Relevant bibliographies by topics / Steam generators

Contents

  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Steam generators'

Author: Grafiati
Published: 4 June 2021
Last updated: 10 May 2023

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Journal articles on the topic "Steam generators"

1

Yefimov, Olexander, Larysa Tiutiunyk, Tetyana Harkusha, Tetyana Yesipenko, and Anastasiia Motovilnik. "Mathematical model of the horizontal steam generator PGV-1000." NTU "KhPI" Bulletin: Power and heat engineering processes and equipment, no. 1-2 (December 28, 2022): 53–57. http://dx.doi.org/10.20998/2078-774x.2022.01.06.

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The materials of the article consider the design characteristics of horizontal steam generators PGV-1000 for WWER NPPs. The NPP steam generator, in particular the PGV-1000 steam generator, is a specific heat exchange unit. This unit, together with a nuclear reactor and a steam turbine, is one of the main equipment of multi-circuit (double-circuit) steam turbine NPPs. The steam generator produces working steam using heat dissipated from the reactor core by the cooling medium and sent to the heat exchange surface of the steam generator. NPP steam generators, connecting the contours of the coolant and the working substance, equally belong to each of them. The heat-absorbing medium in the steam generator is the working substance (water, steam). PGV-1000 type NPP steam generators with pressurized water reactors produce dry saturated steam. The requirement to maintain high purity of the coolant is due to the heat transfer surfaces of such steam generators made of austenitic stainless steel with electropolished surfaces. The design characteristics of modern horizontal steam generators such as PGV-1000 of various modifications provide high technical and economic performance of NPP units with WWER-1000 and high maintainability, which allows to extend the service life of domestic NPPs.
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Egorov, Mikle, Ivan Kasatkin, Ivan Kovalenko, Irina Krectunova, Nataliya Lavrovskaya, and Nadezhda Litvinova. "Russian and foreign steam generators for NPP power units with wet steam turbines." E3S Web of Conferences 178 (2020): 01007. http://dx.doi.org/10.1051/e3sconf/202017801007.

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The main aim of the current study is to analyze advantages and shortcomings of horizontal and vertical types of steam generator design. Design solutions and experience of operation of steam generators of horizontal type accepted in Russia and of vertical type applied by Westinghouse, Combustion Engineering, Siemens, Mitsubishi, Doosan were analyzed within the framework of the present study. It was established that steam generator equipment of horizontal type is characterized by disadvantages of design, technological and operational nature. Thus, horizontal steam generators with dimensions permissible for railroad transportation and, for VVER-1200 with reactor vessel diameter equal to 5 m, by water transport as well, have exhausted the possibilities for further significant increase of the per unit electric power. The demonstrated advantages of vertical-type steam generators are as follows: 1) absence of stagnant zones within the second cooling circuit; 2) uniformity of heat absorption efficiency of the heating surface that ensures improved conditions for moisture separation; 3) increased temperature drop with parameters of generated steam elevated by 0.3 – 0.4 MPa. Conclusion was made on the advisability of introduction of steam generators with vertical-type layout in the Russian nuclear power generation.
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Yefimov, Olexander, Larysa Tiutiunyk, Harkusha Tetyana, Yesypenko Tetyana, and Anastasiia Motovilnik. "DESIGN CHARACTERISTICS OF THE HORIZONTAL STEAM GENERATOR PGV-1000." Bulletin of the National Technical University "KhPI". Series: Innovation researches in students’ scientific work, no. 2 (February 27, 2023): 8–12. http://dx.doi.org/10.20998/2220-4784.2022.02.02.

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The materials of the article consider the design characteristics of horizontal steam generators PGV-1000 for WWER NPPs. The NPP steam generator, in particular the PGV-1000 steam generator, is a specific heat exchange unit. This unit, together with a nuclear reactor and a steam turbine, is one of the main equipment of multi-circuit (double-circuit) steam turbine NPPs. The steam generator produces working steam using heat dissipated from the reactor core by the cooling medium and sent to the heat exchange surface of the steam generator. NPP steam generators, connecting the contours of the coolant and the working substance, equally belong to each of them. The heat-absorbing medium in the steam generator is the working substance (water, steam). PGV-1000 type NPP steam generators with pressurized water reactors produce dry saturated steam. The requirement to maintain high purity of the coolant is due to the heat transfer surfaces of such steam generators made of austenitic stainless steel with electropolished surfaces. WWER reactors do not allow the coolant to boil in the core, so the temperature of the coolant at the outlet of the reactor (at the inlet to the steam generator) is always lower than the saturation temperature corresponding to the water pressure in the coolant circuit. Underheating of the coolant to the saturation temperature (approximately 25 ÷ 30 °С) guarantees the exclusion of vaporization even in the most loaded reactor channels. The design characteristics of modern horizontal steam generators such as PGV-1000 of various modifications provide high technical and economic performance of NPP units with WWER-1000 and high maintainability, which allows to extend the service life of domestic NPPs.
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4

Green, S. J., and G. Hetsroni. "PWR steam generators." International Journal of Multiphase Flow 21 (December 1995): 1–97. http://dx.doi.org/10.1016/0301-9322(95)00016-q.

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5

Egorov, Mikhail Yu. "Vertical steam generators for VVER NPPs." Nuclear Energy and Technology 5, no. 1 (March 20, 2019): 31–38. http://dx.doi.org/10.3897/nucet.5.33980.

