Journal articles: 'High-temperature steam generator'

1

Rice, I. G. "Split Stream Boilers for High-Temperature/High-Pressure Topping Steam Turbine Combined Cycles." Journal of Engineering for Gas Turbines and Power 119, no. 2 (April 1, 1997): 385–94. http://dx.doi.org/10.1115/1.2815586.

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Research and development work on high-temperature and high-pressure (up to 1500°F TIT and 4500 psia) topping steam turbines and associated steam generators for steam power plants as well as combined cycle plants is being carried forward by DOE, EPRI, and independent companies. Aeroderivative gas turbines and heavy-duty gas turbines both will require exhaust gas supplementary firing to achieve high throttle temperatures. This paper presents an analysis and examples of a split stream boiler arrangement for high-temperature and high-pressure topping steam turbine combined cycles. A portion of the gas turbine exhaust flow is run in parallel with a conventional heat recovery steam generator (HRSG). This side stream is supplementary fired opposed to the current practice of full exhaust flow firing. Chemical fuel gas recuperation can be incorporated in the side stream as an option. A significant combined cycle efficiency gain of 2 to 4 percentage points can be realized using this split stream approach. Calculations and graphs show how the DOE goal of 60 percent combined cycle efficiency burning natural gas fuel can be exceeded. The boiler concept is equally applicable to the integrated coal gas fuel combined cycle (IGCC).

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ZHANG, Zhen, Xing-tuan YANG, and Huai-ming JU. "ICONE23-1474 DESIGN AND THERMAL-HYDRAULIC ANALYSIS OF SUPERCRITICAL STEAM GENERATOR IN HIGH-TEMPERATURE GAS-COOLED REACTOR." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2015.23 (2015): _ICONE23–1—_ICONE23–1. http://dx.doi.org/10.1299/jsmeicone.2015.23._icone23-1_216.

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Wang, Yan, Lei Shi, and Yanhua Zheng. "Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor." Science and Technology of Nuclear Installations 2017 (2017): 1–8. http://dx.doi.org/10.1155/2017/8521410.

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

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Zhao, Lin, Bikram Bhatia, Lenan Zhang, Elise Strobach, Arny Leroy, Manoj K. Yadav, Sungwoo Yang, et al. "A Passive High-Temperature High-Pressure Solar Steam Generator for Medical Sterilization." Joule 4, no. 12 (December 2020): 2733–45. http://dx.doi.org/10.1016/j.joule.2020.10.007.

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Liu, Song, Yaping Wei, Shiqiang Chen, Liu Qin, and Guoqiang Xie. "Development and Application of an Ultrahigh-Temperature Steam Generator." Advances in Materials Science and Engineering 2020 (September 29, 2020): 1–4. http://dx.doi.org/10.1155/2020/4243170.

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A device which could produce high-temperature steam quickly was designed. The overall structure of the heating device and the heating pipe of key components were introduced mainly, and the working process of the heating device was analyzed and discussed. This high-temperature steam device with high heating efficiency, produces steam fast and could realize precise temperature control. This device could expand application research of polymer materials and composite materials and provide key guide parameters in the process of technical research.

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Yang, Xingtuan, Yanfei Sun, Huaiming Ju, and Shengyao Jiang. "Procedure of Active Residual Heat Removal after Emergency Shutdown of High-Temperature-Gas-Cooled Reactor." Science and Technology of Nuclear Installations 2014 (2014): 1–10. http://dx.doi.org/10.1155/2014/583597.

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After emergency shutdown of high-temperature-gas-cooled reactor, the residual heat of the reactor core should be removed. As the natural circulation process spends too long period of time to be utilized, an active residual heat removal procedure is needed, which makes use of steam generator and start-up loop. During this procedure, the structure of steam generator may suffer cold/heat shock because of the sudden load of coolant or hot helium at the first few minutes. Transient analysis was carried out based on a one-dimensional mathematical model for steam generator and steam pipe of start-up loop to achieve safety and reliability. The results show that steam generator should be discharged and precooled; otherwise, boiling will arise and introduce a cold shock to the boiling tubes and tube sheet when coolant began to circulate prior to the helium. Additionally, in avoiding heat shock caused by the sudden load of helium, the helium circulation should be restricted to start with an extreme low flow rate; meanwhile, the coolant of steam generator (water) should have flow rate as large as possible. Finally, a four-step procedure with precooling process of steam generator was recommended; sensitive study for the main parameters was conducted.

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LEE, CHOON YEOL, JOONG HO KIM, JOON WOO BAE, and YOUNG SUCK CHAI. "DEVELOPMENT OF THE HIGH TEMPERATURE FRETTING WEAR SIMULATOR FOR STEAM GENERATOR." International Journal of Modern Physics B 24, no. 15n16 (June 30, 2010): 2603–8. http://dx.doi.org/10.1142/s0217979210065337.

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In nuclear power plant, fretting wear due to a combination of impact and sliding motions of the U-tubes against the supports and/or foreign objects caused by flow induced vibration, can make a serious problem in steam generator. A test rig, fretting wear simulator, is developed to elucidate fretting wear mechanism qualitatively and quantitatively. The realistic condition of steam generator of high temperature up to 320°C, high pressure up to 15 MPa, and water environment could be achieved by a test rig. The fretting wear simulator consists of main frame, water loop system, and control unit. Actual contact region under a realistic condition of steam generator was isolated using autoclave. Effects of various parameters such as the amounts of impact and sliding motions, applied loads and initial gaps and so forth are considered in this research. After the experiment, wear damage was measured by a three-dimensional profiler and the surface was also studied by SEM microscopically. Initial results were also presented.

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LEE, CHOON YEOL, JOON WOO BAE, YOUNG SUCK CHAI, and KYOOSIK SHIN. "EXPERIMENTAL STUDY ON IMPACT FRETTING WEAR OF INCONEL TUBES UNDER HIGH TEMPERATURE AND PRESSURE." International Journal of Modern Physics B 25, no. 31 (December 20, 2011): 4253–56. http://dx.doi.org/10.1142/s0217979211066702.

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In nuclear power plant, fretting wear caused by flow induced vibration (FIV) accompanied with impact force can make serious problems between U -tubes and egg-crates which are located in steam generators. In order to guarantee the reliability of the steam generator, design based on consideration of the damage due to the fretting wear of the U -tube is inevitable. The purpose of this study is to elucidate fretting wear mechanism qualitatively and quantitatively. First, finite element models are developed to analyze the dynamic characteristics and estimate the impact force in steam generators. Based on the numerical results, fretting wear simulation is performed according to the environment to which the actual steam generators in nuclear power plant are exposed. Initial experimental results are obtained for various experimental parameters and the effect of work rate and temperature on fretting wear is evaluated.

