Critical Heat Flux
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Nucleate-Boiling Heat Transfer to Water at Atmospheric Pressure
Scholars' Mine Masters Theses Student Theses and Dissertations 1966 Nucleate-boiling heat transfer to water at atmospheric pressure H. D. Chevali Follow this and additional works at: https://scholarsmine.mst.edu/masters_theses Part of the Chemical Engineering Commons Department: Recommended Citation Chevali, H. D., "Nucleate-boiling heat transfer to water at atmospheric pressure" (1966). Masters Theses. 2971. https://scholarsmine.mst.edu/masters_theses/2971 This thesis is brought to you by Scholars' Mine, a service of the Missouri S&T Library and Learning Resources. This work is protected by U. S. Copyright Law. Unauthorized use including reproduction for redistribution requires the permission of the copyright holder. For more information, please contact [email protected]. NUCLEATE-BOILING HEAT TRANSFER TO WATER AT ATMOSPHERIC PRESSURE BY H. D. CHEV ALI - 1 &J '-I 0 ~~ A THESIS submitted to the faculty of THE UNIVERSITY OF MISSOURI AT ROLLA in partial fulfil lment of the requirements for the Degree of MASTER OF SCIENCE IN CHEMICAL ENGINEERING Roll a, Missouri 1966 Approved by / (Advisor ) ii ABSTRACT The purpose of this investigation was to study the hysteresis effect, the effect of micro-roughness and orientation of the heat-trans fer surface, and the effect of infra-red-radiation-treated heat-transfer surfaces on the nucleate-boiling curve. It has been observed that there exists no hysteresis effect for water boiling from a cylindrical copper surface in the nucleate-boiling region over the range studied. The nucleate-boiling curve has been found to be independent of micro-roughness and orientation of the heat transfer surface. There was no detectable change in the nucleate-boil ing characteristics of the heat-transfer surface when the surface was treated with infra-red radiation. -
Arevas Thermal-Hydraulic Platform Content
Focus on Technology AREVAs Thermal-Hydraulic Platform Content Infrastructure and Technology • Thermal-Hydraulic Platform - Unique in the World • Infrastructure for Full-Scale Thermal-Hydraulic Test Facilities • Fluid-Dynamic and Thermal-Hydraulic Analysis • Similitude Tests, Optimization of Components and Processes • BENSON - Thermal-Hydraulic Separate Effect Tests • Vibrations and Mechanical Tests, Optimization of Power Plant Components • Seismic and Vibration Testing • Flow-Induced Vibration Tests, Optimization of Power Plant Components • Flow Model Tests, Optimization of Power Plant Components and Processes • Test Facilities for Power and Process Industry Applications Integral Loops • INKA - Karlstein Integral Test Stand • PKL - PWR Integral System Test Facility Qualification of Components • KOPRA - Component Test Facility, Qualification and Testing of Components at Full Scale • KOPRA - Core Component Test Section, Qualification of Primary-System Components • KOPRA - Test Section for Control Rod Drive Mechanisms • KOPRA - Valve Test Section • KOPRA - Special Valve Analyzing and Testing • KATHY Loop for Critical Heat Flux (CHF) Tests • PETER - PWR Fuel Element Tests at Erlangen • Fuel Assembly’s Components Testing • Reactor Steam Generator Component Testing • GAP The World’s largest Valve Test Facility • APPEL - AREVA Pump Test Loop • DEREST - Debris Retention System Test Facility • JAVAPlus Test Facility for Qualifying FCVSPlus • KADYSS - Test Facility for Qualifying Pump Seal System • Environmental Qualification of Containment Components -
Experimental Determination of Heat Transfer Coefficients in Uranium Zirconium Hydride Fuel Rod
2005 International Nuclear Atlantic Conference - INAC 2005 Santos, SP, Brazil, August 28 to September 2, 2005 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 85-99141-01-5 EXPERIMENTAL DETERMINATION OF HEAT TRANSFER COEFFICIENTS IN URANIUM ZIRCONIUM HYDRIDE FUEL ROD Amir Z. Mesquita, Hugo C. Rezende e Antônio Carlos L. da Costa Centro de Desenvolvimento da Tecnologia Nuclear (CDTN / CNEN – MG) Campus da UFMG - Pampulha 30.123-970 Belo Horizonte, MG [email protected] , [email protected] e [email protected] ABSTRACT This work presents the experiments and theoretical analysis to determine the temperature parameter of the uranium zirconium hydride fuel elements, used in the TRIGA IPR-R1 Research Nuclear Reactor. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally. It was also presented a correlation for the gap conductance between the fuel and the cladding. In the case of nuclear fuels the heat parameters become functions of the irradiation as a result of change in the chemical and physical composition. The value of the heat transfer coefficients should be determined experimentally. 1. INTRUDUCTION The TRIGA IPR-R1 Nuclear Research Reactor, of the Nuclear Technology Development Center - CDTN, localized in Belo Horizonte (Brazil) is a Mark I type, manufactured by General Atomic, cooling by light water, open-pool design and having as fuel an alloy of zirconium hydride and uranium enriched at 20% in 235 U. The heat, generated by the fissions is transferred from fuel elements to the cooling system through the interface fuel/cladding (gap) and the cladding to the coolant. -
