Journal articles: '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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Wu, Wencheng, and Chaung Lin. "Optimal Reliable Control System Design for Steam Generators in Pressurized Water Reactors." Nuclear Technology 106, no. 2 (May 1994): 216–24. http://dx.doi.org/10.13182/nt94-a34977.

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Bury, Tomasz. "Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach." Nukleonika 60, no. 2 (June 1, 2015): 339–45. http://dx.doi.org/10.1515/nuka-2015-0042.

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Abstract The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

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Melikhov, Vladimir, Oleg Melikhov, Yury Parfenov, and Alexey Nerovnov. "Simulation of the Thermal Hydraulic Processes in the Horizontal Steam Generator with the Use of the Different Interfacial Friction Correlations." Science and Technology of Nuclear Installations 2011 (2011): 1–9. http://dx.doi.org/10.1155/2011/181393.

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

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Suh, Gyoo Won, and Hee Cheon No. "Dynamic Modeling and Optimum Level Controller Design for Steam Generators in Pressurized Water Reactors." Nuclear Science and Engineering 90, no. 3 (July 1985): 236–47. http://dx.doi.org/10.13182/nse85-a17765.

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El-Tokhy, Mohamed S., and Imbaby I. Mahmoud. "Temperature Coupling Analysis Between Nuclear Steam Generators and Heat Exchanger Inside Pressurized Water Reactors." Nuclear Science and Engineering 195, no. 1 (September 8, 2020): 69–100. http://dx.doi.org/10.1080/00295639.2020.1787013.

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Umminger, Klaus, Lars Dennhardt, Simon Schollenberger, and Bernhard Schoen. "Integral Test Facility PKL: Experimental PWR Accident Investigation." Science and Technology of Nuclear Installations 2012 (2012): 1–16. http://dx.doi.org/10.1155/2012/891056.

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Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

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Höhne, T., E. Krepper, and U. Rohde. "Application of CFD Codes in Nuclear Reactor Safety Analysis." Science and Technology of Nuclear Installations 2010 (2010): 1–8. http://dx.doi.org/10.1155/2010/198758.

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Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety (NRS) analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR) have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.

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Wang, Yan. "The Prospect of the Modeling Simulation for the Passive Containment Cooling System of the Advanced PWR in China." Applied Mechanics and Materials 88-89 (August 2011): 82–87. http://dx.doi.org/10.4028/www.scientific.net/amm.88-89.82.

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The passive containment cooling system (PCCS) is one main component of the various passive safety systems in the advanced 3rd generation pressurized water reactor. Several containment analysis codes are modified for the simulation and analysis on the PCCS by researchers. These codes for the PCCS were validated by comparison with transient test data from some separate effect tests and integral tests, and used to evaluate the heat-removed capability of the PCCS under the postulated events, such as a loss-of-coolant accident (LOCA) and a main steam line break (MSLB). The advanced 3rd generation pressurized water reactors of AP1000 project were build in China, and the future advanced PWR with higher power will be researched and designed. So it is very important to develop a new code with self-owned intellectual property rights, which is used for the simulation and analysis of the PCCS performance. In this paper, the main ideas of the code design, including the framework of the code, the important mechanism models and the code validation, etc., will be introduced. In addition, the computational fluid Dynamics (CFD) codes also had been applied to solve some thermal-hydraulic problems in Nuclear Reactor Safety (NRS) analysis.

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Cârdu, Mircea, and Marian Dumitru. "On a simplified system for steam production in nuclear power plants with pressurized water reactors." Energy Conversion and Management 40, no. 18 (December 1999): 1931–37. http://dx.doi.org/10.1016/s0196-8904(99)00084-9.

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Tian, Zihao, and Lixin Yang. "Numerical Investigation on a New Type of Two-Stage Steam Separator in Pressurized Water Reactors." Energy Procedia 142 (December 2017): 3962–67. http://dx.doi.org/10.1016/j.egypro.2017.12.304.

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Mukhopadhyay, D., S. K. Gupta, and V. Venkat Raj. "Analysis of Possibility of Pressure Tube Cold Pressurization During ECCS Injection for a Small Break LOCA." Journal of Pressure Vessel Technology 124, no. 4 (November 1, 2002): 483–86. http://dx.doi.org/10.1115/1.1465438.