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9

Esch, Markus, Dietrich Knoche, and Antonio Hurtado. "Numerical discretization analysis of a HTR steam generator model for the thermal-hydraulics code trace." Nuclear Technology and Radiation Protection 29, suppl. (2014): 31–38. http://dx.doi.org/10.2298/ntrp140ss31e.

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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.

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LEE, CHOON YEOL, JOONG HO KIM, JOON WOO BAE, and YOUNG SUCK CHAI. "ERRATUM: "DEVELOPMENT OF THE HIGH TEMPERATURE FRETTING WEAR SIMULATOR FOR STEAM GENERATOR"." International Journal of Modern Physics B 25, no. 20 (August 10, 2011): 2789. http://dx.doi.org/10.1142/s0217979211058997.

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11

Seubert, B. T., T. P. Fluri, and W. J. Platzer. "Numerical Investigation of a High Temperature Stratified Storage with Integrated Steam Generator." Energy Procedia 49 (2014): 1003–14. http://dx.doi.org/10.1016/j.egypro.2014.03.108.

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12

Liu, Yan, Hui-lie Shi, Chun Gui, Xian-yuan Wang, and Rui-feng Tian. "Effect of Saturated Steam Carried Downward on the Flow Properties in the Downcomer of Steam Generator." Energies 12, no. 19 (September 24, 2019): 3650. http://dx.doi.org/10.3390/en12193650.

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The saturated water separated by the steam separator in a natural circulation steam generator may carry a small amount of saturated steam into the downcomer. The steam contacts subcooled water and condenses directly in the downcomer causing the variations in the pressure and steam quality and likely affecting the stability of the water cycle in the secondary loop. It is not conducive to core heat extraction and thus affects nuclear safety. The mathematical model of the downcomer was established in this study based on the internal structures of a natural circulation steam generator. The volume-of-fluid (VOF) and large-eddy-simulation (LES) models were used for analysis on FLUENT software (ANSYS, Pittsburgh, PA. USA) platform. The influence of direct contact condensation of top-down flowing steam on the flow properties in the downcomer of the steam generator under high pressure was studied. The trend of the temperature, pressure, and the void fraction were obtained by combining these models with the condensation model. Further, a one-dimensional calculation program based on the differential drop was also developed to assess the flow field in the downcomer. The calculation results are in good agreement with the experimental results which indicated that, when affected by the saturated steam carried downward, the flow temperature close to the exit of the downcomer rises slightly due to the absorption of the heat released by the steam condensation. Furthermore, the density corrected by the pressure-drop is more reliable than that corrected by the temperature. After the velocity in the downcomer has increased to a certain value, the sensitivity of steam quality to the subcooling degree in the downcomer begins to decline. The results in this paper can be used to perform stability analyses and to design steam generators. The results of research are helpful to the stability analysis and the design of a steam generator, and to improve the accuracy of the measurement of the steam generator operating parameters, thus enhancing the safety of Pressurize Water Reactor operating system.

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13

Li, Zhimo, Jiachun Li, Xiangli Dong, Bo Chen, and Qing Li. "Design of an electromagnetic induction steam generator device based on air source heat pump." MATEC Web of Conferences 355 (2022): 02059. http://dx.doi.org/10.1051/matecconf/202235502059.

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Aiming at the current problems of coal-fired boilers and electromagnetic induction steam generators for environmental pollution and high energy consumption, this article combines air source heat pumps and electromagnetic induction heating technology, and at the same time carries out the structure of the condensate tank and electromagnetic induction steam generator. Redesign. Through trial production and experimentation of the prototype, the results show that compared with traditional coal-fired boilers and separate electromagnetic induction heating technology to generate steam, this device not only achieves energy saving and environmental protection, but also the stability of the steam outlet temperature and the amount of steam generated. Compared with the use of electromagnetic induction heating alone, it has increased by 20%. It can be seen that the use of air source heat pump’ electromagnetic induction heating technology to generate steam saves energy and increases the amount of steam generated.

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14

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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15

Park, Gi Sung, Gyung Guk Kim, and Seon Jin Kim. "Sliding wear behaviors of steam generator tube materials in high temperature water environment." Journal of Nuclear Materials 352, no. 1-3 (June 2006): 80–84. http://dx.doi.org/10.1016/j.jnucmat.2006.02.043.

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16

Kim, Hyungmo, Eunmi Ko, Yong-Sun Ju, Jung Yoon, Jaehyuk Eoh, and Hyeong-Yeon Lee. "High-Temperature Design and Integrity Evaluation of a IHX-Combined Steam Generator in Generation IV Reactor." Transactions of the Korean Society of Mechanical Engineers - A 44, no. 11 (November 30, 2020): 813–20. http://dx.doi.org/10.3795/ksme-a.2020.44.11.813.

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17

Grabezhnaya, V., and A. Mikheyev. "EXPERIMENTAL STUDY OF THERMAL HYDRAULICS ON THE MODEL OF HELICAL COILED STEAM GENERATOR HEATED BY LIQUID LEAD WITH LONGITUDINAL AND TRANSVERSE FLOW." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2019, no. 1 (March 26, 2019): 132–51. http://dx.doi.org/10.55176/2414-1038-2019-1-132-151.