Prediction of Critical Heat Flux (CHF) for Vertical Round Tubes with Uniform Heat Flux in Medium Pressure Regime
Korean J. Chem. Eng., 21(1), 75-80 (2004) Prediction of Critical Heat Flux (CHF) for Vertical Round Tubes with Uniform Heat Flux in Medium Pressure Regime W. Jaewoo Shim† and Joo-Yong Park Department of Chemical Engineering, Dankook University, Seoul 140-714, Korea (Received 2 July 2003 • accepted 8 November 2003) Abstract−The description of critical heat flux (CHF) phenomena under medium pressure (10 bar≤P≤70.81 bar) regime is complex due to the large specific volume of vapor and the effect of buoyancy that are inherent in the con- ditions. In this study, a total of 2,562 data points of CHF in uniformly heated round vertical tube for water were collected from 5 different published sources. The data consisted of the following parameter ranges: 93.7≤G (mass ≤ 2 ≤ ≤ ≤ ≤ ≤ ≤ 2 flux) 18,580 kg/m s, 0.00114 D (diameter) 0.03747 m, 0.008 L (length) 5 m, 0.26 qc (CHF) 9.72 MW/m , and −0.21≤L (exit qualities)≤1.09. A comparative analysis is made on available correlations, and a new correlation is presented. The new CHF correlation is comprised of local variables, namely, “true” mass quality, mass flux, tube diameter, and two parameters as a function of pressure only. This study reveals that by incorporating “true” mass quality in a modified local condition hypothesis, the prediction of CHF under these conditions can be obtained quite accurately, overcoming the difficulties of flow instability and buoyancy effects. The new correlation predicts the CHF data are significantly better than those currently available correlations, with average error 2.5% and rms error 11.5% by the heat balance method. -
Experimental Study on Critical Heat Flux Behavior in Single Fuel Pin with and Without Wire Spacer
EXPERIMENTAL STUDY ON CRITICAL HEAT FLUX BEHAVIOR IN SINGLE FUEL PIN WITH AND WITHOUT WIRE SPACER Dan Tri LE*, Noriaki IN ABA** and Minoru TAKAHASHI** *Department of Nuclear Engineering, Tokyo Institute of Technology Nl-18, 2-12-1 Ookayama, Meguro-ku, Tokyo, 152-8550, Japan [email protected]. ac .j p **Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology Nl-18, 2-12-1 Ookayama, Meguro-ku, Tokyo, 152-8550, Japan inaba@2phase .nr.titech. ac.j p [email protected] Abstract: Light water reactor could have fast neutron spectrum with high conversion ratio nearly equal unity by using tight lattice fuel assembly with wire spacer. Besides, the critical heat flux (CHF) data base for grid spacer has a great quantity of studies but it is not in case of wire type. In order to investigate coolability in tight lattice core in boiling water reactor (BWR), an experiment of critical heat flux (CHF) was conducted using single pin flow channels with a heater pin with and without wire. We determined the critical heat flux for this system by varying the inlet temperature from 333 to 373 K and mass fluxes from 200 to 700 kg/m1 2s and the pressure was atmospheric pressure. The result show the CHF data in two-phase flow condition base on inlet and outlet condition in both case of heater pin with and without wire. The CHF values were higher with wire than without wire due to the effect of wire and spiral flow. Keywords: Critical Heat Flux, Tight Lattice, Two-phase Flow, Wire Spacer. -
Or Thermal Hydraulics in Nuclear Fission and Fusion Dr Michael Bluck Imperial College Introduction U Role of TH in Nuclear
‘Plumbing’, or Thermal hydraulics in Nuclear Fission and Fusion Dr Michael Bluck Imperial College Introduction u Role of TH in nuclear u Coolability u Normal operation u Accident scenarios u Issues in fission u TH Analysis; Systems codes & correlations, CFD u Boiling, critical heat flux, natural circulation u Bridging the scales u Issues in fusion u Tok om a k s u Fusion blankets u Magnetohydrodynamics Role of TH in nuclear: Coolability u The core of a nuclear reactor converts atomic binding energy to heat u Heat is converted to electrical energy by means of (typically) steam turbines u Inter alia, we need a means to convert heat into steam – thermal hydraulics u Efficiently (higher temperatures, improved heat transfer,…) u Reliably (backup systems, natural circulation,…) u Maintaining the extraction of heat from the core is vital – coolability u Normal operation u Scheduled or unscheduled transients u Accident scenarios u Failure to maintain coolability is catastrophic u Chernobyl, Fukushima, etc u Regulator requires coolability to be established in all design scenarios Role of TH in nuclear: Normal Operation By U.S.NRC. - http://www.nrc.gov/reading- rm/basic-ref/students/animated-pwr.html, Public Domain u Ty p i c al P W R u Two l o o p d e s i g n u Primary loop u Pressurized to 15 Mpa u Tin ~ 550K, Tout ~590K u Power generation limited by ability to remove core heat Issues in fission: Accident scenarios u Coolability may be lost due to: u Pump or power failure u Structural failure (blockages) u Loss of coolant accident (LOCA) u Anything else anyone can think of…. -