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ECCS is designed to keep the reactor fuel temperatures within safe limits. The paper describes an additional criterion for Indian pressurized heavy water reactors (IPHWRs) evolving from the need to avoid a small break loss of cooling accident (LOCA) developing into a more severe accident. During a small break loss of coolant accident (LOCA) in PHWRs, the hydro-accumulators ride on the system and inject emergency coolant. The atmospheric steam discharge valves (ASDVs) open and cool the system due to energy discharge. In addition, the pressure control system tends to maintain the pressure. Depending on the system design, this could lead to cold pressurization of the system. This paper examines this issue.

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Isobe, Yoshihiro, Mitsuyuki Sagisaka, Shinobu Yoshimura, and Genki Yagawa. "Economic Evaluation of Maintenance Strategies for Steam Generator Tubes Using Probabilistic Fracture Mechanics and a Financial Method." Solid State Phenomena 120 (February 2007): 119–26. http://dx.doi.org/10.4028/www.scientific.net/ssp.120.119.

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As an application of probabilistic fracture mechanics (PFM) and a financial method, a risk-benefit model was developed for the purpose of optimizing maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). To justify whether or not it is worth while implementing the selected maintenance strategy in terms of an economic point of view, net present value (NPV) was calculated as an index which is one of the most fundamental financial indices for decision-making based on the discounted cash flow (DCF) method.

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Jeon, Soon-Hyeok, Geun Song, and Do Hur. "Micro-Galvanic Corrosion of Steam Generator Materials within Pores of Magnetite Flakes in Alkaline Solutions." Metals 8, no. 11 (November 2, 2018): 899. http://dx.doi.org/10.3390/met8110899.

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In secondary coolant system of the pressurized water reactors, the reduced corrosion products such as metallic Cu and Pb particles were accumulated in the pores of the magnetite flakes and electrically contacted to the steam generator materials. The micro-galvanic corrosion behavior of steam generator materials (steam generator tube materials: Alloy 600 and Alloy 690, steam generator tube sheet materials: SA508 Gr.3) contacted to the corrosion products (magnetite, Cu, and Pb) was investigated in an alkaline solution. The steam generator materials considered in this study were all the anodic elements of the galvanic pair because their corrosion potentials were lower than those of the corrosion products. The corrosion rate of the steam generator materials was increased by the galvanic coupling with the each corrosion products, and was more accelerated with increasing the area ratio of the corrosion products to the steam generator materials. Among the corrosion products, Cu has the largest galvanic effect on steam generator materials in the pores when area ratio of cathode to anode is 10.

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Dileep, B. N., P. M. Ravi, Managanvi Sangamesh, and N. Karunakara. "An advanced method for quick detection of heavy water leak in steam generators of pressurized heavy water reactors." Nuclear Engineering and Design 359 (April 2020): 110447. http://dx.doi.org/10.1016/j.nucengdes.2019.110447.

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Song, Geun Dong, Jeoh Han, Soon-Hyeok Jeon, and Do Haeng Hur. "Magnetite-Accelerated Corrosion of SA508 Tubesheet Material and Its Effect on Steam Generator Tube Denting." Coatings 9, no. 9 (September 9, 2019): 575. http://dx.doi.org/10.3390/coatings9090575.

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The objective of this work is to investigate the magnetite-accelerated corrosion phenomenon of SA508 used as tubesheet material in simulated secondary side environments of pressurized water reactors through immersion and electrochemical tests. The presence of sulfate ions induced the fast growth of a corrosion product layer on SA508, and this phenomenon was accelerated when the SA508 was coupled to magnetite. From the perspective of electrochemical behavior, it was found that SA508 behaves as an anodic member in the coupling system with magnetite, resulting in an increased corrosion rate.

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Hamedani, A., O. Noori-Kalkhoran, R. Ahangari, and M. Gei. "Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)." Kerntechnik 85, no. 1 (December 1, 2020): 54–67. http://dx.doi.org/10.1515/kern-2020-850108.

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Abstract Steam generators are one of the most important components of pressurized-water reactors. This component plays the role of heat transfer and pressure boundary between primary and secondary side fluids. The Once Through Steam Generator (OTSG) is an essential component of the integrated nuclear power system. In this paper, steady-state analysis of primary and secondary fluids in the Integral Economizer Once Through Steam Generator (IEOTSG) have been presented by Single Heated Channel (SHC) and subchannel modelling. Models have been programmed by MATLAB and FORTRAN. First, SHC model has been used for this purpose (changes are considered only in the axial direction in this model). Second, the subchannel approach that considers changes in the axial and also radial directions has been applied. Results have been compared with Babcock and Wilcox (B&W) 19- tube once through steam generator experimental data. Thermal- hydraulic profiles have been presented for steam generator using both of models. Accuracy and simplicity of SHC model and importance of localization of thermal-hydraulic profiles in subchannel approach have been proved.