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The study of heat transfer in spiral coiled tubes is of great interest in view of the widespread use of such channels in engineering practice, in particular, in nuclear power engineering in the form of steam generators at research reactors and nuclear power plants. In the projected BREST-OD-300 reactor facility (RF), a configuration of helical coiled tubes is considered as a steam generator. Thermal hydraulic tests of the model steam generator RF BREST-OD-300 (version 2000) with helical coiled tubes with longitudinal coolant flow were carried out in SSC RF - IPPE at the SPRUT facility in 2011-2013 years. The test program was aimed to study heat transfer and thermal hydraulic stability of the steam generating tubes. Throughout the range of variation the regime parameters, regimes with a reversal circulation in the water loop have not been detected. Despite the fact that the results of conducted tests on the steam generator model gave extensive information on heat transfer in different zones of the steam generating tube, however, the insufficient number of heat transfer tubes in the module (only three) does not allow to conclude that the full hydrodynamic stability of BREST RF steam generator in all possible modes of operational parameters. On the other hand, in a real construction the motion of the heating coolant is omitted with flow around the bundle of tubes close to the transverse flow. Therefore, insufficient reasoning for the transferring results obtained on a three-tube model to a full-scale steam generator served as the basis for testing of a multi-tube full-height fragment model of a reduced diameter one row of the tube bundle actual steam generator. During the tests, there were no noises typical of the unstable operation of water loop. There were no oscillations of water and steam temperature, respectively, at the inlet to the collectors and out of the collectors. At high lead temperatures, the temperature of the superheated steam was always close to the inlet temperature of the lead. The tests showed the absence of the thermohydraulic instability, as in the case of longitudinal and transverse coolant flows in the investigated modes of lead and water parameters. Other parameters being equal, the steam temperature at the outlet of the steam generating tube in case of transverse flow was higher than in the case of longitudinal flow. The experimental data obtained during the testing are primarily necessary for the verification of codes that allow the correct calculation of the various operating modes of the BREST-OD-300 steam generator.

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Shigeta, Jun-Ichi, Yoshio Hamao, Hiroshi Aoki, and Ichiro Kajigaya. "Development of a Coal Ash Corrosivity Index for High Temperature Corrosion." Journal of Engineering Materials and Technology 109, no. 4 (October 1, 1987): 299–305. http://dx.doi.org/10.1115/1.3225981.

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Current development of Advanced Steam Cycle coal-fired power plants requires superheater and reheater tubing alloys which can withstand severe conditions for high temperature corrosion. A corrosion equation to predict corrosion rates for candidate alloys has been developed by a study of deposits removed from steam generator tubes and from test probes installed in a boiler, supplemented by laboratory studies using synthetic coal ash. The corrosion equation predicts corrosion for a particular coal as a function of its content of sulfur, acid-soluble alkalies, and acid-soluble aklaline earths. Good agreement was obtained between the corrosion equation and 6000-hour tests using probes of TP347H and 17-14 CuMo.

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Park, Chi Yong, and Jeong Keun Lee. "Wear Scar Progression of Impact-Fretting at Elevated Temperature for Steam Generator Tubes in Nuclear Power Plants." Key Engineering Materials 326-328 (December 2006): 1251–54. http://dx.doi.org/10.4028/www.scientific.net/kem.326-328.1251.

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Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.

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20

Dong, Li Yu, Zhi Wei Zhou, and Yang Ping Zhou. "Mathematical Model and Dynamic Characteristics of Spiral-Style Super-Critical Steam Generator Used HTGR." Advanced Materials Research 347-353 (October 2011): 1678–82. http://dx.doi.org/10.4028/www.scientific.net/amr.347-353.1678.

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Modular HTGR nuclear power plant because of inherent safety and high thermal efficiency shows good prospects for development. The current high-temperature reactor demonstration power plant (HTR-PM) using two thermal power of 250MW of modular HTGR with an electric power 211MWe turbine unit. As one development goals of multi-reactor with one turbine unit, millions of kilowatt nuclear power plant will use more reactor module and steam generator module more like demonstration power plant (HTR-PM) with 1000MWe supercritical turbine generator unit. spiral-style super-critical steam generator design, modeling is a key factor. Analyzing the structure and the characteristic of moderate spiral coil steam generator which is used in Modular HTGR demonstration power plant, from the mechanism of equipments, based on the law of quality conservation, energy conservation, momentum conservation, authors build up the full scope real time simulation mathematical model of super critical steam generator. The dynamic experiments of feed water disturbance, power disturbance, Helium flux disturbance are made on the basis of the model. The experiments show that the model of super critical steam generator has excellent dynamic characteristics.

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Qiu, Jin Rong, Chang Hong Peng, and Yun Guo. "Sensitivity Analyses of Steam Generator Tube Thickness on Induced-SGTR in SBO Accident." Applied Mechanics and Materials 341-342 (July 2013): 1338–41. http://dx.doi.org/10.4028/www.scientific.net/amm.341-342.1338.

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For nuclear power plant, station black-out (SBO) is the events that contribute significantly to the level-I core damage risk. For an SBO, it is assumed that both the off-site power and on-site diesel generators fail to supply alternating-current power for the plant systems. SBO induced steam generator tube rupture (SGTR) is a concern because the steam generator (SG) tubes are parts of the reactor coolant pressure boundary and failure of the SG tubes may lead to fission products bypassing the containment. The SG tube integrity may be challenged by high temperature and high pressure conditions and may have a potential to fail due to creep rupture. This study focuses on the probability of SBO induced SGTR accidents under the station blackout (SBO) with RCS integrity, seal LOCA and steam relief valves remaining stuck open for the reference plant. At last, the sensitivity of the tube thick is studied.

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Gue´rout, F. M., and N. J. Fisher. "Steam Generator Fretting-Wear Damage: A Summary of Recent Findings." Journal of Pressure Vessel Technology 121, no. 3 (August 1, 1999): 304–10. http://dx.doi.org/10.1115/1.2883707.

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Flow-induced vibration of steam generator (SG) tubes may sometimes result in fretting-wear damage at the tube-to-support locations. Fretting-wear damage predictions are largely based on experimental data obtained at representative test conditions. Fretting-wear of SG materials has been studied at the Chalk River Laboratories for two decades. Tests are conducted in fretting-wear test machines that simulate SG environmental conditions and tube-to-support dynamic interactions. A new high-temperature force and displacement measuring system was developed to monitor tube-to-support interaction (i.e., work-rate) at operating conditions. This improvement in experimental fretting-wear technology was used to perform a comprehensive study of the effect of various environment and design parameters on SG tube wear damage. This paper summarizes the results of tests performed over the past 4 yr to study the effect of temperature, water chemistry, support geometry and tube material on fretting-wear. The results show a significant effect of temperature on tube wear damage. Therefore fretting-wear. The results show a significant effect of temperature or tube wear damage. Therefore, fretting-wear tests must be performed at operating temperatures in order to be relevant. No significant effect of the type of water treatment on tube wear damage was observed. For predominantly impacting motion, the wear of SG tubes in contact with 410 stainless steel is similar regardless of whether Alloy 690 or Alloy 800 is used as tubing material or whether lattice bars or broached hole supports are used. Based on results presented in this paper, an average wear coefficient value is recommended that is used for the prediction of SG tube wear depth versus time.