Assessment of Critical Heat Flux Correlations in Narrow Rectangular Channels Alberto Ghione, Brigitte Noel, Paolo Vinai, Christophe Demazière
Assessment of Critical Heat Flux correlations in narrow rectangular channels Alberto Ghione, Brigitte Noel, Paolo Vinai, Christophe Demazière To cite this version: Alberto Ghione, Brigitte Noel, Paolo Vinai, Christophe Demazière. Assessment of Critical Heat Flux correlations in narrow rectangular channels. NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Oct 2016, Gyeongju, South Korea. hal-02124694 HAL Id: hal-02124694 https://hal.archives-ouvertes.fr/hal-02124694 Submitted on 9 May 2019 HAL is a multi-disciplinary open access L’archive ouverte pluridisciplinaire HAL, est archive for the deposit and dissemination of sci- destinée au dépôt et à la diffusion de documents entific research documents, whether they are pub- scientifiques de niveau recherche, publiés ou non, lished or not. The documents may come from émanant des établissements d’enseignement et de teaching and research institutions in France or recherche français ou étrangers, des laboratoires abroad, or from public or private research centers. publics ou privés. NUTHOS-11: The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety Gyeongju, Korea, October 9-13, 2016 . N11P0387 Assessment of Critical Heat Flux correlations in narrow rectangular channels Alberto Ghione (a) , Brigitte Noel Commissariat à l’Énergie Atomique et aux énergies alternatives, CEA DEN/DM2S/STMF/LATF; 17 rue des Martyrs, Grenoble, France [email protected], [email protected] Paolo Vinai, Christophe Demazière (a) Chalmers University of Technology, Division of Subatomic and Plasma Physics Department of Physics; Gothenburg, Sweden [email protected], [email protected] ABSTRACT The aim of the work is to assess different CHF correlations when applied to vertical narrow rectangular channels with upward low-pressure water flow. -
Heat and Mass Transfer: Fundamentals and Applications 5Th Edition Download Free
HEAT AND MASS TRANSFER: FUNDAMENTALS AND APPLICATIONS 5TH EDITION DOWNLOAD FREE Yunus A Cengel | 9780077654764 | | | | | ISBN 13: 9780073398181 Noting that the heat transfer area in this case is A 2rL and the thermal conductivity is constant, the one-dimensional transient heat conduction equation in a cylinder becomes 1 r r. The density of the cylinder is and the specific heat is c. The topic of critical heat flux is covered in Chap ter 10 Boiling and Condensation. Taking the positive direction of x to be from the inner surface to the outer surface, the correct Heat and Mass Transfer: Fundamentals and Applications 5th edition for the convection boundary condition is a k dT 0 dx Heat and Mass Transfer: Fundamentals and Applications 5th edition k b k h T 0 T dT 0 dx Answer a k 1 1 T hT 1 1 dT L dx 2 d k dT L dx T L T h 2 2 Heat and Mass Transfer: Fundamentals and Applications 5th edition h T 2 1 e None of them 2 dT 0 h T1 0 T 1 dx Consider steady one-dimensional heat conduction through a plane wall, a cylindrical shell, and a spherical shell of uniform Heat and Mass Transfer: Fundamentals and Applications 5th edition with constant thermophysical properties and no thermal energy generation. I have taken their services earlier for textbook solutions which helped me to score well. Properties The Heat and Mass Transfer: Fundamentals and Applications 5th edition surface has an absorptivity of 0. The total rate of heat transfer from the black ball is a W. -
Henryk Anglart
Thermal- Hydraulics in Nuclear Systems Henryk Anglart Thermal-Hydraulics in Nuclear Systems 2010 Henryk Anglart All rights reserved i his textbook is intended to be an introduction to selected thermal-hydraulic topics for students of energy engineering and applied sciences as well as for professionals working in the nuclear engineering field. The basic aspects of T thermal-hydraulics in nuclear systems are presented with a goal to demonstrate how to solve practical problems. This ‘hands-on’ approach is supported with numerous examples and exercises provided throughout the book. In addition, the book is accompanied with computational software, I C O N K E Y implemented in the Scilab ( www.scilab.org ) environment. The software is available for download at Note Corner www.reactor.sci.kth.se/downloads and is shortly described Examples in Appendices C and D. Computer Program More Reading The textbook is