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Morghi, Youssef, Jesus Puente, Amir Mesquita, and Ana Baliza. "INVESTIGATION OF COUNTER-CURRENT FLOW LIMITATION FOR AIR-WATER IN A PWR HOT LEG EXPERIMENTAL LOOP FOR DIFFERENT GEOMETRY." International Journal of Engineering Technologies and Management Research 5, no. 2 (February 10, 2020): 198–212. http://dx.doi.org/10.29121/ijetmr.v5.i2.2018.164.

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Gas/liquid two-phase stratified flows in horizontal channels are frequently encountered in nuclear reactors, oil and gas pipelines, steam generators, refrigeration equipment, reflux condensers, packed columns, and heat pipes. The phenomenon known as countercurrent flow limitation, or flooding, is the limiting condition where the flow rates of neither the gas nor the liquid can be further increased without changing the flow pattern. This is the condition where the maximum air mass flow rate at which the down-flowing water mass flow rate is equal to the inlet water mass flow rate. This limiting condition, also known as onset of flooding, can occur in vertical or horizontal geometry. This work is a review of recent experimental investigations of countercurrent flow limitation (CCFL) for various hot-leg geometries of pressurized water reactors (PWRs). We compare results with those obtained from the Nuclear Technology Development Centre (CDTN) in 2005. Recent experimental results in the literature are in good agreement with the 2005 findings.

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Ward, Leonard W. "Evaluation of the Loss of Residual Heat Removal Systems in Pressurized Water Reactors with U-Tube Steam Generators." Nuclear Technology 100, no. 1 (October 1992): 25–38. http://dx.doi.org/10.13182/nt92-a34751.

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Jain, J. K., Sundar Singh, M. K. Mathur, N. P. Srivastava, and K. P. Dwivedi. "Experiences in Manufacturing and Qualifying of Stainless Steel AISI 403 Grade Materials for PHWR Reactors of NPCIL." Advanced Materials Research 794 (September 2013): 194–200. http://dx.doi.org/10.4028/www.scientific.net/amr.794.194.

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NPCIL uses stainless steel AISI 403 grade material extensively for very critical nuclear components. A very special application of this material is for manufacturing END FITTING BODY of COOLANT CHANNEL ASSEMBLIES of Pressurized Heavy Water Reactor (PHWR) of Nuclear Power Plants. Presently, two types of PHWR Nuclear Plants viz 220 MWe and 540 MWe are under operation in India. Next higher rating viz 700 MWe PHWR Nuclear Plants are at various stages of construction and due for commissioning in near future. There are number of Pressure Tubes in PHWR Nuclear Plants wherein fuel is loaded and are called coolant channel assemblies. End fitting bodies are assembled on either end of PRESSURE TUBES of coolant channel assembly. The Coolant (Heavy water) flows inside these assemblies and collects the heat produced during nuclear reaction in pressure tubes. This heat is utilized to produce Steam in Steam Generators. After giving up heat in Steam Generator, coolant returns to pressure tube and cycle continues. Material for End Fittings has been selected keeping in mind the mechanical properties at room and at elevated temperatures are compatible with mating component which are manufactured using Zircaloy and SS 410. It is also an essential that coefficient of thermal expansion is also compatible for mating material. Indigenization of the material for manufacturing of End Fittings with short delivery period has always been a challenge for commercial Nuclear Reactors. Inspection of these types of critical nuclear reactor components becomes critical due to the high safety standards and complexity in structures. There are very few manufacturers for this material in India. NPCIL took up a challenge for developing a new supplier in Northern Region of India for developing this material. A leading company in northern region of INDIA came forward and took up the challenge and succeeded for this venture. This paper deals with development of vender to manufacture of AISI 403 grade material by melting, refining ( ESR process), forging, heat treatment, testing & provide the guidelines to tackle the issue faced during this developmental work.

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Lee, Yong-Beom, Ji-Min Lee, Do-Haeng Hur, Jong-Hyeon Lee, and Soon-Hyeok Jeon. "Effects of Advanced Amines on Magnetite Deposition of Steam Generator Tubes in Secondary System." Coatings 11, no. 5 (April 27, 2021): 514. http://dx.doi.org/10.3390/coatings11050514.