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Tan, Jibo, Xiaoqiang Liu, Xuelian Xu, En-Hou Han, Xinqiang Wu, and Xiang Wang. "Environmentally Assisted Fatigue Evaluation Model of Alloy 690 Steam Generator Tube in High-Temperature Water." Corrosion 72, no. 5 (May 2016): 655–64. http://dx.doi.org/10.5006/1803.

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Lu, B. T., J. L. Luo, and Y. C. Lu. "Passivity degradation of nuclear steam generator tubing alloy induced by Pb contamination at high temperature." Journal of Nuclear Materials 429, no. 1-3 (October 2012): 305–14. http://dx.doi.org/10.1016/j.jnucmat.2012.06.021.

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BOGDANOVI, B. "A process steam generator based on the high temperature magnesium hydride/magnesium heat storage system." International Journal of Hydrogen Energy 20, no. 10 (October 1995): 811–22. http://dx.doi.org/10.1016/0360-3199(95)00012-3.

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Srinivasan, R., P. Chellapandi, and C. Jebaraj. "Structural design approach of steam generator made of modified 9Cr-1Mo for high temperature operation." Transactions of the Indian Institute of Metals 63, no. 2-3 (April 2010): 629–34. http://dx.doi.org/10.1007/s12666-010-0094-x.

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27

Zhang, Tong, Guihui Qiu, Hongying Yu, Peng Zhou, Shicheng Wang, Kaige Zhang, Qi Guo, Lu Ren, and Jian Xu. "The Fouling Behavior of Steam Generator Tube at Different Positions in the High-Temperature Water." Metals 11, no. 5 (April 22, 2021): 684. http://dx.doi.org/10.3390/met11050684.

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The fouling behavior of a steam generator (SG) tube was investigated at different positions after 500 h of immersion in high-temperature water. A triple-layer structure of fouling appeared at both the crevice position and the free span position, namely, the large, dispersedly distributed deposition layer on the top; the small and faceted outer layer; and the relatively continuous inner layer. There was no obvious positional effect on the thickness of the inner layer. However, in the crevice position, the density of the deposited particle and the thickness of the outer layer was much higher than those of the free span position. The tube support plate (TSP) made of 410 stainless steel contributed significantly to the fouling behavior of the SG tube in the crevice between the SG tube and the TSP.

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Begum, Shahida, A. N. Mustafizul Karim, and M. A. Shafii. "Investigation on Wall Thinning and Creep Damage in Boiler Tube due to Scale Formation." Advanced Materials Research 538-541 (June 2012): 1781–84. http://dx.doi.org/10.4028/www.scientific.net/amr.538-541.1781.

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Boiler is a closed vessel in which the water is heated up to convert it from the liquid phase to superheat steam at specified pressure by addition of heat. The tubes are operated continuously at high temperature due to the formation of scale which has lower conductivity than that of steel. The scale can be formed for various reasons of which tube geometries, flue gas and steam temperature are prominent. The remaining wall thickness decreases due to the formation of scale which eventually causes failure of the boiler tubes. In this investigation an iterative technique was used to determine the temperature distribution across the tube with the increase of operating time. The operating time was considered up to 160,000 hours. The remaining life of the steam generator tube was found by finding hoop stress and Larson Miller Parameter from the Larson Miller Parameter curve for SA213-T22 material. The remaining life of the steam generator tube was used to find cumulative creep damage. By utilizing finite element modelling software, ANSYS 9/ ANSYS 11 the temperature distribution across the steam generator tube was evaluated. The temperature distribution along with Larson Miller Parameter predicted the oxide scale thickness. It was also observed that different input parameters have pronounced affect on the formation of oxide scale inside the steam generator tube. By increasing the heat transfer rate across the wall, the oxide scale thickness was increased more rapidly than normal condition. It was also observed that due to formation of scale the thermal conductivity in the boiler tubes was affected and the remaining life of boiler tubes was decreased and accelerated creep damage.

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Jiang, Shuang, Jun Cai, J. W. Zhang, and Y. D. Wang. "Stress Corrosion Analysis of Steam Generator Tube of Nuclear Power Plant by Finite Elements." Applied Mechanics and Materials 441 (December 2013): 295–99. http://dx.doi.org/10.4028/www.scientific.net/amm.441.295.

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The steam generator is operated under the pressure of 15.5MPa and the temperature 327°C. Under this working condition the junction of the heat transfer tube and tube-sheet is accident-prone areas of the steam generator. By using ANSYS to carry out a thermal structure coupling analysis on the junction, it is found that a large tensile stress along axial and tangential directions occurs on outer surface of the heat transfer tube and is located at the region of 2.0 mm to 6.0 mm high over the tube-sheet. The stress attributes to the rapidly change of temperature on the tube outer surface and leads to the stress corrosions.

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Orović, Josip, Vedran Mrzljak, and Igor Poljak. "Efficiency and Losses Analysis of Steam Air Heater from Marine Steam Propulsion Plant." Energies 11, no. 11 (November 2, 2018): 3019. http://dx.doi.org/10.3390/en11113019.

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Air heaters are commonly used devices in steam power plants. In base-loaded conventional power plants, air heaters usually use flue gases for air heating. In this paper, the air heater from a marine steam propulsion plant is analyzed, using superheated steam as a heating medium. In a marine propulsion plant, flue gases from steam generator are not hot enough for the air heating process. In a wide range of steam system loads, the analyzed steam air heater has low energy power losses and high energy efficiencies, ranging from 98.41% to 99.90%. Exergy analysis of the steam air heater showed that exergy destruction is quite high, whereas exergy efficiency ranged between 46.34% and 67.14%. Air heater exergy destruction was the highest, whereas exergy efficiency was the lowest at the highest steam system loads, which was an unexpected occurrence because the highest loads can be expected in the majority of marine steam plant operations. The change in the ambient temperature significantly influences steam air heater exergy efficiency. An increase in the ambient temperature of 10 °C reduces analyzed air heater exergy efficiency by 4.5%, or more, on average.

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Dibyo, Sukmanto, and Ign Djoko Irianto. "DESIGN ANALYSIS ON OPERATING PARAMETER OF OUTLET TEMPERATURE AND VOID FRACTION IN RDE STEAM GENERATOR." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 19, no. 1 (March 10, 2017): 33. http://dx.doi.org/10.17146/tdm.2017.19.1.3228.

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HTGR is one of the next generation reactor types. HTGR is currently considered as one of the leading reactors for the future nuclear power plant. The steam generator is one of the main components in HTGR as well as in RDE. In the steam generator, the heat is transferred by high temperature helium gas in the shell side to water in the tube side to generate the superheated steam. the purpose of this work is to design the operating parameter of outlet temperature and void fraction of steam based on feed water mass flow rate and inlet temperature variations in RDE steam generator. In this work, the Chemcad program was used. Both inlet and outlet temperature of helium gas have been set up as boundary conditions. The result shows that using the mass flow rate of 4.3 kg/s - 4.8 kg/s and water inlet temperature of 110 oC - 160 oC, the superheated steam outlet temperature (void fraction = 1.0) is obtained in the range of 275.5 oC – 600 oC.This analysis is beneficial to assess 10 MW RDE design especially in the steam generator system operating parameters.Keywords: outlet temperature, void fraction, superheated steam, RDE steam generator ANALISIS DESAIN PARAMETER OPERASI UNTUK TEMPERATUR KELUARAN DAN FRAKSI UAP PADA PEMBANGKIT UAP RDE. Reaktor daya HTGR adalah salah satu tipe reaktor generasi lanjut. HTGR saat ini merupakan desain reaktor yang dipertimbangkan untuk pembangkit listrik unggulan dimasa mendatang. Pembangkit uap merupakan salah satu komponen utama pada HTGR begitu pula pada RDE. Di dalam pembangkit uap, panas dari gas helium temperatur tinggi pada sisi shell di transfer ke air pada sisi tube pembangkit uap untuk menghasilkan uap lewat jenuh. Tujuan analisis ini adalah mendesain parameter operasi terhadap temperatur keluaran dan fraksi uap berdasarkan variasi laju alir massa air umpan dan temperatur masuk pada RDE. Dalam analisis digunakan program Chemcad, temperatur gas helium masuk dan keluar ditentukan sebagai kondisi batas. Hasil menunjukkan bahwa dengan menggunakan laju alir massa 4,3 kg/detik - 4,8 kg/detik dan temperatur masukan air umpan dari 110 oC -160 oC dapat diperoleh uap lewat jenuh (fraksi uap= 1,0) pada temperatur keluaran dalam rentang 275,5 oC - 600 oC. Analisis ini berguna untuk memberikan kajian desain RDE 10 MW khususnya parameter operasi sistem pembangkit uap.Kata-kata kunci: temperatur keluaran, fraksi uap, uap lewat jenuh, pembangkit uap RDE

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Nelson, N. Rino. "Behaviour of flange joints in Steam Generator under thermal loads." E3S Web of Conferences 309 (2021): 01082. http://dx.doi.org/10.1051/e3sconf/202130901082.

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Pressure vessels such as steam generators are subjected to high temperature, in addition to high pressure during the operating condition. Flanges and bolts are made up of different materials whose coefficient of thermal expansion varies. Usually, thermal expansion in bolts is greater than that of flanges. At elevated temperatures bolts expand more than that of flanges, resulting in decrease of compression in connected members achieved during assembly stage, which in turn decreases the contact stress in gasket. This can lead to leakage of internal fluid. The loss in gasket contact stress due to differential thermal expansion can be nullified by using sleeves of higher thermal expansion between the flange-nut and flange-bolt head interfaces. At higher temperatures sleeves expand more than bolts and flanges, pushing the flanges closer towards each other, thus decreasing gap created due to differential thermal expansion. The behaviour of gasketed blind flange joint with and without sleeves is analysed and the performances are compared under thermal loads. The non-linear behaviour of gaskets is included by specifying the loading and unloading characteristics with hysteresis.

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Silvestri, G. J., R. L. Bannister, T. Fujikawa, and A. Hizume. "Optimization of Advanced Steam Condition Power Plants." Journal of Engineering for Gas Turbines and Power 114, no. 4 (October 1, 1992): 612–20. http://dx.doi.org/10.1115/1.2906634.

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The modern pulverized-coal power plant is the product of continuous design experience and component improvement in the 20th century. In recent years, studies of the effect of high temperatures on turbine materials have led to major worldwide research and development programs on improving the thermal cycle by raising turbine-inlet pressure and temperature. This paper reviews the importance of various parameters in trying to optimize a turbine cycle designed for advanced steam conditions. Combinations of throttle pressure (between 3500 psi [24.1 MPa] and 10,000 psi [70MPa]), throttle and reheat temperature(1000°F [538°C] to 1400°F [760°C]), and number of reheats are explored to establish a realistic turbine cycle design. Assessments and trade-offs are discussed, as applicable. Critical cycle components, feedwater cycle arrangements, and reheat pressure selections are analyzed in establishing an optimized steam turbine-boiler cycle for a 1000 MW turbine-generator. Applicability of results to smaller advanced steam turbines is given. A brief update on the high-temperature Wakamatsu turbine project in Japan is also given.

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Tan, Jibo, Xinqiang Wu, En-Hou Han, Wei Ke, Xiaoqiang Liu, Fanjiang Meng, and Xuelian Xu. "Corrosion fatigue behavior of Alloy 690 steam generator tube in borated and lithiated high temperature water." Corrosion Science 89 (December 2014): 203–13. http://dx.doi.org/10.1016/j.corsci.2014.08.027.

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Krivina, L. A., and Yu P. Tarasenko. "Anticorrosive protection of the steam generator of the heat exchanger by the high-temperature oxidation method." IOP Conference Series: Materials Science and Engineering 709 (January 3, 2020): 044012. http://dx.doi.org/10.1088/1757-899x/709/4/044012.

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Khoshhal, Abhas, Masoud Rahimi, Afshar Ghahramani, and Ammar Abdulaziz Alsairafi. "Computational fluid dynamics modeling of high temperature air combustion in an heat recovery steam generator boiler." Korean Journal of Chemical Engineering 28, no. 5 (April 12, 2011): 1181–87. http://dx.doi.org/10.1007/s11814-010-0481-3.

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JU, Huaiming, Yu YU, Zhiyong HUANG, Youje ZHANG, Zhiyong LIU, and Jun LI. "Experiment and Verification Test of the Once-through Steam Generator of the 10 MW High-temperature Gas-cooled Reactor Flow Stability of the Once-through Steam Generator." Journal of Nuclear Science and Technology 41, no. 4 (April 2004): 524–28. http://dx.doi.org/10.1080/18811248.2004.9715515.