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Steam generators for NPPs are the important large-sized metal consuming equipment of nuclear power installations. Efficiency of steam generator operation determines the overall service life of the whole nuclear facility. The main aim of the current study is to analyze advantages and shortcomings of horizontal and vertical types of steam generator design. This analysis is aimed at the development of recommendations for designing advanced steam generators for future Russian units of NPPs with VVER reactors of increased power. Design solutions and fifty-year experience of operation of 400 steam generators of horizontal type accepted in Russia and of vertical type applied by Westinghouse, Combustion Engineering, Siemens, Mitsubishi, Doosan were analyzed within the framework of the present study. Advantages and drawbacks of both types of equipment determining the development of conditions of the operating processes were also identified and systematized. Currently NPPs equipped with VVER are characterized with extended surface area of containment shells due to the application of four-loop design configuration and horizontal-type steam generators. It was established that steam generator equipment of horizontal type is characterized by such inherent disadvantages of design, technological and operational nature as the following: 1) small height and volume of the vapor space above the evaporation surface reducing separation capabilities and the capacity of the equipment as a whole; 2) impossibility of organizing separate single-phase pre-boiling section. Because of the above, horizontal steam generators with dimensions permissible for railroad transportation and, for VVER-1200 with reactor vessel diameter equal to 5 m, by water transport as well, have exhausted the possibilities for further significant increase of the per unit electric power. The demonstrated advantages of vertical-type steam generators were as follows: 1) absence of stagnant zones within the second cooling circuit, and, consequently, of hold-ups in them; 2) uniformity of heat absorption efficiency of the heating surface ensuring, as well, improved conditions for moisture separation; 3) high degree of moisture removal from steam-water mixture due to the combination of moisture separating elements of chevron and swirl-vane types; 4) increased temperature drop with parameters of generated steam elevated by 0.3 – 0.4 MPa. Conclusion was made on the advisability of introduction of steam generators with vertical-type layout in the Russian nuclear power generation. Practical tasks that need to be addressed in order to ensure introduction of vertical steam generators at NPPs with high-power VVER reactors were formulated.
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Kovalchuk, V., I. Kozlov, O. Dorozh, and A. Machkov. "EFFICIENCY OF STEAM GENERATORS AT NUCLEAR POWER PLANTS." Odes’kyi Politechnichnyi Universytet Pratsi 2, no. 64 (2021): 28–35. http://dx.doi.org/10.15276/opu.2.64.2021.04.

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The possibility of a comprehensive assessment of steam generators efficiency at nuclear power plants with water-water reactors, based on the indicator of OEE (overall equipment effectiveness) is considered. It is proposed to consider efficiency as the probability of functioning from the standpoint of availability, performance and product quality.The aim of the work is to evaluate the possibility of using the OEE indicator to analyze the efficiency of NPP steam generators in complex conditions: reactor − steam generator − turbine. Achieving this goal will provide a comprehensive indicator of monitoring the efficiency of steam generating systems and have a tool for systematic monitoring of steam generators. To assess the organizational and environmental efficiency of the organizational structure, individual, group and integrated indicators are proposed, which reflect the share or decrease of the absolute indicator in the system compared to the baseline. The study is based on the analysis of long-term performance of units with steam generators PG-1000, which are comparable. It is shown that the main element of the steam generation system, which determines its efficiency, is the heat generating source. The contribution to the efficiency of all aspects of operation is estimated. It is shown that the efficiency index of OEE allows to characterize the efficiency of steam generators operation at nuclear power plants with water-water reactors, and can be used to monitor and control the process of their operation. In result of research, it is defined that steam generator efficiency increases in process of achievement of the maximum value of its productivity.
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CHO, SUNG-KEUN, CHANG-SUNG SEOK, BONG-KOOK BAE, and JAE-MEAN KOO. "EVALUATION OF THE HOOP TENSILE PROPERTIES OF A STEAM GENERATOR TUBE." International Journal of Modern Physics B 20, no. 25n27 (October 30, 2006): 4129–34. http://dx.doi.org/10.1142/s0217979206040970.

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The steam generators in a pressurized water reactor (PWR) are large heat exchangers that use the heat from the primary reactor coolant to make steam on the secondary-side to drive turbine generators. Hoop stress is known to be the main cause of fracture of inner pressurized tubes such as the steam generator tube. However, because the steam generator tube is too small to be manufactured to a standard tensile specimen in the hoop direction, the axial tensile properties of the steam generator tube (or original material properties) instead of hoop tensile properties have been used to estimate the fracture properties of a steam generator tube. In this study, we have conducted not only axial tensile tests but also ring tensile tests. From these test, both the axial and hoop tensile properties of steam generator tubes were obtained, and the reliability of the hoop tensile properties were confirmed by burst test of a real steam generator tube.
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Walter, Heimo, and Wladimir Linzer. "Flow Stability of Heat Recovery Steam Generators." Journal of Engineering for Gas Turbines and Power 128, no. 4 (March 1, 2004): 840–48. http://dx.doi.org/10.1115/1.2179469.