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Ethanolamine (ETA) is widely used to control the pH value in the secondary water of pressurized water reactors. However, it is necessary to consider other advanced amines which can replace ETA due to its serious human hazards and environmental treatment problems. The purpose of this study is to contemplate the effects of three advanced amines (ETA, 3-methoxypropylamine (MPA), and dimethylamine (DMA)) on the magnetite deposition behavior of a thermally treated (TT) Alloy 690 tubes by using a steam generator (SG) tube fouling loop in simulated secondary water. All particles were identified as a magnetite and were polyhedral or spherical in shape. When using ETA, MPA, and DMA, the average porosity of the deposits was about 34.7%, 33.0%, and 24.6%, respectively. The amount of deposits was largest when ETA was added, and it decreased by 41% when adding MPA and 55% when adding DMA. The mechanism of magnetite deposition was discussed in terms of zeta potentials of both the magnetite particles and the Alloy 690TT surface and magnetite solubility depending on the amines. To compare the potential for replacing ETA with other advanced amines, the various factors such as SG integrity, human hazards, and environmental treatment problems were discussed.

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Nakata, Alexandre Ezzidi, Masanori Naitoh, and Chris Allison. "NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 21, no. 3 (November 12, 2019): 119. http://dx.doi.org/10.17146/tdm.2019.21.3.5630.

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Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) of the European Commission (EC) in Holland (Europe) for a Pressurized Water Reactor (PWR). The obtained results of both projects have shown very large discrepancies between the used severe accident codes for both reactor types BWR and PWR. Consequently, the results for a real plant analysis by these integral codes, may not be correct after the beginning of core melt. Discrepancies of results of ex-vessel phenomena in the containment between the codes are in general larger. Therefore, there is a strong need for a reliable new generation mechanistic severe accident code which can simulate severe accident scenarios from an initiating event till containment failure with better accuracy not only for existing light water reactors but also for new generation IV reactor types. SAMPSON mechanistic ex-vessel modules coupled with SCDAPSIM and a new thermal-hydraulic module ASYST-ISA with particularly newly developed options for the reactor coolant system (RCS) and material properties applicable to new reactor deigns, is proposed as a best etimate new generation severe accident code for several reasons which are described in this paper.Keywords: Severe accident, SAMPSON, SCDAPSIM, ASYST-ISA, Steam explosion, Hydrogen detonation

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Saini, Nadish, and Igor A. Bolotnov. "Two-Phase Turbulence Statistics from High Fidelity Dispersed Droplet Flow Simulations in a Pressurized Water Reactor (PWR) Sub-Channel with Mixing Vanes." Fluids 6, no. 2 (February 6, 2021): 72. http://dx.doi.org/10.3390/fluids6020072.

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In the dispersed flow film boiling regime (DFFB), which exists under post-LOCA (loss-of-coolant accident) conditions in pressurized water reactors (PWRs), there is a complex interplay between droplet dynamics and turbulence in the surrounding steam. Experiments have accredited particular significance to droplet collision with the spacer-grids and mixing vane structures and their consequent positive feedback to the heat transfer recorded in the immediate downstream vicinity. Enabled by high-performance computing (HPC) systems and a massively parallel finite element-based flow solver—PHASTA (Parallel Hierarchic Adaptive Stabilized Transient Analysis)—this work presents high fidelity interface capturing, two-phase, adiabatic simulations in a PWR sub-channel with spacer grids and mixing vanes. Selected flow conditions for the simulations are informed by the experimental data found in the literature, including the steam Reynolds number and collision Weber number (Wec={40,80}), and are characteristic of the DFFB regime. Data were collected from the simulations at an unprecedented resolution, which provides detailed insights into the continuous phase turbulence statistics, highlighting the effects of the presence of droplets and the comparative effect of different Weber numbers on turbulence in the surrounding steam. Further, axial evolution of droplet dynamics was analyzed through cross-sectionally averaged quantities, including droplet volume, surface area and Sauter mean diameter (SMD). The downstream SMD values agree well with the existing empirical correlations for the selected range of Wec. The high-resolution data repository from the simulations herein is expected to be of significance to guide model development for system-level thermal hydraulic codes.

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Jeon, Soon-Hyeok, Hee-Sang Shim, Ji-Min Lee, Jeoh Han, and Do Haeng Hur. "Simulation of Porous Magnetite Deposits on Steam Generator Tubes in Circulating Water at 270 °C." Crystals 10, no. 9 (August 20, 2020): 729. http://dx.doi.org/10.3390/cryst10090729.

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In the secondary side of pressurized water reactors (PWRs), the main corrosion product accumulated on the steam generator (SG) tubes is magnetite, which has a porous structure. The purpose of this work is to simulate the porous magnetite deposited to the SG tubes using a loop system. We newly developed a circulating loop system for a porous magnetite deposition test. A test section was designed as a single hydraulic flow channel, and a cartridge heater was fabricated and mounted into a commercial SG tube to provide an equal heating source for the primary water. After the deposition test, the simulated magnetite deposits were characterized for comparison to real SG tube deposits collected from an operating PWR plant. The magnetite deposits produced using the loop system were appropriate for simulating the real SG tube deposits because the particle characteristics, phase, and porous morphology are closely similar to those of real deposit samples. Using the loop system, the chemical impurities such as Na and Cl can be easily concentrated within the pores of the simulated magnetite deposits. These simulated magnetite samples are expected to be widely utilized in various research fields such as the heat transfer degradation and magnetite accelerated corrosion of SG tubes.

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Bosholm, J. "Entfernung eisenoxidhaltiger Beläge aus Dampferzeugern von Druckwasserreaktoren / Removal of deposits containing iron oxide from steam generators of pressurized water reactors." Kerntechnik 58, no. 1 (January 1, 1993): 37–39. http://dx.doi.org/10.1515/kern-1993-580115.

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Song, Sung Jin, Chang Hwan Kim, Deok Hyun Lee, Myung Sik Choi, Do Haeng Hur, and Jung Ho Han. "An Automated System for Detection of Through-Wall Cracks in Racks in Steam Generator Tubes." Key Engineering Materials 297-300 (November 2005): 2071–76. http://dx.doi.org/10.4028/www.scientific.net/kem.297-300.2071.

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Through-wall axial cracks occurred by primary water stress corrosion are one of the serious defects in steam generator (SG) tubes (made of alloy 600) in pressurized water reactors. Therefore, it is necessary to detect and size them by eddy current testing (ECT) conducted during in-service inspection of SG tubes. To address this issue, it has been recently proposed an effective method, namely „M-shape profile“ approach, which relies on the difference in the amplitude between the pancake and plus point coils in a MRPC probe. Even though the M-shape curve approach is straightforward in principle, it requires time-consuming data processing if performed by human operators. In order to get rid of this tedious task, an automated system is developed in the present work. This paper addresses the principle of the M-shape approach together with the automated system and its performances for the detection of natural axial cracks in SG tubes. The results observed in the present work demonstrate the high potential of the developed system as a very promising tool for detecting through-wall cracks in many practical field applications.

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36

Polley, Martin Victor. "A Correlation Between Operation of Primary Coolants at Low pHTwith Steam Generator Channel Head Dose Rates in Westinghouse Pressurized Water Reactors." Nuclear Technology 71, no. 3 (December 1985): 557–67. http://dx.doi.org/10.13182/nt85-a33678.

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ADIWIBOWO, PRIYO HERU. "KARAKTERISITIK FLOW PATERN PADA ALIRAN DUA FASE GAS-CAIRAN MELEWATI PIPA VERTIKAL." Jurnal Teknik Industri 11, no. 2 (February 9, 2012): 117. http://dx.doi.org/10.22219/jtiumm.vol11.no2.117-122.

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Multi-phase flows are widely encountered in several engineering and industrial facilities, such as conventional steam power plants, evaporators and condensers, pressurized-water nuclear reactors, a wide variety of petroleum industries, chemicals and food processing industries. Surely, in the complex pipeline installation of these systems, vertical pipe will be commonly used for pipe connection. The purpose of this work is to investigate the flow pattern of gas-liquid two phase in the vertical pipe. Experiments will be performed in a 36 mm ID acrylic pipe vertical. Superifical liquid velocities and volumetric gas quality will be varied 0.3~1,1 m/s and 0.05~0.2 respectively. Digital camera will be used for flow pattern visualization in the vertical pipe. It was observed that effect of vertical pipe on flow pattern formed cluster bubbly flow for low volumetric gas quality with high superifical liquid velocities. For superifical liquid velocities with medium volumetric gas quality formed homogeneous bubbly flow and high volumetric gas quality is dense bubbly flow.

