Relevant bibliographies by topics / Pressurized water reactors Steam-turbines

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

Academic literature on the topic 'Pressurized water reactors Steam-turbines'

Author: Grafiati
Published: 4 June 2021
Last updated: 1 February 2022

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Journal articles on the topic "Pressurized water reactors Steam-turbines"

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Hutchinson, Harry. "Powering Down." Mechanical Engineering 125, no. 04 (April 1, 2003): 46–48. http://dx.doi.org/10.1115/1.2003-apr-4.

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This article reviews that after a half century of safety testing for the nuclear industry, a key heat-transfer lab is losing its home. Columbia University’s Heat Transfer Research Facility has been the only place to go for key safety testing. Since the days of the Atoms for Peace program during the Eisenhower years, the lab has tested generations of nuclear reactor fuel assemblies. The lab’s clients over the years have included all the designers of pressurized water reactors in the United States and others from much of the world. The tests are primarily concerned with one small, but significant feature of a reactor core. A core contains as many as 3000 fuel assemblies, bundles of long, slender rods containing enriched uranium. Controlled fission among the bundles heats water to begin the series of heat-transfer cycles that send steam to the turbines that will drive generators.
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Feliachi, Ali, and Lotfi A. Belblidia. "Optimal Level Controller for Steam Generators in Pressurized Water Reactors." IEEE Power Engineering Review PER-7, no. 6 (June 1987): 30–31. http://dx.doi.org/10.1109/mper.1987.5527113.

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Feliachi, A., and L. A. Belblidia. "Suboptimal level controller for steam generators in pressurized water reactors." IEEE Transactions on Energy Conversion 3, no. 2 (June 1988): 278–84. http://dx.doi.org/10.1109/60.4731.

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Feliachi, Ali, and Lotfi A. Belblidia. "Optimal Level Controller for Steam Generators in Pressurized Water Reactors." IEEE Transactions on Energy Conversion EC-2, no. 2 (June 1987): 161–67. http://dx.doi.org/10.1109/tec.1987.4765824.

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Roach, William H. "Detection of steam generator tube leaks in pressurized water reactors." Nuclear Engineering and Design 89, no. 1 (November 1985): 81–89. http://dx.doi.org/10.1016/0029-5493(85)90144-x.

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Györke, Gábor, and Attila R. Imre. "Physical-chemical Background of the Potential Phase Transitions during Loss of Coolant Accidents in the Supercritical Water Loops of Various Generation IV Nuclear Reactor Types." Periodica Polytechnica Chemical Engineering 63, no. 2 (March 18, 2019): 333–39. http://dx.doi.org/10.3311/ppch.12770.

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Loss of coolant accidents (LOCA) are a serious type of accidents for nuclear reactors, when the integrity of the liquid-loop breaks. While in traditional pressurized water reactors, pressure drop can cause flash boiling, in Supercritical-Water Cooled reactors, the pressure drop can be terminated by processes with fast phase transition (flash boiling or steam collapse) causing pressure surge or the expansion can go smoothly to the dry steam region. Modelling the pressure drop of big and small LOCAs as isentropic and isenthalpic processes and replacing the existing reactor designs with a simplified supercritical loop, limiting temperatures for various outcomes will be given for 24.5 and 25 MPa initial pressure. Using the proposed method, similar accidents for chemical reactors and other equipment using supercritical fluids can be also analyzed, using only physical-chemical properties of the given supercritical fluid.
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Kuan, Cheng Chung, Chaung Lin, and Chang Chia Hsu. "Fuzzy Logic Control of Steam Generator Water Level in Pressurized Water Reactors." Nuclear Technology 100, no. 1 (October 1992): 125–34. http://dx.doi.org/10.13182/nt92-a34758.

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Tsai, Chiung Wen, Zhi Hao Ren, Jia Lei Ruan, Ting Wang, and Jing Gang Li. "Analysis of Main Steam Line Break for a Pressurized Water Reactor." Applied Mechanics and Materials 764-765 (May 2015): 181–85. http://dx.doi.org/10.4028/www.scientific.net/amm.764-765.181.