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This paper presents the results of theoretical flow stability analyses of two different types of natural circulation heat recovery steam generators (HRSG)—a two-drum steam generator—and a HRSG with a horizontal tube bank. The investigation shows the influence of the boiler geometry on the flow stability of the steam generators. For the two-drum boiler, the steady-state instability, namely, a reversed flow, is analyzed. Initial results of the investigation for the HRSG with a horizontal tube bank are also presented. In this case, the dynamic flow instability of density wave oscillations is analyzed.
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Blokhina, A., S. Lyakishev, and O. Korotkova. "POWER CALCULATION OF STRAIGHT-PIPE STEAM GENERATOR WITH SODIUM COOLANT." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, no. 1 (March 26, 2021): 152–61. http://dx.doi.org/10.55176/2414-1038-2021-1-152-161.

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The article investigates the influence of coolant flow profile nonlinearity through straight-pipe steam generators tube assembly for fast neutron reactor with sodium coolant on heat exchanger power. When designing steam generators, a very important task is to correctly calculate the output parameters of the steam generator, especially the power. For the plants without reference solutions it is necessary to perform a deep analysis of the factors affecting on calculated parameters and to incorporate these parameters in the codes. An example of a new plant that do not have analogues is a shell-type steam generator for perspective fast neutron plants with liquid metal sodium coolant. The application of new solutions in steam generators design requires experimental and calculational justification of thermal hydraulic with the use of modern calculation codes. Power calculation of steam generator is carried out by thermohydraulic code “KORSAR/GP”, “PGN-2K”. One of the assumptions in coolant path parameters calculation model (tube space) is a uniform velocity profile by cross-section of tube assembly. It’s also accepted, that each heat exchange tube has the same expense of feed water. On the other hand calculational CFD and experimental studies at aerodynamic model of steam generator showed the presence of significant unevenness of coolant expense by tube space cross-section which is not taken into account in thermohydraulic calculations. The article contains the methodic of accounting for the known uneven coolant flow profile by tube space cross-section in liquid metal steam generator calculation. Based on the results obtained, measures to improve power output and reliability are proposed.
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Korolev, Vladimir I. "ANALYSIS OF THERMOHYDRAULIC CHARACTERISTICS OF STEAM GENERATOR CASSETTES OF THE RITM-200 REACTOR PLANTS OF THE UNIVERSAL NUCLEAR ICEBREAKERS." Vestnik Gosudarstvennogo universiteta morskogo i rechnogo flota imeni admirala S. O. Makarova 14, no. 5 (December 14, 2022): 759–74. http://dx.doi.org/10.21821/2309-5180-2022-14-5-759-774.

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Currently, a new icebreaking fleet including universal nuclear icebreakers of the new generation of project 22220 is being intensively built in Russia. The nuclear icebreakers of this project such as “Arctic” and “Siberia” are already in operation. The “Ural” “nuclear icebreaker is almost ready for commissioning. “Yakutia” and “Chukotka” icebreakers as well as the icebreaker-leader of project 10510 are under construction. The integrated layout of the main equipment of the reactor plant had been used for the first time on the nuclear icebreakers of the new generation of these projects. The application of the new principle of equipment layout required switching to a fundamentally new method of forming the heat exchange surface of steam generators. As a result, straight-tube modular structures assembled into cassettes of steam generators were used. The approach to providing feed water supply and steam removal from steam generators has also changed, which significantly affects the thermohydraulic processes in the cassettes and leads to a decrease in the temperature of steam at the outlet of steam generators. In addition, the heat exchange surface of the modules in the longitudinal section forms annular channels with two-way heating by a coolant, which also affects the steam generators characteristics formation. In particular, the coefficients of heat transfer and heat transfer over the phase sections of heat exchange, the average logarithmic heat heads and the distribution of the heat exchange surfaces of the sections have changed compared to the previously used coil steam generators. At the same time, the hydraulic resistance of the flows through the working fluid and the coolant has decreased which reduces energy costs for pumping media and creates favorable conditions for improving natural circulation. In this paper, the task is to analyze the features of the thermohydraulic characteristics of steam generator cassettes, to compare them with similar characteristics of coil steam generators in operation.
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Dissertations / Theses on the topic "Steam generators"

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CABRAL, EDUARDO L. L. "Modelo matematico para o estudo do comportamento dinamico de geradores de vapor de tubos em U com circulacao natural." reponame:Repositório Institucional do IPEN, 1985. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9256.

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IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Waite, Jason S. "The application of brush seals to steam turbine generators." Thesis, Northumbria University, 1999. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.367419.

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Rummens, Helena E. C. "The thermalhydraulics of tube-support fouling in nuclear steam generators." Thesis, National Library of Canada = Bibliothèque nationale du Canada, 1999. http://www.collectionscanada.ca/obj/s4/f2/dsk1/tape8/PQDD_0021/NQ48334.pdf.

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Tulloch, Sam. "The use of dispersants in pressurised water reactor steam generators." Thesis, University of Birmingham, 2011. http://etheses.bham.ac.uk//id/eprint/706/.