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38

Staehle, R. W., and J. A. Gorman. "Quantitative Assessment of Submodes of Stress Corrosion Cracking on the Secondary Side of Steam Generator Tubing in Pressurized Water Reactors: Part 1." CORROSION 59, no. 11 (November 2003): 931–94. http://dx.doi.org/10.5006/1.3277522.

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Staehle, R. W., and J. A. Gorman. "Quantitative Assessment of Submodes of Stress Corrosion Cracking on the Secondary Side of Steam Generator Tubing in Pressurized Water Reactors: Part 3." CORROSION 60, no. 2 (February 2004): 115–80. http://dx.doi.org/10.5006/1.3287716.

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Staehle, R. W., and J. A. Gorman. "Quantitative Assessment of Submodes of Stress Corrosion Cracking on the Secondary Side of Steam Generator Tubing in Pressurized Water Reactors: Part 2." CORROSION 60, no. 1 (January 2004): 5–63. http://dx.doi.org/10.5006/1.3299232.

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41

Wazzan, A. R., H. Procaccia, J. David, A. Fromal, and P. Pitner. "Thermal-hydraulic characteristics of pressurized water reactors during commercial operation: VII. Thermal-hydraulics of the PWR Paluel 1 steam generator during commercial operation." Nuclear Engineering and Design 105, no. 3 (January 1988): 285–93. http://dx.doi.org/10.1016/0029-5493(88)90250-6.

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Hwang, Seong Sik, Nam Gung Chan, Jang Yul Park, Man Kyo Jung, Hong Pyo Kim, and Joung Soo Kim. "Leak Behavior of SCC Degraded SG Tubes at a Constant Pressure." Key Engineering Materials 345-346 (August 2007): 1345–48. http://dx.doi.org/10.4028/www.scientific.net/kem.345-346.1345.

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A primary water stress corrosion cracking (SCC) and an outside diameter SCC have occurred in the steam generator (SG) tubes of nuclear power plants around the world. It is important to establish the repair criteria for the degraded tubes to assure a reactor integrity, and yet maintain the plugging ratio within the limits needed for an efficient operation. For assessment and management of the degradation, it became crucial to understand initial leak behaviors under a small pressure and leak rate evolution under a constant pressure of SCC flaws. Stress corrosion cracked tube specimens were prepared by using a room temperature cracking technique, and leak behaviors of these tubes were measured at room temperature. Water pressure inside the tube was increased slowly in a step like manner with a designated holding time. Water leak rates just after a ligament rupture were measured by collecting the leaked water in a plastic container for a designated time. A leak rate was calculated by dividing the amount of water by the time. Under 3.45 MPa, a small water droplet was formed, but it did not grow after a 10 minute holding period at a constant pressure of 3.45 MPa. A throughwall crack seemed to open at around 8.28 MPa (1200 psi). Some tubes with 100 % through wall cracks did not show a leakage at 10.8 MPa, which is a typical pressure difference of pressurized water reactors (PWRs) during a normal operation. The higher the pressure was applied, the larger the rates of increase with the time were. Axial cracks showed a lower leak pressure than that of the circumferential cracks, which might be from a higher hoop stress than the axial stress. A large open and long axial crack showed an increasing leak rate with the time at a constant pressure.

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Lee, Ji-Min, Dong-Seok Lim, Soon-Hyeok Jeon, and Do Haeng Hur. "Zeta Potentials of Magnetite Particles and Alloy 690 Surfaces in Alkaline Solutions." Materials 13, no. 18 (September 9, 2020): 3999. http://dx.doi.org/10.3390/ma13183999.

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Magnetite particles deposited on the secondary side of a steam generator (SG) can degrade the integrity and performance of pressurized water reactors. Therefore, it is necessary to produce the data of fundamental interfacial electrokinetic properties of magnetite particles and SG tube materials. This study investigated the zeta potentials of magnetite nanoparticles and Alloy 690 surfaces, which were dependent on the pH value, pH agent, and the presence of NaCl. The zeta potentials of the magnetite nanoparticles increased in the negative direction as the pH increased, regardless of the pH agent. At the same pH value, the absolute values of the zeta potentials with different pH agents were: ethanolamine < ammonia < morpholine. In the presence of NaCl, the zeta potentials of the particles further increased negatively. The meaning of the measured zeta potentials was discussed in terms of the dispersion stability and the agglomeration of the particles. Based on the relationship between the zeta potentials of the particles and Alloy 690 surfaces, the magnetite deposition on Alloy 690 was also discussed. Furthermore, the empirical formulas for the pH-dependent zeta potentials of magnetite particles in each alkaline solution were suggested.