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The thermal hydraulics (T-H) code, GINKGO, has been developed by China Nuclear Power Technology Research Institute (CNPTRI) for the transient analyses of pressurized water reactors. GINKGO is designed to simulate the non-loss-of-coolant-accidents (non-LOCAs), and the transients caused by the breaks in secondary side. This paper presents the GINKGO models as well as the analysis of double-ended MSLB that identifies the reactor core characteristics under reactivity feedback. The analysis results show that the variations of reactivity and nuclear power are governed by the competition between positive moderator feedback and negative reactivity insertion by borated water. Finally, the boron concentration is high enough to suppress the nuclear power to a lower power level of ~5%, terminating this reactivity insertion accident (RIA).
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Perovic, V., A. Perovic, G. C. Weatherly, and A. M. Brennenstuhl. "Microstructure and Microchemistry of Inconel 600 STEAM Generator Tubing." Microscopy and Microanalysis 6, S2 (August 2000): 356–57. http://dx.doi.org/10.1017/s1431927600034279.

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Inconel 600 is an austenitic Ni-Cr-Fe alloy which is extensively used for tubing in steam generators of pressurized light water reactors (PWR) and CANDU heavy water reactors, because of its excellent mechanical properties and corrosion resistance. However, there have been instances of intergranular stress corrosion cracking of tubes in operating steam generators. The chemistry and the structure of grain boundaries and grain boundary precipitation have emerged as factors of prime importance in understanding stress corrosion cracking and intergranular attack of nickel-base alloys (see e.g. ref. l).In this study analytical electron microscopy was used to determine the microstructure of grain boundary and matrix precipitates, grain boundary chromium content and dislocation substructure of selected steam generating tubes of CANDU reactors. The results of the in-service materials are compared with as-received material. Two JEOL 2010 STEM instruments were used in this study.
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Yang, Guangze, Véronique Pointeau, Etienne Tevissen, and Alexandre Chagnes. "A review on clogging of recirculating steam generators in Pressurized-Water Reactors." Progress in Nuclear Energy 97 (May 2017): 182–96. http://dx.doi.org/10.1016/j.pnucene.2017.01.010.

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Dissertations / Theses on the topic "Pressurized water reactors Steam-turbines"

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Elliott, Alicia M. "Multiphysics modeling of activity transport and evolution of CRUD and steam generator oxides in pressurized water reactors." Thesis, Massachusetts Institute of Technology, 2018.

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This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 121-123).
Fouling deposits of corrosion products on fuel cladding, known as crud, in the core of light water reactors can cause a variety of operational issues. Buildup of radioactive crud and corrosion products on ex-core structures, such as steam generators and piping, can cause increased radiation fields and higher dose exposures for plant workers. To better understand the mechanisms of corrosion product activity transport and evolution in the primary coolant loop, a crud source term and activity transport code that can predict the concentration of active isotopes in a primary loop over time and plant operating parameters was developed, implementing mechanistic models for soluble corrosion product dissolution and precipitation. The code described in this thesis tracks activated isotope deposition throughout the primary loop with spatial and temporal resolution, without the use of empirical rate constants derived from plant measurements, to predict primary loop activity buildup. Developed in C++ using the MOOSE Framework, this code can be easily coupled to other multiphysics codes through the MOOSE MultiApp system. A set of input file generation scripts, written in Python, were developed to calculate thermodynamic parameters for chemical reactions added to the simulation, and easily set up simulation input files in a "user-friendly" format. The open source code described in this work, Ouroboros, is available freely for future improvements and adaptations to implement additional mechanisms and more rigorous models. This code is the first step towards a long term effort to develop an open source, fully mechanistic crud source term model including all mechanisms for activity transport in pressurized water reactors.
by Alicia M. Elliott.
S.M.
S.M. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering
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Nakka, Ravi Kumar Nasrazadani Seifollah. "Flow accelerated corrosion experience at Comanche Peak Steam Electric Station." [Denton, Tex.] : University of North Texas, 2008. http://digital.library.unt.edu/permalink/meta-dc-6072.

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Namduri, Haritha Nasrazadani Seifollah. "Formation and quantification of corrosion deposits in the power industry." [Denton, Tex.] : University of North Texas, 2007. http://digital.library.unt.edu/permalink/meta-dc-3635.

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Nakka, Ravi Kumar. "Flow Accelerated Corrosion Experience at Comanche Peak Steam Electric Station." Thesis, University of North Texas, 2008. https://digital.library.unt.edu/ark:/67531/metadc6072/.

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Flow accelerated corrosion (FAC) is a major concern in the power industry as it causes thinning of the pipes by the dissolution of the passive oxide layer formed on the pipe surface. Present research deals with comparing the protection offered by the magnetite (Fe3O4) versus maghemite (γ-Fe2O3) phases thickness loss measurements. Fourier transform infrared spectroscopy (FTIR) is used in distinguishing these two elusive phases of iron oxides. Representative pipes are collected from high pressure steam extraction line of the secondary cycle of unit 2 of Comanche Peak Steam Electric Station (CPSES). Environmental scanning electron microscopy (ESEM) is used for morphological analysis. FTIR and X-ray diffraction (XRD) are used for phase analysis. Morphological analysis showed the presence of porous oxide surfaces with octahedral crystals, scallops and "chimney" like vents. FTIR revealed the predominance of maghemite at the most of the pipe sections. Results of thickness measurements indicate severe thickness loss at the bend areas (extrados) of the pipes.
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Namduri, Haritha. "Characterization of Iron Oxide Deposits Formed at Comanche Peak Steam Electric Station." Thesis, University of North Texas, 2003. https://digital.library.unt.edu/ark:/67531/metadc5521/.

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The presence of deposits leading to corrosion of the steam generator (SG) systems is a major contributor to operation and maintenance cost of pressurized water reactor (PWR) plants. Formation and transport of corrosion products formed due to the presence of impurities, metallic oxides and cations in the secondary side of the SG units result in formation of deposits. This research deals with the characterization of deposit samples collected from the two SG units (unit 1 and unit 2) at Comanche Peak Steam Electric Station (CPSES). X-ray diffraction (XRD), Fourier transform infrared spectrophotometry (FTIR), scanning electron microscopy (SEM), and energy dispersive spectroscopy (EDS) techniques have been used for studying the compositional and structural properties of iron oxides formed in the secondary side of unit 1 and unit 2. Magnetite (Fe3O4) was found to be predominant in samples from unit 1 and maghemite (g-Fe2O3) was found to be the dominant phase in case of unit 2. An attempt has been made to customize FTIR technique for analyzing different iron oxide phases present in the deposits of PWR-SG systems.
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Namduri, Haritha. "Formation and Quantification of Corrosion Deposits in the Power Industry." Thesis, University of North Texas, 2007. https://digital.library.unt.edu/ark:/67531/metadc3635/.

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The presence of deposits on the secondary side of pressurized water reactor (PWR) steam generator systems is one of the main contributors to the high maintenance costs of these generators. Formation and transport of corrosion products formed due to the presence of impurities, metals and metallic oxides in the secondary side of the steam generator units result in formation of deposits. This research deals with understanding the deposit formation and characterization of deposits by studying the samples collected from different units in secondary side system at Comanche Peak Steam Electric Station (CPSES). Fourier transform infrared spectrophotometry (FTIR), scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS) have been used for studying the phases, morphologies and compositions of the iron oxides formed at Unit 1 and Unit 2 of secondary side of steamgenerator systems. Hematite and magnetite were found to be the dominant phases of iron oxides present in the units. Fe, Cr, O, Ni, Si, Cl and Cu were found in samples collected from both the units. A qualitative method was developed to differentiate iron oxides using laser induced breakdown spectroscopy (LIBS) based on temporal response of iron oxides to a high power laser beam. A quantitative FTIR technique was developed to identify and quantify iron oxides present in the different components of the secondary side of the steam generator of CPSES. Amines are used in water treatment to control corrosion and fouling in pressurized water reactors. CPSES presently uses an amine combination of dimethylamine (DMA), hydrazine and morpholine to control the water chemistry. Along with the abovementioned amines, this study also focuses on corrosion inhibition mechanismsof a new amine DBU (1, 8-diazabicyclo [5.4.0] undec-7-ene). Electrochemical impedance spectroscopy and polarization curves were used to study the interaction mechanism between DBU solution and inconel alloys 600 and 690 at steamgenerator operating temperatures and pressures. Of all the amines used in this study (DMA, DBU, ETA, and morpholine), DMA was more effective at keeping the passive film formed on the alloy 600 surface from failing at both ambient and high temperatures. Morpholine was found result in higher corrosion resistance compared to the other amines in case of alloy 690.
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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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Choi, Jung In. "Nonlinear digital computer control for the steam generator system in a pressurized water reactor plant." Thesis, Massachusetts Institute of Technology, 1987. http://hdl.handle.net/1721.1/42103.