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Environmental degradation promoted by the presence of sludge piles in the steam generators of Pressurised Water Reactors (PWR) can pose a threat to their safe and continuous operation. The use of dispersants can reduce the rate at which sludge accumulates. Polyacrylic acid (PAA) is currently the only dispersant used in PWRs. Settling rate tests identified several dispersants with the potential to outperform PAA, notably Hydroxyethylidene-1,1-diphosphonic Acid (HEDP). To estimate the dispersant concentration required during plant operation, optimum concentrations were identified for both PAA and HEDP. Nuclear magnetic resonance spectrometry was used to investigate the thermal stability of HEDP between 230 and 270oC, revealing that HEDP decomposes more rapidly than PAA. The dominant HEDP decomposition product was shown to be orthophosphate but several other long lived intermediate products were detected. The effect of dispersants on the environmental degradation of grade 316 stainless steel was determined by electrochemical methods and by constant extension rate tests. Rates of general corrosion measured by linear polarisation resistance and electrochemical impedance spectroscopy were shown to be very low (on the order 10-5 mm/year) under aerated and deaerated conditions both at room temperature and at 70oC. Corrosion rates were slightly reduced in the presence of PAA and HEDP. Constant extension rate tests demonstrated that neither PAA nor HEDP promote stress corrosion cracking at 250oC. It was concluded HEDP would not be suitable for use in PWRs due to its rapid thermal degradation rate. The decomposition products were shown to rapidly concentrate in steam generators thereby preventing accurate control of water chemistry.
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Rummens, Helena E. C. (Helena Elisabeth Cornelia) Carleton University Dissertation Engineering Mechanical and Aerospace. "The Thermalhydraulics of tube-support fouling in nuclear steam generators." Ottawa, 1999.

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Rehfeldt, Sebastian [Verfasser], and Günter [Akademischer Betreuer] Scheffnecht. "Radiative heat transfer in oxy-fuel steam generators / Sebastian Rehfeldt ; Betreuer: Günter Scheffnecht." Stuttgart : Universitätsbibliothek der Universität Stuttgart, 2016. http://d-nb.info/1118507487/34.

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Damon, Randolph. "A probabilistic fracture mechanics model for the tubing degradation of the Koeberg steam generators." Master's thesis, University of Cape Town, 2004. http://hdl.handle.net/11427/5571.

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Includes bibliographical references (leaves 115-124).
The susceptibility of Steam Generator (SG) Tubes (particularly tubes manufactured from Inconel 600) to primary water stress corrosion cracking (PWSCC) has been a major concern to the nuclear industry since 1971, when this phenomenon was first observed outside the laboratory. Since the susceptibility of Inconel 600 to PWSCC in pure water was first established as a certainty, various electricity utilities have devoted considerable resources to determining the exact nature of this degradation phenomenon and to predicting its consequences. Whereas, the study of the nature of this degradation mechanism has led to many conflicting conclusions, the predictive methods developed have been more successful. Initially, the establishment of the Leak Before Risk of Break principle has allowed various utilities to justify SG operation with cracked tubes. Later, the development of probabilistic simulation methods, most notably by EdF and the Jozef Stefan Institute (JSI), have led to further justification of the existing maintenance regimes and also allowed the ability to do sensitivity studies with regards to various influencing parameters. The current maintenance regime at the Koeberg Nuclear Power Station (near Cape Town, South Africa) utilises tube plugging (when cracks exceed a certain length limit) and online leak detection (based on detection of radioactive Nitrogen 16) as the main means of ensuring safe SG operation. However, the plugging limit used is based on a conservative deterministic analysis, which may be penalising in some situations. Thus, the need for risk-based SG life-time optimisation was evident.
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RODRIGUES, LUIZ A. H. "Modelagem teorica-experimental da equacao da quantidade de movimento para geradores de vapor de reatores PWR." reponame:Repositório Institucional do IPEN, 1994. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10391.

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IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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D'Angelo, Ricardo Andres. "The effects of dispersants on the fouling and the blockage deposits in pressurized water reactor steam generators." Thesis, Paris 6, 2015. http://www.theses.fr/2015PA066767.

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L’acide polyacrylique (APA) est injecté dans les générateurs vapeurs (GV) des réacteurs à eau sous pression pour limiter leur encrassement et leur colmatage. Le but de cette étude est de déterminer l’action de l’APA sur les dépôts présents dans les GV, et le caractère préventif du dispersant vis-à-vis de la formation de dépôts. Dans un premier temps, des dépôts de magnétite ont été soumis aux conditions rencontrées en phase liquide dans les GV, avec et sans APA en boucle d’essai. La concentration en fer en sortie de la boucle augmente significativement à partir d’une injection de 1 mg.kg-1 d’APA. Quelle que soit la concentration d’APA testée, la caractérisation après essai des échantillons montre une diminution localisée au sommet des octaèdres de magnétite et une couche d’hématite en surface du dépôt. Cependant cette couche a une épaisseur d’autant plus importante que la concentration en APA testée est grande. Dans la seconde partie, l’effet de l’APA sur la formation du dépôt sur un tube type GV a été investigué, dans une boucle simulant les conditions diphasiques des GV. Les résultats montrent une augmentation de la concentration en fer en sortie de la boucle avec l’injection d’APA. Les caractérisations effectuées sur les échantillons ont, par ailleurs, montré une diminution de l’épaisseur de la couche d’oxyde formée et un changement de la magnétite en hématite à l’interface liquide-vapeur. Ces deux études et une étude complémentaire sur la complexation de fer par l’APA à 25oC, ont permis de proposer des hypothèses expliquant le mécanisme d’action de l’APA sur les espèces du fer
Polyacrylic acid (PAA), is injected into Pressurized Water Reactor steam generators (SGs) to limit the fouling and the blockage of the SGs units, which decreases thermal performance and increase maintenance costs. The aim of this study is to determine the action mechanism(s), the efficiency and the pre-emptive nature of PAA on the fouling and blockage deposits in SGs. As a first step in this work, magnetite deposits were formed. These deposits were used to investigate the effects of PAA on pre-existing magnetite deposits in a dedicated experimental loop having the physico-chemical conditions of the SGs in liquid flow. Iron concentration at the loop exit increased with the injection of at least 1 mg.kg-1 PAA. Sample characterizations show a localized diminution of the magnetite octahedra peaks at all tested PAA concentrations and the formation of a thicker hematite layer at higher PAA concentrations. This work was followed by the investigation of the effect of PAA on magnetite suspensions and the formation of deposit layers onto the passive layer in the presence of PAA. This was performed in a dedicated biphasic loop having the physico-chemical conditions of SGs. Results indicated an increased iron concentration at the loop exit with the injection of PAA. Characterizations of the samples indicated the apparent diminution of oxide layer thickness and a change in oxide phase from magnetite to hematite at the liquid-vapor interface. The study of these effects and a complexation study performed at 25 oC, led to the formation and proposal of hypotheses to explain the underlying action mechanisms of PAA
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Gerken, Lisa M. "Long-Term Cooling of an SBLOCA: Boron Precipitation in the Core, Boron Dilution in the Steam Generators." Thesis, Virginia Tech, 2014. http://hdl.handle.net/10919/24906.

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When soluble boron is used to control reactivity, there are two particular events which can challenge long-term core cooling (LTCC) during the small break loss-of-coolant accident (SBLOCA): boron precipitation and boron dilution. The initial consequences of the SBLOCA are mitigated by the emergency safety systems, but the core continues to boil. As boron is less volatile than steam, the steam is virtually boron free. All the boron remains in the core, the boron concentration in the core rises. If the solubility limit is reached, precipitation could occur. The boron precipitation event was historically considered to be bounded by the large break accident. However, there are characteristics of the SBLOCA which cannot be neglected and an SBLOCA specific methodology is required. On the opposite end of the boron concentration spectrum is the SBLOCA boron dilution event. The steam generators remove heat from the primary and condense the steam. The condensation of the boron-free steam can result in the accumulation of a deborated slug of water. If natural circulation restarts, the slug can be transported toward the core and potentially reduce the core boron concentration enough to induce a recriticality. This thesis describes two analytical methodologies for these SBLOCA LTCC events. The two methodologies have a similar approach. Both use transient system analyses for inputs to and justification of the follow-on boron concentration calculations. For boron precipitation, a maximized concentration is calculated with the Small Break Boron Precipitation model. For boron dilution, a minimized core inlet concentration is calculated using computational fluid dynamics.
Master of Science
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Books on the topic "Steam generators"

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Tapping, R. L. Steam generators: Problems and prognosis. Chalk River, Ont: Heat Exchanger Technology Branch, Chalk River Laboratories, 1997.

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Annaratone, Donatello. Steam generators: Description and design. Berlin: Springer, 2008.

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International Seminar on Horizontal Steam Generators (4th 1997 Lappeenranta, Finland). Fourth International Seminar on Horizontal Steam Generators. Lappeenranta, Finland: Lappeenranta University of Technology, 1997.

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International Seminar on Horizontal Steam Generators (3rd 1994 Lappeenranta, Finland). Third International Seminar on Horizontal Steam Generators. Lappeenranta, Finland: Lappeenranta University of Technology, 1995.

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American Society of Mechanical Engineers. Winter Meeting. Thermal hydraulics of nuclear steam generators/heat exchangers. New York: American Society of Mechanical Engineers, 1988.

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International, Seminar on Horizontal Steam Generators (5th 2001 Lappeenranta Finland). Proceedings of Fifth International Seminar on Horizontal Steam Generators. Lappeenranta: Lappeenranta University of Technology, 2001.

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Junge, Hans Dieter. Pocket dictionary of steam generators.: Taschenwörterbuch Dampferzeuger. Deutsch, Englisch. Berlin: Verlag für Architektur und technische Wissenschaften, 1986.

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IEEE Power Engineering Society. Power Generation Committee., ed. IEEE recommended practice for functional and performance characteristics of control systems for steam turbine-generator units. New York, NY: Institute of Electrical and Electronics Engineers, 1992.

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Walter, Zörner, ed. Steam turbine generators process control and diagnostics: Modern instrumentation for the greatest economy of power plants. Erlangen: Publicis MCD, 1996.

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IEEE Power Engineering Society. Power Generation Committee., ed. IEEE recommended practice for functional and performance characteristics of control systems for steam turbine-generator units. New York, NY: Institute of Electrical and Electronics Engineers, 1985.

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Book chapters on the topic "Steam generators"

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Jackson, Benjamin W. "Steam Generators." In Power Plant Engineering, 185–217. Boston, MA: Springer US, 1996. http://dx.doi.org/10.1007/978-1-4613-0427-2_7.

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Kolev, Nikolay Ivanov. "Steam Generators." In Multiphase Flow Dynamics 5, 311–61. Cham: Springer International Publishing, 2015. http://dx.doi.org/10.1007/978-3-319-15156-4_8.

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Kolev, Nikolay I. "Steam generators." In Multiphase Flow Dynamics 4, 293–306. Berlin, Heidelberg: Springer Berlin Heidelberg, 2009. http://dx.doi.org/10.1007/978-3-540-92918-5_8.