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38

Yunus, M., A. A. Budiman, S. Zhe, Kiswanta, W. Chunlin, M. Subekti, S. Bakhri, and S. Jun. "Simulation System for PeLUIt 150 MW Nuclear Reactor by using vPower." Journal of Physics: Conference Series 2048, no. 1 (October 1, 2021): 012034. http://dx.doi.org/10.1088/1742-6596/2048/1/012034.

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Abstract In developing the PeLUIt 150 MW nuclear power plant based on the High Temperature Gas-cooled Reactor (HTGR) technology, with the helium-coolant and output thermal power of 150 MW, the PeLUIt simulator is also developed for training the operators and educating other technical personnel. Referred to the balance of plant (BOP) design of the PeLUIt, the simulator utilized the vPower simulation platform to simulate the secondary loop for power generation with a water-steam Rankine cycle. The paper focuses on developing the secondary loop’s main components: steam generator, steam turbine, condenser, deaerator, and feedwater pump. The reactor module in the primary loop is simplified as a heat source with 150 MW output. The steam generator that connects the primary and secondary loops is modeled with the heat exchanger module by transferring heat from helium to water/steam. Meanwhile, pressure and flow parameters can also be simulated for both helium and water/steam flows in steady-state and transient operating conditions. The steady-state simulation results are almost the same as the design data. The differences in the main steam temperature, feedwater pressure, and feedwater temperature, are 0.03%, 0.53%, and 0.02%, respectively. Meanwhile, the transient condition carried out in the loss of coolant accident showed a decrease in flowrate of 43.31 kg/s and an increase in temperature of feedwater and main-steam of 52.32 and 15.38 °C, respectively. In addition, there was a pressure drop of around 10.37 (feedwater) and 10.16 MPa (main-steam).

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Feng, Hong Cui, Wei Zhong, Yan Ling Wu, and Shui Guang Tong. "The Effects of Parameters on HRSG Thermodynamic Performance." Advanced Materials Research 774-776 (September 2013): 383–92. http://dx.doi.org/10.4028/www.scientific.net/amr.774-776.383.

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Changes of inlet temperature, mass flow rate and composition of flue gas, or of water/steam pressure and temperature in heat recovery steam generator (HRSG), all will modify the amount of waste heat recovered from flue gas; this brings forward a desire for the optimization of the design of HRSG. For single pressure HRSGs with given structures and specified values of inlet temperature, mass flow rate and composition of flue gas, the steam mass flow rate and gas outlet temperature of the HRSG are analyzed as functions of several parameters. This analysis is based on the laws of thermodynamics, incorporated into the energy balance equations for the heat exchangers. Those parameters are superheated steam pressure and temperature, feedwater temperature and pinch point temperature difference. It was shown that the gas outlet temperature could be lowered by selecting appropriate water/steam parameters and pinch point temperature difference. While operating with the suggested parameters, the HRSG can generate more high-quality steam, a fact of great significance for waste heat recovery from wider ranges of sources for better energy conservation.

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Grabezhnaya, V., A. Mikheyev, A. Alekhin, A. Kryukov, and A. Tikhomirov. "EXPERIMENTAL JUSTIFICATION OF DESIGN CHARACTERISTICS OF STEAM GENERATOR RP BREST-OD-300." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2021, no. 2 (June 26, 2021): 218–35. http://dx.doi.org/10.55176/2414-1038-2021-2-218-235.

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The project BREST-OD-300 reactor plant (RP) with a fast neutron reactor and a lead coolant in the primary circuit is being developed in NIKIET JSC. As a steam generator (SG), a helical-type steam generator with coiled tubes with subcritical pressure water in the second circuit is considered. To substantiate the design characteristics of the secondary coolant at the State Research Center of the Russian Federation - IPPE, thermohydraulic tests of various SG models were carried out at the SPRUT stand Initially, tests were carried out on a model of a coiled steam generator consisting of two three-tube modules with a longitudinal lead flow around a three-tube bundle of coiled tubes. The influence of operating parameters on thermohydraulic characteristics and hydrodynamic stability is shown in the case of operation of one module, as well as in the joint operation of two models in the investigated range of operating parameters. At the second stage, tests of a standard steam generator model were carried out with lead flowing around 18 heat exchange tubes. In the multitube model, the downward movement of the heating coolant took place with the flow around the bundle of heat transfer tubes close to the transverse flow. Data were obtained on the hydrodynamic stability of steam generating tubes and the entire model as a whole when operating in the entire range of changes in operating parameters, which are necessary for creating a databank and further verification of calculation codes describing the ongoing thermohydraulic processes. During the tests in both models of the steam generator, there were no noises inherent in unstable operating modes of the circuit. No pulsations of water and steam temperature were found, respectively, in the inlet and outlet collectors. At high lead temperatures, the temperature of the superheated steam was always close to the lead inlet temperature. A series of works devoted to the study of heat transfer from the side of a lead coolant with a transverse flow around a package of heat exchange tubes in normal heat transfer modes and with freezing of lead has been completed. Studies have been carried out on the effect of oxygen concentration in lead on heat transfer.

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Aminov, R. Z., and A. N. Egorov. "HYDROGEN-OXYGEN STEAM GENERATOR FOR A CLOSED HYDROGEN COMBUSTION CYCLE." Alternative Energy and Ecology (ISJAEE), no. 13-15 (August 11, 2018): 68–79. http://dx.doi.org/10.15518/isjaee.2018.13-15.068-079.

