DOCSLIB.ORG

Scaling Analysis for the Direct Reactor Auxiliary Cooling System for Ahtrs

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Scaling Analysis for the Direct Reactor Auxiliary Cooling System for Ahtrs Scaling Analysis for the Direct Reactor Auxiliary Cooling System for AHTRs THESIS Presented in Partial Fulfillment of the Requirements for the Degree Master of Science in the Graduate School of The Ohio State University By Qiuping Lv, B.S. Graduate Program in Nuclear Engineering The Ohio State University 2013 Master's Examination Committee: Prof. Xiaodong Sun, Advisor Prof. Thomas E. Blue Prof. Richard N. Christensen Copyright by Qiuping Lv 2012 Abstract The Advanced High Temperature Reactor (AHTR) or Fluoride Salt Cooled High Temperature Reactor (FHR) is one of the advanced rector concepts that have been proposed for Gen IV reactors. The AHTR combines four main proven nuclear technologies, namely, the liquid salt of molten salt reactors, the coated particle fuel (TRISO particle) of high-temperature gas-cooled reactors, the pool configuration and passive safety system of sodium-cooled fast reactors, and the Brayton power cycle technology. The AHTR is capable of providing very high temperature (750 to 1,000ºC) heat for various industrial processing needs, hydrogen production, and electricity generation. The Direct Reactor Auxiliary Cooling System (DRACS) is a passive heat removal system that was derived from the Experimental Breeder Rector-II (EBR-II), and then improved in later fast reactor designs. The DRACS has been proposed for AHTR as the passive decay heat removal system. The DRACS features three coupled natural circulation/convection loops relying completely on buoyancy as the driving force. In the DRACS, two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX) are used to couple these natural circulation/convection loops. In addition, a fluidic diode is employed to restrict parasitic flow during normal operation of the reactor and to activate the DRACS in accidents. ii While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for AHTRs built and tested in the literature. In this report, a detailed modular design of the DRACS for a 20-MWth FHR is first developed. As a starting point, the DRACS is designed to remove 1% of the nominal power, i.e., the decay power being 200 kW. The design process for the prototypic DRACS involves selection of the salts, identification of the reactor core, design of the DHX and NDHX, design of the fluidic diode, design of the air chimney, selection of the loop pipes, and finally determination of the loop height based on pressure drop analysis. FLiBe with high enrichment in Li-7 and FLiNaK have been selected as the primary and secondary salts, respectively. A 16-MWth pebble bed core proposed by University of California at Berkeley (UCB) is adopted in the design. Shell-and-tube heat exchangers have been designed based on Delaware Method for the DHX and NDHX. A vortex diode that has been tested with water in the literature is adopted in the present design. Finally, pipes with inner diameter of 15 cm are selected for both the primary and secondary loops of the DRACS. The final DRACS design features a total height less than 13 m. The design presented here has the potential to be used in the planned small-scale FHR test reactor and will also benefit and guide the DRACS design for a commercial AHTR. Following the prototypic DRACS design is the detailed scaling analysis for the DRACS, which will provide guidance for the design of scaled-down DRACS test facilities. Based on the Boussinesq assumption and one-dimensional formulation, the governing equations, i.e., the continuity, integral momentum, and energy equations are non-dimensionalized by introducing appropriate dimensionless parameters, including the iii dimensionless length, temperature, velocity, etc. The key dimensionless numbers, i.e., the Richardson, friction, Stanton, time ratio, Biot, and heat source numbers that characterize the DRACS system, are obtained from the non-dimensional governing equations. Based on the dimensionless numbers and non-dimensional governing equations, similarity laws are proposed. In addition, a scaling methodology has also been developed, which consists of the core scaling and loop scaling. Due to the importance of the core heat transfer in establishing the DRACS steady state, core scaling is started with, from which the convection time ratio is obtained. The loop scaling is accomplished by utilizing the convection time ratio obtained from the core scaling and by making two assumptions that are related to the power and loop height of the test facility. The consistence between the core and the loop scaling is examined through the reference volume ratio that can be obtained from both scaling processes. The scaling methodology and similarity laws have been applied to obtain a scaled-down low-temperature DRACS test facility (LTDF) and a scaled-down high-temperature DRACS test facility (HTDF). iv Dedication This document is dedicated to my family. v Acknowledgments There are many people I want to thank. First and foremost, I must give my deepest gratitude to my advisor, Prof. Xiaodong Sun, for his guidance, supervision, attention and support throughout my study and research. In addition, I sincerely thank Prof. Richard Christensen and Prof. Thomas Blue for their constructive comments and suggestions during all the research meetings. Helps from Dr. Xia Wang, Dr. Grady Yoder and Dr. Dane Wilson (ORNL), and Dr. Piyush Sabharwall (INL) are greatly appreciated. I also would like to thank my friends and the colleagues in the Ohio State University Nuclear Engineering Program. They have been helpful in improving my English. Special thanks to my wife and my parents for always being understanding and supportive. Finally, the support from the U.S. Department of Energy for our DRACS project is gratefully acknowledged. vi Vita 2009................................................................B.S. in Physics, Nanjing University, China 2009-2010 ......................................................Teaching Assistant, Physics Department, Iowa State University 2010-2011 ......................................................Graduate Fellow, Nuclear Engineering Program, The Ohio State University 2011 to present ..............................................Graduate Research Associate, Nuclear Engineering Program, The Ohio State University Publications 1. X. Wang, Q. Lv, X. Sun, R.N. Christensen, T.E. Blue, G. Yoder, D. Wilson, and P. Sabharwall, “A Modular Design of a Direct Reactor Auxiliary Cooling System for AHTRs,” Transaction of the American Nuclear Society, Vol. 104, 2011 American Nuclear Society Annual Meeting, June 26-30, 2011, Hollywood, FL, pp. 1077-1080. 2. X. Wang, Q. Lv, X. Sun, R.N. Christensen, T.E. Blue, G. Yoder, D. Wilson, and P. Sabharwall, “Scaling Analysis for the Direct Reactor Auxiliary Cooling System for AHTRs,” Transaction of the American Nuclear Society, Vol. 105, 2011 American Nuclear Society Winter Meeting, October 30 – November 3, 2011, Washington, DC, pp. 1027-1030. 3. X. Wang, Q. Lv, X. Sun, R.N. Christensen, T.E. Blue, G. Yoder, D. Wilson, and P. Sabharwall, “Design of a Scaled-down DRACS Test Facility for an AHTR,” Transaction of the American Nuclear Society, Vol. 105, 2011 American Nuclear Society Winter Meeting, October 30 – November 3, 2011, Washington, DC, pp. 1031-1034. vii 4. Q. Lv, X. Wang, I. Adams, X. Sun, R.N. Christensen, T.E. Blue, G. Yoder, D. Wilson, and P. Sabharwall, “Design of a Scaled-down Low-temperature DRACS Test Facility for an AHTR,” Transactions of the American Nuclear Society, Vol. 106, 2012 American Nuclear Society Annual Meeting, June 24-28, 2012, Chicago, IL, pp. 1071-1074. 5. Q. Lv, I. Adams, X. Wang, X. Sun, R.N. Christensen, T.E. Blue, G. Yoder, D. Wilson, and P. Sabharwall, “A MATLAB Code for Thermal Performance Evaluation of a Low- Temperature DRACS Test Facility”, Transaction of the American Nuclear Society, Vol. 107, 2012 American Nuclear Society Winter Meeting, November 11-15, 2012, San Diego, CA, pp. 1374-1377. Fields of Study Major Field: Nuclear Engineering viii Table of Contents Abstract ............................................................................................................................... ii Dedication ........................................................................................................................... v Acknowledgments.............................................................................................................. vi Vita .................................................................................................................................... vii Table of Contents ............................................................................................................... ix List of Tables .................................................................................................................... xii List of Figures .................................................................................................................. xiv Nomenclature ................................................................................................................... xvi Chapter 1: Introduction ....................................................................................................... 1 1.1 Gen IV Reactors ........................................................................................................ 1 1.2 Molten Salt Reactor (MSR) ...................................................................................... 3 1.3 Advanced High Temperature Reactor (AHTR) .......................................................
Load more