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Kolev, Nikolay Ivanov. "Steam generators." In Multiphase Flow Dynamics 5, 301–54. Berlin, Heidelberg: Springer Berlin Heidelberg, 2011. http://dx.doi.org/10.1007/978-3-642-20601-6_8.

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Armbruster, Stanley A. "Steam Turbine Generators." In Power Plant Engineering, 218–49. Boston, MA: Springer US, 1996. http://dx.doi.org/10.1007/978-1-4613-0427-2_8.

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Guidez, Joël, and Gérard Prêle. "The Steam Generators." In Superphenix, 147–57. Paris: Atlantis Press, 2017. http://dx.doi.org/10.2991/978-94-6239-246-5_12.

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Girard, Sylvain. "Clogging of Recirculating Nuclear Steam Generators." In Physical and Statistical Models for Steam Generator Clogging Diagnosis, 3–13. Cham: Springer International Publishing, 2014. http://dx.doi.org/10.1007/978-3-319-09321-5_2.

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Collier, J. G. "Nuclear Steam Generators and Waste Heat Boilers." In Two-Phase Flow Heat Exchangers, 659–82. Dordrecht: Springer Netherlands, 1988. http://dx.doi.org/10.1007/978-94-009-2790-2_22.

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Todreas, Neil E., Mujid S. Kazimi, and Mahmoud Massoud. "Thermal Analysis of Steam Generators and Condensers." In Nuclear Systems Volume II, 393–442. 2nd ed. Boca Raton: CRC Press, 2021. http://dx.doi.org/10.1201/9780429157608-11.

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Cao, Huasong, Peiwei Sun, and Jianmin Zhang. "Power Optimization of Steam Generators Under Fault Conditions." In Proceedings of The 20th Pacific Basin Nuclear Conference, 817–33. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2314-9_75.

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Conference papers on the topic "Steam generators"

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Conlon, William M., Pete Johnson, and Robert Hanson. "Superheated Steam From CLFR Solar Steam Generators." In ASME 2011 Power Conference collocated with JSME ICOPE 2011. ASMEDC, 2011. http://dx.doi.org/10.1115/power2011-55174.

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AREVA Solar has designed, constructed and demonstrated the first successful Once Through Solar Steam Generator (SSG) to deliver superheated steam without intermediate heat transfer fluids. Deployed at the Kimberlina Solar Thermal Power Station, SSG4 represents the state of the art for solar steam production, for stand-alone power generation and augmentation of fossil fueled steam cycles. The ASME Section I boiler was designed, constructed, stamped and commissioned during 2010, and includes a novel Model Predictive Control system capable of maintaining any two of three steam conditions (flow, pressure, temperature) under varying solar input. During field trials in September 2010, exit steam conditions were maintained at 60 +/− 3 bar and 370 +/− 20C during steady and transient conditions, while steam flow consistently exceeded predictions. In a “lights-out” test, simulating complete instantaneous cloud cover, SSG4 had sufficient thermal inertia to supply more than 18 minutes of superheated steam. AREVA Solar’s SSGs incorporate a 400m long tube bundle within an elevated insulated cavity receiver, onto which sunlight is concentrated by reflectors. The multi-pass tube bundle arranges superheater tubes in the high flux regions, and economizer/evaporator tubes in lower flux regions. This assures sufficient heat flux to sustain superheated steam temperatures throughout the operating day, and also reduces the average bundle temperature to reduce radiant heat losses. Boiler tubes were prepared in AREVA Solar’s factory to improve their absorption of solar energy and reduce radiant heat losses. The inverted cavity maintains a stagnant air layer between the tube bundle and a glass cover below the boiler tube supports, to reduce convective heat loss. SSG4 was designed for a Maximum Allowable Working Pressure of 105 bara, and a Maximum Mean Wall Temperature of 482C in the superheater section. AREVA Solar is the first Concentrated Solar Power provider with an ASME “S” Stamp and National Board authorization. Following the initial trials at 370C, the SSG is expected to operate at 450C superheated steam temperature. This paper describes the design, construction, commissioning, and testing of the Compact Linear Fresnel Reflector (CLFR) SSG4.
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Hou, Su-Xia, Yun Tai, and Fu-Yu Zhao. "Analysis of Two-Phase Flow Instability for Steam Generator." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48533.

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Two-phase flow instability is an important problem that affects the running of steam generators in nuclear reactor systems. In this paper, two-phase flow instability in parallel channels of a steam generator are analyzed to disclose the mechanism of flow instability by using the frequency domain method. The mathematical expressions of heat transfer and flow for a steam generator are proposed, and the transfer function of the closed-loop system is deduced by using linearization and Laplace transfer. The steam generator’s stability is judged according to Nyquist stability criterion. Depending on this fundamental principal, the computer code is developed to analyze the stability of steam generators. The results displayed two conclusions; firstly, the increase of inlet orifices or mass flow rate enhances the stability of generator steam; secondly, the coupling interactions between channels and their external loop effects on the stability of generator steam are not ignored. The result show that the effects are non-monotonic on the stability of generator steam.
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Prabhu, Padmanabha J., and Damian A. Testa. "Steam Generator Asset Management Program." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29471.