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Garud, Y. S. "Incremental Damage Formulation and Its Application to Assess IGSCC Growth of Circumferential Cracks in a Tube." Corrosion 47, no. 7 (July 1, 1991): 523–27. http://dx.doi.org/10.5006/1.3585288.

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Abstract Recent cracking indications in the roll-transition (RT) zone of steam-generator tubes of several pressurized water reactors emphasize the greater need for quantitative assessment of the cracking response, especially for the safe and economic continued operation of the reactors. To address this need, the author recently proposed and examined a strain rate-based incremental damage formulation using the concept of an effective stress. Details of the formulation and its application to the tube cracking phenomenon are presented in this paper. The crack growth model consists of a set of constitutive relations for the material deformation (including creep), an effective stress relation that couples the local and nominal deformation responses and accounts for the crack size and its growth rate, and the corrosion damage relation in terms of the local strain rate and environmental parameters. In this model, the crack initiation and propagation phases do not need to be treated separately; nor is it a requirement to assume a pre-existing crack or a crack-like defect. The formulation allows determination of the instantaneous crack growth rate based on the CERT data. Its application to the RT cracking is illustrated with simplifying assumptions for the state of stress and the shape of crack likely to exist in the RT zone; influence of the stress and temperature on the crack growth response is evaluated. Practical implications of both the modeling approach and the results of its application are briefly discussed.

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Krieg, R., T. Malmberg, G. Messemer, T. Stach, and E. Stratmanns. "Slug Impact Loading on the Vessel Head during a Postulated In-Vessel Steam Explosion in Pressurized Water Reactors–Assessments and Discussion of the Investigation Strategy." Nuclear Technology 111, no. 3 (September 1995): 369–85. http://dx.doi.org/10.13182/nt95-a15867.

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Yu, Hao, and Minjun Peng. "Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario." Science and Technology of Nuclear Installations 2019 (August 25, 2019): 1–10. http://dx.doi.org/10.1155/2019/1510967.

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Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

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Kim, Dong Jin, Hyuk Chul Kwon, Seong Sik Hwang, and Hong Pyo Kim. "A Passive Film Formed on Alloy 600 in High Temperature Aqueous Solution." Solid State Phenomena 135 (February 2008): 103–6. http://dx.doi.org/10.4028/www.scientific.net/ssp.135.103.

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Alloy 600(74wt% Ni, 15wt% Cr, 9wt% Fe) is used as a material for steam generator tubing in pressurized water reactors(PWR) due to its high corrosion resistance under a PWR environment whose primary side and secondary side are operated under 340oC, 150bar and 290oC, 50bar, respectively. However, in spite of its outstanding corrosion resistance, a stress corrosion cracking(SCC) which has been one of the most important degradation issues, has occurred occasionally owing to the severe high temperature and pressure condition of a PWR. It is expected that a SCC is deeply related to an electrochemical property of a passive film formed under a high temperature and pressure condition. Therefore an understanding of the basic electrochemical behaviors regarding an anodic dissolution and a passivation of the bare surface of metals and alloys provides important information about a SCC mechanism and control. In the present work, the passive oxide films on Alloy 600 were investigated as a function of the dissolved oxygen content and the pH by using a potentiodynamic polarization, electrochemical impedance spectroscopy(EIS) and a depth profiling by Auger electron spectroscopy(AES). For this purpose, 0.5M H3BO3, 0.5M Na2SO4 and 0.1M NaOH aqueous solutions with/without dissolved oxygen at 300oC were used as the test solutions. The resistance of an oxide formed in an aerated solution was smaller than that in a deaerated solution while the oxide film in the aerated solution was thicker than the oxide film in the deaerated solution.

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Kim, Dong Jin, Hyuk Chul Kwon, Seong Sik Hwang, and Hong Pyo Kim. "Material Property of a Passive Oxide Formed on Alloy 600." Advanced Materials Research 26-28 (October 2007): 937–40. http://dx.doi.org/10.4028/www.scientific.net/amr.26-28.937.