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Rebak, Raúl Basilio. "Environmentally induced cracking in alloy 600 and SA 302 steel in pressurized water reactor steam generator conditions /." The Ohio State University, 1993. http://rave.ohiolink.edu/etdc/view?acc_num=osu1487846885777343.

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Brink, Michael Joseph. "Hardware-in-the-loop simulation of pressurized water reactor steam-generator water-level control, designed for use within physically distributed testing environments." The Ohio State University, 2013. http://rave.ohiolink.edu/etdc/view?acc_num=osu1357273230.

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Books on the topic "Pressurized water reactors Steam-turbines"

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Kurtz, R. J. Steam generator group project: Annual report, 1985. Washington, DC: Division of Engineering Safety, Office of Nuclear Regulatory Researach, U.S. Nuclear Regulatory Commission, 1987.

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Zverkov, V. V. I͡A︡dernai͡a︡ paroproizvodi͡a︡shchai͡a︡ ustanovka s VVĖR-440. Moskva: Ėnergoatomizdat, 1987.

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Stubbe, E. J. Assessment study of RELAP-5 MOD-2 cycle 36.01: Based on the Doel-2 steam generator tube rupture incident of June 1979. Washington, D.C: U.S. Nuclear Regulatory Commission, 1986.

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Yidong, Zhou, and Huang Xingrong, eds. He dian chang he zheng qi gong ying xi tong: Nuclear steam supply system of nuclear power plants. Beijing Shi: Yuan zi neng chu ban she, 2010.

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Rogers, J. M. An analysis of semiscale Mod-2C S-FS-1 steam line break test using RELAP5/MOD2. Washington, DC: U.S. Nuclear Regulatory Commission, 1992.

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He dian chang zheng qi dong li zhuan huan xi tong: Steam power conversion system of nuclear power plants. Beijing Shi: Yuan zi neng chu ban she, 2010.

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Frank, L. Steam generator operating experience update for 1987-1988. Washington, D.C: Division of Engineering and Systems Technology, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1989.

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Odar, F. Assessment of the TRAC-M codes using Flecht-Seaset reflood and steam cooling data. Washington, DC: U.S. Nuclear Regulatory Commission, 2001.

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Lee, R. Y. Thermal-hydraulic research plan for Babcock and Wilcox plants. Washington, DC: Division of Reactor and Plant Systems, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1988.

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Lee, R. Y. Thermal-hydraulic research plan for Babcock and Wilcox plants. Washington, DC: Division of Reactor and Plant Systems, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1988.

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Book chapters on the topic "Pressurized water reactors Steam-turbines"

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Wan Mohamad, Wan Munirah, Tahir Ahmad, and Azmirul Ashaari. "Modeling Steam Generator System of Pressurized Water Reactor Using Fuzzy Arithmetic." In Communications in Computer and Information Science, 237–46. Singapore: Springer Singapore, 2016. http://dx.doi.org/10.1007/978-981-10-2777-2_21.

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Meier, Paul F. "Nuclear." In The Changing Energy Mix, 120–51. Oxford University Press, 2020. http://dx.doi.org/10.1093/oso/9780190098391.003.0005.

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With the exception of nuclear submarines and some military applications, nuclear energy is only used to generate electricity. In the United States, uranium and plutonium are the fuels of choice, while some other countries, notably India, are developing thorium as the nuclear fuel. There are two main types of nuclear reactors—the pressurized water reactor (PWR) and the boiling water reactor (BWR). The PWR is the more common design, where the water used to generate steam and drive the turbine is isolated from the reactor core. In contrast, the water that moderates reactor heat in the BWR is also used to generate the steam, so this water must be contained to prevent radioactive contamination. In the United States, nuclear energy accounts for about 20% of electricity generation. Worldwide uranium reserves are about 6 million tonnes based on a price of $130/kg, but if this price constraint is relaxed, the supply of uranium is virtually unlimited since it is present in seawater at parts per billion levels.
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Muscroft, J. "2. Modern large 3000 rev/min steam turbines for pressurized water reactor power stations." In Technology of turbine plant operating with wet steam, 15–24. Thomas Telford Publishing, 1989. http://dx.doi.org/10.1680/totpowws.13957.0003.

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"Corrosion Fatigue Cracking of a Steam Generator Vessel From a Pressurized Water Reactor." In Handbook of Case Histories in Failure Analysis, 136–39. ASM International, 1992. http://dx.doi.org/10.31399/asm.fach.v01.c9001051.

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Bittanti, S., R. Cori, F. Pretolani, L. Rassu, and D. Roncaglioni. "IDENTIFICATION OF A PRESSURIZED WATER REACTOR STEAM GENERATOR BY STATE-SPACE MULTI-VARIABLE MODELS." In Control Science and Technology for Development, 71–75. Elsevier, 1986. http://dx.doi.org/10.1016/b978-0-08-033473-8.50016-4.

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Conference papers on the topic "Pressurized water reactors Steam-turbines"

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Sun, Peiwei, and Chong Wang. "Coordinated Control of a Small Pressurized Water Reactor." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81156.

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Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.
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Jiang, Huijing, Ning Bai, Huanfen Zhan, Feng Shen, Bin Gao, and Xuee Wang. "Small Modular Pressurized Water Reactors Combined With Conventional Thermal Power Plant." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66690.

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Small modular reactor is investigated worldwide with the advantages of lower initial investment and short construction period. Generally, the economy of small modular pressurized water reactor (PWR) is not as good as large PWR, so various applications of small PWR are investigated, such as marine reactor, heat supply and sea water desalination. Limited to the parameters of steam generator, the generating efficiency for the pressurized water reactor nuclear power plant is about 33%, while the steam temperature of supercritical fossil power plant can exceed 600°C and generating efficiency is more than 45%. The essence of a hybrid power plant is to use a fossil fuel to superheat wet steam in an outer steam superheating device, after the steam generator to improve the parameters of working fluid. On one hand, the innovative hybrid nuclear power plant which combines nuclear reactor with conventional thermal energy can improve the efficiency of small PWR. On the other hand, this hybrid power plant has lower carbon emission compared with traditional thermal power plant. This paper describes two different coupling schemes of small pressurized water reactor combined with supercritical thermal power plant using steam turbine. Efficiency of hybrid power plant is influenced by the coupling scheme, steam parameter of the superheating device outlet, the proportion of nuclear energy, efficiency of assemblies and so on. The plant efficiency becomes higher with the improvement of parameter of the superheating device’s outlet steam, and it is higher when the proportion of nuclear energy becomes lower. Take the 660MWt integrated small pressurized water reactor as an example, when the proportion of nuclear energy accounts for 48%, the thermal efficiency of this innovative hybrid power plant is about 43%,while the net efficiency is 41%, that is much higher than the efficiency of traditional pressurized water reactor, improving about 24 percent. As to the carbon emission, it depends on the coal consumption rate of power supply. The coal consumption rate of this hybrid power plant is 158g /kWh, while the consumption rate of thermal hybrid power plant is 280g/kWh, reducing about 44 percent. Also, the fundamental solutions of technical problems for this innovative hybrid power plant are discussed in the paper. Furthermore, several useful outcomes and suggestions for key equipments are put forward, such as the scheme of a superheating device and high-temperature steam turbine, and the possibility of using a lava boiler as the superheating device to improve the steam parameters after the steam generator is analyzed.
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Cao, Huasong. "Simulation of Small Steam Generator Tube Break (SSGTB) in a Small Pressurized Water Reactor (SPWR)." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66637.

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Lots of efforts have been made to Research & Development of Small pressurized water reactors (SPWRs). Steam generator tube break occurs due to wear and corrosion frequently in the reactor. Among the breaks, Small Steam Generator Tube Break (SSGTB) is difficult to detect. Therefore, it is necessary to investigate the features of SSGTB. A small pressurized water reactor model has been established in this paper by Relap5. The model includes reactor core, pressurizer, steam generator, main coolant pump and auxiliary safety system. The core flow, pressure of pressurizer, core outlet temperature and secondary outlet steam temperature obtained based on steady-state calculation is compared with design data to verify the model correct. SSGTB is simulated by introducing a small break in the steam generator tube. The important parameters of reactor are recorded and analyzed. The procedure of SSGTB is analyzed and the system response features are summarized.
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Ponomaryov, Paul, Yifeng Zhou, Cristina-Maria Mazza, and Igor Pioro. "Modifications of Generic Pressurized-Water Reactor Units to Increase Cycle Efficiency." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60353.