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The paper analyzes the problems of combustion hydrogen in an oxygen medium for produce high-temperature steam that can be used to produce electricity at various power plants. For example, at the nuclear power plants, the use of a H2-O2 steam generator as part of a hydrogen energy complex makes it possible to increase its power and efficiency in the operational mode due to steam-hydrogen overheating of the main working fluid of a steam-turbine plant. In addition, the use of the hydrogen energy complex makes it possible to adapt the nuclear power plants to variable electric load schedules in conditions of increasing the share of nuclear power plants and to develop environmentally friendly technologies for the production of electricity. The paper considers a new solution of the problem of effective and safe use of hydrogen energy at NPPs with a hydrogen energy complex.Technical solutions for the combustion of hydrogen in the oxygen medium using direct injection of cooling water or steam in the combustion products have a significant drawback – the effect of “quenching” when injecting water or water vapor which leads to a decrease in the efficiency of recombination during cooling of combustion products that is expressed in an increase fraction of non-condensable gases. In this case, the supply of such a mixture to the steam cycle is unsafe, because this can lead to a dangerous increase in the concentration of unburned hydrogen in the flowing part of the steam turbine plant. In order to solve this problem, the authors have proposed a closed hydrogen cycle and a hydrogen vapor overheating system based on it, and carried out a study of a closed hydrogen combustion system which completely eliminates hydrogen from entering the working fluid of the steam cycle and ensures its complete oxidation due to some excess of circulating oxygen.The paper considers two types of hydrogen-oxygen combustion chambers for the system of safe generating of superheated steam using hydrogen in nuclear power plant cycle by using a closed system for burning hydrogen in an oxygen medium. As a result of mathematical modeling of combustion processes and heat and mass transfer, we have determined the required parameters of a hydrogen-oxygen steam generator taking into account the temperature regime of its operation, and a power range of hydrogen-oxygen steam generators with the proposed combustion chamber design.

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Iliev, Iliya, Angel Terziev, and Hristo Beloev. "Condensing economizers for large scale steam boilers." E3S Web of Conferences 180 (2020): 01004. http://dx.doi.org/10.1051/e3sconf/202018001004.

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An energy analysis and assessment of the feasibility of the development and implementation of a condensing economizer (CE) for steam generator TGM-96A has been made. The condensing economizer is characterized with a very high heating capacity of 21 to 23.5 Gcal/h and will cool the fuel gases from 123°C down to 47°C. The utilized heat will be used to heat three water loops: two loops with District Heating Network water (DHN) and one with raw water from the Dnieper River. The expected improvement in the efficiency of the steam generator is 10.55%. Complete thermal calculations of the condensing economizer have been carried out along with a definition by flow and temperature of the heat carriers. Based on the calculations of the heat and mass transfer processes, the thermal capacity of the economizer under the conditions of partial condensation of water vapor from the flue gases was estimated. The analysis presents an objective assessment of the project’s investments, as well as other financial indicators and environmental benefits that give the investor the opportunity to put the project into operation.

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Liu, Yun Jing, and Yu Long Wu. "Adaptive Fuzzy PID Control Applied in Boiler System of Power Plant." Applied Mechanics and Materials 599-601 (August 2014): 794–97. http://dx.doi.org/10.4028/www.scientific.net/amm.599-601.794.

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The boiler system in power plant consists mainly of a steam-water system and combustion system, which produce a high-pressure superheated steam to drive a generator in order to produce power. The steam temperature control of the boiler system is a critical operational consideration. Parameters of boiler producing process must be controlled strictly. But the boiler system has high nonlinearity, large delay, strong coupling and load disturbance. Therefore it is difficult to develop practical mathematical model of the control object using the traditional PID control. In order to realize the optimal control of steam temperature, this paper presents the adaptive fuzzy-PID control Strategy. Simulation results based on MATLAB show that the proposed strategy can largely improve the system response performance compared with the traditional PID control

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Jeong, Sung Hoon, and Young Ze Lee. "Wear Characteristics of Tube-Support Components for a Nuclear Steam Generator under Fretting Conditions." Solid State Phenomena 120 (February 2007): 181–86. http://dx.doi.org/10.4028/www.scientific.net/ssp.120.181.

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Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature flows in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration) which cause fretting wear in contact part of tube-support. The reduction of tube thickness due to fretting wear of tube-support can threaten the safety of nuclear power plant. Therefore, a research on the fretting wear characteristics of tube-support is required. This work is focused on investigations of fretting wear characteristics and wear mechanisms of tube-support. Results show that the wear rate of tube is proportional to that of support and that with increasing the water temperature the wear volume of tube-support decreases because the oxidation rate decreases due to reduction of the oxygen concentration in contact surfaces. Also, the wear mechanisms of tube-support are abrasive and oxidational wear.

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45

Liu, Yang, Xiao Yuan, and Yu Bao. "Experimental Evaluation Study on High Temperature Tolerance Foam for Profile Control of Steam Huff-n-Puff." Geofluids 2022 (March 8, 2022): 1–10. http://dx.doi.org/10.1155/2022/1433998.

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In the process of steam huff and puff to develop heavy oil reservoirs, steam channeling is the main factor that restricts the development effect, which causes that the swept area of steam is limited and unable to heat heavy oil adequately. It is important to find an economical and efficient method to prevent steam channeling from happening at steam huff and puff production; thus, a DF-2 high temperature tolerance foam system is proposed to assist steam huff and puff production in this paper. Firstly, utilizing high-temperature aging tank and high-speed stirrer device, the foam quality of DF-2 foam and four kinds of other common industrial foams were evaluated and compared; this experiment result shows that the DF-2 foam has the best temperature tolerance. Whereafter, three groups of single-sandpack experiments were carried out to investigate the change of DF-2 foam resistance factor under different conditions, which included the influence of gas liquid ratio, injection pattern, and crude oil, and the blocking capabilities of DF-2 foam were studied. Through the dual-sandpack experiment, the profile control performance and diversion capacity of the DF-2 foaming agents were studied and evaluated systematically. Finally, the DF-2 high temperature tolerance foam agent was applied to on-site production. The sandpack experiment results indicated that the resistance factor of DF-2 foam agent decreased with gas-liquid ratio, and the optimal gas-liquid ratio was 1 : 1. Coinjection of gas and liquid with foam generator is better than slug injection. Crude oil in sandpack can reduce blocking capability to some extent. During the application of the DF-2 high temperature tolerance foaming agent in the steam huff-n-puff process in M-12 Oilfield block, it had a good performance of profile control and foam blocking, solving the problem of steam channeling and improving the steam sweep area to increase the steam-oil ratio in the heavy oil reservoir, and the production performance of steam huff and puff was improved to a great extent.

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46

Fic, Adam, Jan Składzień, and Michał Gabriel. "Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle." Archives of Thermodynamics 36, no. 1 (March 1, 2015): 3–18. http://dx.doi.org/10.1515/aoter-2015-0001.

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Abstract Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

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Kim, Young Jin, Byung Jin Lee, Kunwoo Yi, Yoon Jae Choe, and Min Chul Lee. "Numerical Study on the Effects of Relative Diameters on the Performance of Small Modular Reactors Driven by Natural Circulation." Energies 13, no. 22 (November 11, 2020): 5881. http://dx.doi.org/10.3390/en13225881.