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Alloy 600 is used as a material for a steam generator tubing in pressurized water reactors(PWR) due to its high corrosion resistance under a PWR environment. In spite of its corrosion resistance, a stress corrosion cracking(SCC) has occurred on the primary side as well as the secondary side of a tubing. It is known that a SCC is related to the electrochemical behaviors of an anodic dissolution and a passivation of a bare surface of metals and alloys. Therefore in the present work, the passive oxide films on Alloy 600 have been investigated as a function of the solution temperature by using a potentiodynamic polarization, electrochemical impedance spectroscopy and a TEM, equipped with EDS. Moreover the semiconductive property was evaluated by using the Mott-Schottky relation. It was found that the passivity depends on the chemical composition and the densification of the oxide film rather than the oxide thickness. As the solution temperature of 0.5M H3BO3 increased, the thickness of the passive film increased but the oxide resistance of the passive film was decreased, indicating that the measured current in the passive region of the potentiodynamic curve is closely related to the stability of the passive film rather than the oxide thickness. It was found that the oxide films were composed of an outer oxide layer with a lower resistance and an inner oxide layer with a relatively higher resistance. From the Mott-Schottky relation, the oxide formed at 300oC showed a p-type semiconductor property unlike the n-type oxide films up to 250oC.

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Loubet, L., P. G. Allinei, P. Schindler, and V. Pointeau. "Implementation of three gamma measuring stations on the Colentec loop in Cadarache for the on line observation of the clogging phenomena in Steam Generator." EPJ Web of Conferences 225 (2020): 03011. http://dx.doi.org/10.1051/epjconf/202022503011.

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The Tube Support Plate blockage, also named clogging, is a complex phenomenon that can occur in the steam generator of Pressurized Water Reactors. This deposit of iron oxides, that can reduce the coolant flux and constraint the primary tubes, could have significant consequences on heat exchanges and the integrity of the primary tubes. Since 2014, an experimental program carried out jointly by CEA-EDF, has been performed on a representative dedicated equipment, the COLENTEC loop at Cadarache. The physico-chemical properties of the deposit formed (chemical composition, thickness...) were determined at the end of each experimental campaign lead at thermohydraulic and chemical stable conditions. ... In order to determine the influence on clogging of parameters such as the chemistry, the temperature, the pressure in an on line process and without performing separate experimental campaigns, a method of measurement on radioactive tracer, the gamma emitter 59Fe, has been studied by our laboratory. First, a modelization with a monte carlo code, MCNP, has been carried out with the geometrical configuration of the Tube Support Plate zone of COLENTEC in order to determine the most adapted measuring station. These calculations had shown that a 30% efficiency, high purity, germanium detector, with a specific collimation would be able to detect the accumulation of 59Fe in the Tube Support Plate of COLENTEC. Then, in order to improve and ensure those important online measurements, our laboratory has proposed the installation of two additional measuring stations. One will be dedicated to a second on-line measurement, with a gamma detector of the same type, but positioned on a section of the loop remote from the previous area, to verify that there is no significant secondary deposit evolution elsewhere on the circuit. The last additional gamma measurement station will consist of a low background system able to carry out 59Fe concentration measurements of liquid samples taken from the loop during the test. This paper presents the context of this experiment and then focuses on the spectrometry measuring stations, modelizations and tests.

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CHO, SUNG-KEUN, CHANG-SUNG SEOK, BONG-KOOK BAE, and JAE-MEAN KOO. "EVALUATION OF THE HOOP TENSILE PROPERTIES OF A STEAM GENERATOR TUBE." International Journal of Modern Physics B 20, no. 25n27 (October 30, 2006): 4129–34. http://dx.doi.org/10.1142/s0217979206040970.

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The steam generators in a pressurized water reactor (PWR) are large heat exchangers that use the heat from the primary reactor coolant to make steam on the secondary-side to drive turbine generators. Hoop stress is known to be the main cause of fracture of inner pressurized tubes such as the steam generator tube. However, because the steam generator tube is too small to be manufactured to a standard tensile specimen in the hoop direction, the axial tensile properties of the steam generator tube (or original material properties) instead of hoop tensile properties have been used to estimate the fracture properties of a steam generator tube. In this study, we have conducted not only axial tensile tests but also ring tensile tests. From these test, both the axial and hoop tensile properties of steam generator tubes were obtained, and the reliability of the hoop tensile properties were confirmed by burst test of a real steam generator tube.

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

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