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Currently, Pressurized Water Reactors (PWRs), Boiling Water Reactors (BWRs) and Pressurized Heavy Water Reactors (PHWRs) have the lowest thermal efficiencies compared to those of other nuclear-power reactors and thermal power plants. Therefore, the objective of this paper is to propose modifications to a generic PHWR to yield an increase in overall plant thermal efficiency. The focus of this paper is primarily on the secondary side of a Nuclear Power Plant (NPP) and is directly dealing with wet-steam High-Pressure (HP) and Low-Pressure (LP) turbine stages and a Moisture Separator and Reheater (MSR). Modifications of the HP and LP turbine stages are based on utilizing moisture removal stages (having up to 60% removal efficiency), which reduce the moisture content as the steam passes through those turbine stages. Reduced energy losses and an increase in mechanical efficiency due to lesser moisture content results in an increase in thermal efficiency. Furthermore, implementing moisture-removal stages in the LP turbine gives the ability to eliminate the reheater in the MSR, thus resulting in an increase of thermal efficiency due to both, the higher mechanical efficiency of an LP turbine and the redirection of live-steam previously used by the MSR to a HP turbine. To be able to show an increase in thermal efficiency based on these modifications of a generic PHWR, the Pickering CANDU-6 nuclear-reactor parameters were used as a reference case in the software, called DE-TOP. The modifications suggested in this paper can be applied to any NPP that uses a Rankine steam-turbine cycle on the secondary side (PWR, PHWR and/or BWR) and recommended for implementation during planned replacement of LP- and/or HP-turbine rotors for new construction of PWRs and PHWRs.
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Ma, Zhegang, Carlo Parisi, Cliff Davis, Sai Zhang, and Hongbin Zhang. "Risk-Informed Analysis for Accident Tolerant Fuel in Pressurized Water Reactors." In 2020 International Conference on Nuclear Engineering collocated with the ASME 2020 Power Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/icone2020-16761.

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Abstract This paper presents the research activities performed by Idaho National Laboratory (INL) for the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk-Informed System Analysis (RISA) Pathway, Enhanced Resilient Plant (ERP) Systems research, using the probabilistic risk assessment (PRA) tool SAPHIRE and the deterministic best estimate tool RELAP5-3D for risk-informed analysis. The ERP research supports DOE and industry initiatives by developing Accident Tolerant Fuel (ATF), the Diverse and Flexible Coping Strategy (FLEX), and passive cooling system designs to enhance existing reactors’ safety features (both active and passive) and to substantially reduce operating costs of nuclear power plants (NPPs) through risk-informed approaches to analyze the plant enhancements and their characterization. The risk-informed analysis used SAPHIRE and RELAP5-3D to evaluate the risk impacts from near-term ATF (FeCrAl and Chromium-coated clads) on a generic Westinghouse three-loop pressurized water reactor (PWR) under the following accident scenarios: station blackout (SBO), loss of feedwater (LOFW), steam generator tube rupture (SGTR), loss-of-coolant accidents (LOCAs), locked rotor transient, turbine trip transient, anticipated transient without scram (ATWS), and main steam line break (MSLB). The RELAP5-3D simulations included the time to core damage, time to 0.5 kilograms hydrogen generation, and total hydrogen generation. The simulation results show there are modest gains of coping time (delay of time to core damage) due to efficacy of the near-term ATF designs in various accident scenarios. The risk benefits on behalf of the core damage frequency (CDF) brought by the ATF designs would be small for most of the scenarios. However, results revealing much less hydrogen being produced at the time of core damage show a clear benefit in adopting ATFs.
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Subudhi, M., R. Morante, and A. D. Lee. "Aging Management of Reactor Coolant System Mechanical Components in Pressurized Water Reactors for License Renewal." In ASME 2002 Pressure Vessels and Piping Conference. ASMEDC, 2002. http://dx.doi.org/10.1115/pvp2002-1372.

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The reactor coolant system (RCS) mechanical components in pressurized water reactors (PWRs) that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer, steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions, determination of the effects of aging on their intended safety functions, and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. This paper presents a number of generic issues, including the time-limited aging analyses, associated with RCS components that require further review by the staff.
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Khoshgoftar Manesh, Mohammad Hasan, and Majid Amidpour. "Analysis of Pressurized Water Reactor With Pinch, Exergy and Thermoeconomic Methods." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48429.

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Exergetic, pinch and thermoeconomic analyses were performed for a 1000 MW PWR power plant similar to BUSHEHR power plant. In these analyses, mass and energy conservation laws were applied to each component of the system. Quantitative balances of the exergy and exergetic cost for each component, and for the whole system was carefully considered. The exergoeconomic model, which represented the productive structure of the system considered, was used to visualize the cost formation process and the productive interaction between components. Simulation of power plant is performed in STEAM PRO 13.0. The computer program developed in this study can determine the production costs of each component of PWR power plant such as electricity production in steam turbines. In addition, this code can provide combined exergy and pinch representation for each component of system. This graphical representation helps us to show energy integration in the nuclear power plant. The code can be also be used to study plant characteristics, namely, thermodynamic performance and sensitivity to changes in process and/or component design variables.
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Tsiklauri, G., D. Newman, G. Meriwether, and V. Korolev. "Pebble Bed Boiling Water Reactor Concept With Superheated Steam." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22045.

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An Advanced Nuclear Reactor concept is presented which extends Boiling Water Reactor technology with micro-fuel elements (MFE) and produces superheated steam. A nuclear plant with MFE is highly efficient and safe, due to ceramic-clad nuclear fuel. Water is used as both moderator and coolant. The fuel consists of spheres of about 1.5 mm diameter of UO2 with several external coatings of different carbonaceous materials. The outer coating of the particles is SiC, manufactured with chemical vapor disposition (CVD) technology. Endurance of the integrity of the SiC coating in water, air and steam has been demonstrated experimentally in Germany, Russia and Japan. This paper describes a result of a preliminary design and analysis of 3750 MWt (1500 MWe) plant with standard pressure of 16 MPa, which is widely achieved in the vessel of pressurized-water type reactors. The superheated steam outlet temperature of 550 °C elevates the steam cycle to high thermal efficiency of 42%.
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Revankar, Shripad T., Ram Anand Vadlamani, and Jovica R. Riznic. "Experimental and Theoretical Analysis of Subcooled Water Discharge Through Simulated Steam Generator Tube Cracks." In ASME 2013 International Mechanical Engineering Congress and Exposition. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/imece2013-62470.

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The steam generator (SG) tubes represent a major fraction of the reactor primary coolant pressure boundary surface area in both Canadian pressurized heavy water reactor (CANDU) reactors and pressurized water reactor (PWR). There is very limited data on the steam generator tube leak rate measurement. Most studies of subcooled choking flow are related to long tubes with L/D greater than 15. Also, all of those data have a channel length greater than 10 mm, which is not indicative of steam generator tubing. Steam generator tubes have a wall thickness typically less than 3 mm. Experiments were conducted on choking flow for various simulated crack geometries for vessel pressures up to 7 MPa with various subcoolings. Measurements were done on subcooled flashing flow rate through well defined simulated crack geometries with L/D at ∼2 and 5–6. Both homogeneous equilibrium and non-equilibrium mechanistic models were developed to model two-phase choking flow through slits. A comparison of the model results with experimental data shows that the homogeneous equilibrium based models grossly under predict choking flow rates in such geometries, while homogeneous non-equilibrium models greatly increase the accuracy of the predictions.
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Cenusa, Victor-Eduard, Florin-Niculae Alexe, Mihaela Norisor, and Madalina Barbu. "Analysing the Heat and Exergy Transfer in Steam Generators of Nuclear Power Plants with Pressurized Light Water Reactors." In 2019 International Conference on ENERGY and ENVIRONMENT (CIEM). IEEE, 2019. http://dx.doi.org/10.1109/ciem46456.2019.8937579.

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