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Most of the small modular reactors (SMRs) under development worldwide present the same components: an integral reactor vessel with a low-positioned core as the heat source and a high-positioned steam generator as the heat sink. Moreover, some SMRs are being designed to be driven by natural circulation during normal power generation. This work focused on such designs and on their performance, considering the changes generated by the geometric and hydraulic parameters of the system. Numerical simulations using mass, momentum, and energy equations that considered buoyancy forces were performed to determine the effects of various geometric and hydraulic parameters, such as diameters and flow resistances, on the reactor’s performance. It was found that nonuniform diameters promote velocity changes that affect the natural circulation flow rate. Moreover, the reactor’s temperature distribution depends on the steam generator tube pitch. Therefore, the hydraulic diameters of the reactor’s coolant passages should be maintained as uniform as possible to obtain a more uniform temperature distribution and a larger mass flow rate in SMRs.

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Xia, Da-Hai, and Jing-Li Luo. "Effects of reduced sulfur on passive film properties of steam generator (SG) tubing: an overview." Anti-Corrosion Methods and Materials 66, no. 3 (May 7, 2019): 317–26. http://dx.doi.org/10.1108/acmm-09-2018-1996.

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Purpose Corrosion is considered as one of the issues that threaten the safe operation of steam generator (SG) tubing. Some sulfur-related specie can cause corrosion degradation of SG tubing. Sulfur-induced corrosion of SG alloys in high temperature and high-pressure water is one of the most complicated processes. The purpose of this study is to study the effect of reduced sulphur on passive film properties of steam generator (SG) tubing. Design/methodology/approach In this paper, the effects of reduced sulfur on passive film properties of SG tubing were reviewed from the aspects of thermodynamic calculations and experimental. Findings Thermodynamic calculations are mainly presented by E-pH diagrams, volt equivalent diagrams and species distribution curves. The stability of sulphur species highly depends on temperature, solution pH, and electrochemical potential. Experimental data indicated that reduced sulfur species can interact with the passive film, which led to changes in film thickness, film structure, semiconductivity and pitting growth rate. Originality/value The state-of-the-art discussed in this paper gives basis for resolving engineering problems regarding with sulfur-induced corrosion.

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Park, Chi Yong, Yong Sung Lee, and Myung Hwan Boo. "Development of Wear Test System for Steam Generator Tubes in Nuclear Power Plants." Key Engineering Materials 297-300 (November 2005): 1418–23. http://dx.doi.org/10.4028/www.scientific.net/kem.297-300.1418.

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In steam generators of nuclear power plants, flow-induced vibration (FIV) can lead to tube damage by fretting-wear occurred due to impact and sliding movement between the tubes and their supports. There have been many studies and test results on wear damage of steam generator tubes but they were not reflected the mechanical and chemical conditions accurately. KEPRI nuclear power laboratory developed a wear test system, which is able to control the motion of impact and sliding simultaneously in the pressurized high temperature water-chemistry conditions. Some wear tests were performed to verify the stable operation for the wear test. This wear test system with new concepts was described briefly, and some data for verifying its performance have been shown in the cases of the selected some test results. In the test, Alloy 690 was used for tube materials and 409 stainless steel for support plates. A little data deviation was obtained and stability of system operation was investigated.

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Susyadi, Susyadi, Hendro Tjahjono, Sukmanto Dibyo, and Jupiter Sitorus Pane. "STUDI KARAKTERISTIK PEMBENTUKAN UAP DALAM PEMBANGKIT UAP HELIKAL PADA REAKTOR MODULAR DAYA KECIL." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 17, no. 2 (June 6, 2015): 59. http://dx.doi.org/10.17146/tdm.2015.17.2.2276.

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Reaktor modular daya kecil (SMR) sangat cocok untuk dibangun Indonesia, terutama pada lokasi-lokasi dengan kapasitas jaringan listrik yang rendah sehingga investigasi lebih jauh tentang reaktor ini sangat diperlukan. Umumnya SMR memiliki bentuk pembangkit uap yang kompak dan terintegrasi di dalam bejana tekan. Disain tersebut menyebabkan perbedaan pendekatan dalam memproduksi uap dibandingkan reaktor nuklir konvensional yang menggunakan pembangkit uap tabung-u terbalik. Oleh karena itu tujuan dari penelitian ini adalah untuk mengetahui karakteristik uap dan pola pembentukkannya di dalam pembangkit uap tipe helikal yang banyak digunakan oleh SMR. Metoda yang dipakai adalah dengan melakukan pemodelan dan perhitungan numerik menggunakan program RELAP5. Dalam pemodelan, aliran air umpan bertekanan dan temperatur rendah dimasukkan ke dalam tabung helikal sementara aliran fluida bertekanan dan temperatur tinggi, yang mewakili pendingin sistem primer reaktor, berada di sisi luar tabung. Hasil perhitungan menunjukkan bahwa uap yang dihasilkan oleh pembangkit uap helikal bersifat lewat jenuh yakni sekitar 25 K di atas titik jenuhnya. Hal ini memberikan keunggulan komparatif dari segi disain dan operasional pada SMR dibanding reaktor konvensional karena uap lewat jenuh yang dihasilkan dapat mengurangi kerugian turbin dan sekaligus meningkatkan efisiensi termodinamika. Kata kunci: pembangkit uap helikal, SMR, PWR, uap lewat jenuh, RELAP5 Small modular reactor (SMR) is very suitable to be deployed in Indonesia especially for locations having low electrical grid capacity, so further investigation on the characteritic of this reactor is needed. In general SMR has a compact and integrated-to-vessel steam generator design. This design implies different approach in producing steam as compared to conventional nuclear power plant having inverted u-tube steam generator. For that reason, this research is intended to investigate the steam characteristic and how it is generated in the helical SG which is widely used in SMR. The method used is through numerical calculation of the SG model using RELAP5 code. In the model, the feed-water which has low pressure and temperature is flown into helical tubes while high pressure and temperatur fluid, which represents reactor primary system coolant, stays in outer side of the tube. Calculation result shows that the steam produced by helical steam generator is superheated, i.e. about 25 K above saturation temperature. This provides comparative advantage to SMR on the design and operational aspects compared to conventional reactors because the superheated steam it produces can reduce turbine losses and at the same time increase thermodynamic efficiency. Keywords: helical steam generator, SMR, PWR, superheated steam, RELAP5

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Journal articles on the topic 'High-temperature steam generator'

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