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The Steam Generator Asset Management Program (SGAMP) is a long term program designed to maximize the performance and reliability of the steam generators. The SGAMP focuses on plant specific conditions and hence is applicable to the original or the replacement steam generators. It is recommended that the utility and the vendor form a joint steam generator management team (SGMT) to develop, monitor and implement a long-term plan to address steam generator operation, maintenance and life extension goals. The SGMT will consist of representatives from operations, chemistry, maintenance and engineering functions and will be responsible for making decisions related to the steam generators. The charter of the SGMT is to develop a steam generator strategic plan that will cost-effectively manage steam generator options. The strategic plan is consistent with the Steam Generator Program Guidelines (NEI 97-06 in the United States). The strategic plan is a living document and is revised periodically to incorporate inspection results, new technology developments, lessons learned and industry experience. Cost-benefit analyses of strategies may be performed to prolong steam generator operability through steam generator performance modeling (tube degradation, fouling, etc.), diagnostic tools, regulatory strategy, condition monitoring and operational assessment strategy, and maintenance strategy. The SGMT will provide input regarding potential maintenance of the steam generators with schedule and cost impacts for each outage. It will also recommend engineering evaluations to be performed in support of program goals and will develop short- and long-term recommendations. These recommendations will address action plans, performance measures and results. Secondary side inspection and cleaning strategy should be developed (techniques and frequency) to maximize performance cost-effectively. This paper is based on Westinghouse experience gained by working with several pressurized water reactor (PWR) plant operators in the United States (US).
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Walter, Heimo, and Wladimir Linzer. "Flow Stability of Heat Recovery Steam Generators." In ASME Turbo Expo 2004: Power for Land, Sea, and Air. ASMEDC, 2004. http://dx.doi.org/10.1115/gt2004-53040.

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In this paper the results of a theoretical stability analysis are presented. The investigation was done for two different types of natural circulation Heat Recovery Steam Generators (HRSG) — a two-drum steam generator and a HRSG with a horizontal tube bank. The investigation shows the influence of the boiler geometry on the stability of the steam generators. For the two-drum boiler the static instability, namely the reverse flow is analysed. First results of the investigations for the HRSG with a horizontal tube bank are also presented. In this case the dynamic flow instability of density wave oscillations is analysed.
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Tanner, Peter, Kent Rasmussen, Pete Johnson, Alex Hoermann, and Weili Cheng. "Control of Solar Steam Generators." In ASME 2011 Power Conference collocated with JSME ICOPE 2011. ASMEDC, 2011. http://dx.doi.org/10.1115/power2011-55175.

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Direct steam generation has distinct cost advantages over alternative methods for producing superheated steam from a solar collector system. We present results from a model predictive controller for a solar steam generator that is configured as a once through to superheat process using direct steam generation. The solar collector system modeled uses a linear Fresnel reflector field with a multi-tube receiver. The model predictive controller uses step response models that are developed offline using a dynamic simulation. The model predictive controller incorporates a feedforward signal based on absorbed solar power. Performance of the controller has been demonstrated at the Kimberlina Solar Power Station near Bakersfield, California. The results show the feasibility of a controller for a solar direct steam generation system that provides stable superheated steam temperature and pressure conditions through weather events.
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Nemcic, Kresimir, and Robert Brems. "Twenty Years of Operation in NPP Krsko." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22130.

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This paper presents the main design and fabrication improvements that were included in the new steam generators that were installed at Nuclear Power Plant Krsko in Slovenia in 2000. These improvements were a result of an on-going R & D effort associated with steam generator technology, aimed at increased reliability and better maintainability of new steam generators. The paper also provides basic information related to 20 years of operation of NPP Krsko with an emphasis on subjects related to steam generator performance and degradation. These include inservice inspection results on steam generators up to replacement, corrective actions, corrosion-erosion counter-measures, and replacement of condenser and moisture spearator reheaters with improved material. The paper also provides data regarding changes in feedwater and steam generator water chemistry, with water chemistry results before and after steam generator replacement. This paper shows that materials issues are very important for steam generator reliability, and it gives the reasons why specific materials were selected for replaced components. Finally, the paper demonstrates how the strategy used at NPP Krsko prolonged the useful life of the old steam generators until replacement steam generators could be designed, fabricated, and installed.
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Kehler, J. H. "Frequency regulation from steam turbine generators." In IEEE Power Engineering Society. 1999 Winter Meeting (Cat. No.99CH36233). IEEE, 1999. http://dx.doi.org/10.1109/pesw.1999.747554.

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Subudhi, M., and E. J. Sullivan. "Age-Related Degradation of Steam Generator Internals Based on Industry Responses to Generic Letter 97-06." In ASME 2002 Pressure Vessels and Piping Conference. ASMEDC, 2002. http://dx.doi.org/10.1115/pvp2002-1376.