organized into five chapters and each of them is divided into several sections. Parts in the book of special interest are designed with icons, as indicated in the table to the left. The “Note Corner” icon indicates a section with additional relevant information, not directly related to the topics covered by the book, but which could be of interest to the reader. All examples are marked with a pen icon. Special icons are also used to mark sections with computer programs and with suggested more reading. The first chapter is concerned with various introductory topics in nuclear reactor thermal-hydraulics. The second chapter deals with the rudiments of thermodynamics, at the level which is necessary to understand the material presented in Chapter 3 (fluid mechanics) and Chapter 4 (heat transfer). -
Simulation of a High Speed Counting System for Sic Neutron Sensors
Transactions of the Korean Nuclear Society Autumn Meeting Gyeongju, Korea, October 27-28, 2016 Preliminary Study of Printed Circuit Heat Exchanger (PCHE) for various power conversion systems for SMART Jinsu Kwon, Seungjoon Baik, Jeong Ik Lee* Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu,Daejeon 305-701, Republic of Korea *Corresponding author: [email protected] 1. Introduction SMART (System-integrated Modular Advanced ReacTor) is a promising advanced small nuclear power reactor. It is a 330 MWth integral type reactor developed by KAERI (Korea Atomic Energy Institute) for multipurpose utilization, which incorporated inherent safety systems, system simplification and component modularization. The steam-Rankine cycle was the most widely used power conversion system for a nuclear power plant. The size of the heat exchanger is important for the modulation. Such a challenge was conducted by Kang et al [1]. They change the steam generator type for the SMART from helical type heat exchanger to Printed Circuit Heat Exchanger (PCHE). Figure 1. PCHE Plates and Diffusion Bond [2] Recently, there has been a growing interest in the supercritical carbon dioxide (S-CO2) Brayton cycle as the most promising power conversion system. The reason is high efficiency with simple layout and compact power plant due to small turbomachinery and compact heat exchanger technology. That is why the S- CO2 Brayton cycle can enhance the existing advantages of Small Modular Reactor (SMR) like SMART, such as reduction in size, capital cost, and construction period. The objective of this paper is comparing the size of PCHE in case of steam Rankine cycle and in case of the S-CO2 Brayton cycle. -
Thermal Hydraulics of Water Cooled Divertors
GA–A23476 THERMAL HYDRAULICS OF WATER COOLED DIVERTORS by C.B. BAXI OCTOBER 2000 DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. GA–A23476 THERMAL HYDRAULICS OF WATER COOLED DIVERTORS by C.B. BAXI This is a preprint of a paper presented at the 21st Symposium on Fusion Technology, September 11–15, 2000 in Madrid, Spain and to be published in Fusion Design and Engineering. Work supported by General Atomics IR&D Funds GA PROJECT 4437 OCTOBER 2000 C.B. BAXI THERMAL HYDRAULICS OF WATER COOLED DIVERTORS ABSTRACT Divertors of several new machines, such as JT-60SU, FIRE, SST-1, ITER-RC and KSTAR, are water-cooled. This paper examines critical thermal hydraulic issues associated with design of such divertors. The flow direction of coolant in the divertor can be either toroidal or poloidal. A quantitative evaluation shows that the poloidal flow direction is preferred, because the flow rate and pumping power are about an order of magnitude smaller compared to the toroidal flow direction. -
Advanced High Temperature Reactor Thermal Hydraulics Analysis and Salt Clean-Up System Description
ORNL/TM-2014/499 Advanced High Temperature Reactor Thermal Hydraulics Analysis and Salt Clean-up System Description Compiled by G. L. Yoder Approved for public release: Contributors distribution is unlimited. A. T. Bopp D. E. Holcomb W. D. Pointer D. Wang September 2014 DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via US Department of Energy (DOE) SciTech Connect. Website http://www.osti.gov/scitech/ Reports produced before January 1, 1996, may be purchased by members of the public from the following source: National Technical Information Service 5285 Port Royal Road Springfield, VA 22161 Telephone 703-605-6000 (1-800-553-6847) TDD 703-487-4639 Fax 703-605-6900 E-mail [email protected] Website http://www.ntis.gov/help/ordermethods.aspx Reports are available to DOE employees, DOE contractors, Energy Technology Data Exchange representatives, and International Nuclear Information System representatives from the following source: Office of Scientific and Technical Information PO Box 62 Oak Ridge, TN 37831 Telephone 865-576-8401 Fax 865-576-5728 E-mail [email protected] Website http://www.osti.gov/contact.html This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof.