National debt was being rescheduled. Since resumption of the financing in August 1983 contracts worth almost 200M dollars have been placed with Canadian Companies for the supply of major equipment for the first two units. The Canadian design is that which was used in the latest 600 MWE CANDU station at Wolsong, Korea. The vast construction site is now well developed with the cooling water systems/channels and service buildings at an advanced stage of completion. The perimeter walls of the first two reactor buildings are already complete and slip-forming for the 3rd Unit is imminent. Many Romanian organizations are involved in the infrastructure which has been established to handle the design, manufacture, construction and operation of the CANDU stations. The Romanian manufacturing industry has made extensive preparations for the supply of CANDU equipment and components, and although a major portion of the first two units will come from Canada their intentions are to become largely self-supporting for the ensuing CANDU program. Quality assurance programs have been prepared already for many of the facilities

  18. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10-6/year. CANDU nuclear plant designers and owner/operators share information and operational experience nationally and internationally through the CANDU Owners' Group (COG). The research program generally emphasizes the unique aspects of the CANDU concept, such as heat removal through the moderator, but it has also contributed significantly to areas generic to most power reactors such as hydrogen combustion, containment failure modes, fission product chemistry, and high temperature fuel behaviour. Abnormal plant operating procedures are aimed at first using event-specific emergency operating procedures, in cases where the event can be diagnosed. If this is not possible, generic procedures are followed to control Critical Safety Parameters and manage the accident. Similarly, the on-site contingency plans include a generic plan covering overall plant response strategy, and a specific plan covering each category of contingency

  19. ENVIRONMENTAL ASSESSMENT OF AN ENHANCED OIL RECOVERY STEAM GENERATOR EQUIPPED WITH A LOW-NOX BURNER. VOLUME 2. DATA SUPPLEMENT

    Science.gov (United States)

    The report is a compendium of detailed test sampling and analysis data obtained in field tests of an enhanced oil recovery steam generator (EOR steamer) equipped with a MHI PM low-NOx crude oil burner. Test data included in the report include equipment calibration records, steame...

  20. Application of concept mapping principles to managing steam generator knowledge at the CNSC

    International Nuclear Information System (INIS)

    The Canadian Nuclear Safety Commission (CNSC) oversee proper regulation required for the operations of more than 180 commercial steam generators housed in Canadian nuclear power plants. The number of inspection reports and other technical reports concerning the operation of these units result in a significant amount of information over time. Consequently the method to access the information easily and in a timely manner can require improvement. The Steam Generator Knowledge Management project was developed as a mode to efficiently manage and integrate all knowledge and resources relevant to steam generators found in a variety of sources used by the CNSC to conduct assessments. From a regulatory point of view, the tool was created to facilitate the assessment process of inspection reports as well as licence renewal requests proposed by licensees. The project provides a concise and logical interface between the user and diverse resources involved in performing regulatory activities. These include links to standards, CNSC license documents, operating experience from other regulators and licensees, electronic banks of research documents, journal articles, studies, and previously submitted licensee reports and responses. The concept of knowledge mapping was applied using Excel and Access software in order to achieve these goals. This software uses an approach toward associating related concepts, which is modeled after the way in which the human brain is believed to acquire and assimilate new knowledge into its existing framework. This results in a network that is intuitively set up and conducive to the accumulation of further knowledge and resources. This paper provides an abridged account of the theory governing concept mapping, not to mention its origins and the impact generated by its application in an organizational milieu. In addition, an indication of those successful integrations of concept mapping into large organizations, both commercial and scientific, is