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This paper presents the results of an aging assessment of the nuclear power industry’s responses to NRC Generic Letter 97-06 on the degradation of steam generator internals experienced at Electricite de France (EdF) plants in France and at a United States pressurized water reactor (PWR). Westinghouse (W), Combustion Engineering (CE), and Babcock & Wilcox (B & W) steam generator models, currently in service at U.S. nuclear power plants, potentially could experience degradation similar to that found at EdF plants and the U.S. plant. The steam generators in many of the U.S. PWRs have been replaced with steam generators with improved designs and materials. These replacement steam generators have been manufactured in the U.S. and abroad. During this assessment, each of the three owners groups (W, CE, and B&W) identified for its steam generator models all the potential internal components that are vulnerable to degradation while in service. Each owners group developed inspection and monitoring guidance and recommendations for its particular steam generator models. The Nuclear Energy Institute incorporated in NEI 97-06, “Steam Generator Program Guidelines,” a requirement to monitor secondary side steam generator components if their failure could prevent the steam generator from fulfilling its intended safety-related function. Licensees indicated that they implemented or planned to implement, as appropriate for their steam generators, their owners group recommendations to address the long-term effects of the potential degradation mechanisms associated with the steam generator internals.
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Brunin, Olivier, and Geoffrey Deotto. "Loads Applied to Clogged Steam Generators Internal Structures." In ASME 2009 Pressure Vessels and Piping Conference. ASMEDC, 2009. http://dx.doi.org/10.1115/pvp2009-77140.

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During their commissioning, steam generators are clean, which means there is no fouling of the heat transfer surface of tubes and no clogging of the flow area on the secondary side. Then sludge appears steadily at a slow pace during operation. Sludge initiates a partial loss of cooling capacity which is modeled by a fouling factor and which mainly results in vapor pressure decrease. Sludge also initiates a reduction of the secondary side flow area, known as clogging. Four safety-related issues are dependant on clogging [1]: the secondary water mass balance, the thermohydraulics oscillations, the tube vibration risk and the resistance of internal structures. This paper focuses on the last of these issues. A numerical application, based on the modeling of a fictitious steam generator, is detailed in this presentation. The order of magnitude is an 8-times increase of the loads in normal operating conditions in case of a typical 60% clogging ratio of the upper tube support plate, and a 12-times increase in case of incidental depressurization transient. These theoretical results emphasize the need to take these loads properly into account in the checking of the mechanical behavior of the internal structure of the steam generators in operation in case of significant sludge deposits.
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Pasha, Akber. "Acceptance Criteria for Heat Recovery Steam Generators Behind Gas Turbines." In ASME 1986 International Gas Turbine Conference and Exhibit. American Society of Mechanical Engineers, 1986. http://dx.doi.org/10.1115/86-gt-201.

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The design of a Heat Recovery Steam Generator behind a gas turbine depends upon various input parameters such as gas turbine exhaust flow, exhaust temperature, etc. Most of the input parameters are either measured with tolerances or calculated based on experimental correlations. The design of the heat recovery steam generator itself utilizes various correlations and empirical values. The errors or measurement tolerances in these variables affect the performance of the steam generator. This paper describes the various design parameters, the possible magnitude of errors in these parameters and the overall effect on the steam generator’s performance. By utilizing the information given in this paper, it is possible to develop a performance envelope based on the possible error margins of the input variables. The steam generator performance can be deemed acceptable if it is within this envelope.
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Reports on the topic "Steam generators"

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Geffraye, G., P. Bazin, and P. Pichon. CCFL in hot legs and steam generators and its prediction with the CATHARE code. Office of Scientific and Technical Information (OSTI), September 1995. http://dx.doi.org/10.2172/107026.

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Uvan Catton, Vijay K. Dhir, Deepanjan Mitra, Omar Alquaddoomi, and Pierangelo Adinolfi. Development of Design Criteria for Fluid Induced Structural Vibrations in Steam Generators and Heat Exchangers. Office of Scientific and Technical Information (OSTI), April 2004. http://dx.doi.org/10.2172/827838.

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Catton, Ivan, Vijay K. Dhir, O. S. Alquaddoomi, Deepanjan Mitra, and Pierangelo Adinolfi. Development of Design Criteria for Fluid Induced Structural Vibration in Steam Generators and Heat Exchangers. Office of Scientific and Technical Information (OSTI), March 2004. http://dx.doi.org/10.2172/822365.

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Belle R. Upadhyaya and J. Wesley Hines. On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers. Office of Scientific and Technical Information (OSTI), September 2004. http://dx.doi.org/10.2172/832717.

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Belle R. Upadhyaya and J. Wesley Hines. On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers. Office of Scientific and Technical Information (OSTI), September 2004. http://dx.doi.org/10.2172/832721.

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Kreider, M. A., G. A. White, R. D. Varrin, and P. J. Ouzts. Heat transfer characteristics of porous sludge deposits and their impact on the performance of commercial steam generators. Office of Scientific and Technical Information (OSTI), December 1998. http://dx.doi.org/10.2172/329562.

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Geelhood, B. D., and K. H. Abel. Neutron measurements of the fuel remaining in the TMI II once-through steam generators (OTSG'S). Office of Scientific and Technical Information (OSTI), February 1989. http://dx.doi.org/10.2172/6456962.

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Upadhyaya, Belle R., J. Wesley Hines, and Baofu Lu. On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers, Volumes 1, 2. Office of Scientific and Technical Information (OSTI), June 2005. http://dx.doi.org/10.2172/843182.

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Kurtz, R., R. Clark, E. Bradley, W. Bowen, P. Doctor, R. Ferris, and F. Simonen. Steam Generator Tube Integrity Program/Steam Generator Group Project. Office of Scientific and Technical Information (OSTI), May 1990. http://dx.doi.org/10.2172/6853141.

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Adams, J. P., and E. S. Peterson. Steam generator secondary pH during a steam generator tube rupture. Office of Scientific and Technical Information (OSTI), December 1991. http://dx.doi.org/10.2172/10139132.

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