Recommended publications
  • Molten Salt Chemistry Workshop
    The cover depicts the chemical and physical complexity of the various species and interfaces within a molten salt reactor. To advance new approaches to molten salt technology development, it is necessary to understand and predict the chemical and physical properties of molten salts under extreme environments; understand their ability to coordinate fissile materials, fertile materials, and fission products; and understand their interfacial reactions with the reactor materials. Modern x-ray and neutron scattering tools and spectroscopy and electrochemical methods can be coupled with advanced computational modeling tools using high performance computing to provide new insights and predictive understanding of the structure, dynamics, and properties of molten salts over a broad range of length and time scales needed for phenomenological understanding. The actual image is a snapshot from an ab initio molecular dynamics simulation of graphene- organic electrolyte interactions. Image courtesy of Bobby G. Sumpter of ORNL. Molten Salt Chemistry Workshop Report for the US Department of Energy, Office of Nuclear Energy Workshop Molten Salt Chemistry Workshop Technology and Applied R&D Needs for Molten Salt Chemistry April 10–12, 2017 Oak Ridge National Laboratory Co-chairs: David F. Williams, Oak Ridge National Laboratory Phillip F. Britt, Oak Ridge National Laboratory Working Group Co-chairs Working Group 1: Physical Chemistry and Salt Properties Alexa Navrotsky, University of California–Davis Mark Williamson, Argonne National Laboratory Working
    [Show full text]
  • Liquid Fluoride Salt Experiment Using a Small Natural Circulation Cell
    ORNL/TM-2014/56 Liquid Fluoride Salt Experiment Using a Small Natural Circulation Cell Graydon L. Yoder, Jr. Dennis Heatherly David Williams Oak Ridge National Laboratory Josip Caja Mario Caja Approved for public release; Electrochemical Systems, Inc., distribution is unlimited. Yousri Elkassabgi John Jordan Roberto Salinas Texas A&M University April 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.
    [Show full text]
  • Thermal Properties of Licl-Kcl Molten Salt for Nuclear Waste Separation
    Project No. 09-780 Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste Separation Fue l C yc le R&D Dr. Kumar Sridharan University of Wisconsin, Madison In collaboration with: Idaho National Laboratory Michael Simpson, Technical POC James Bresee, Federal POC Thermal Properties of LiCl -KCl Molten Salt for Nuclear Waste Separation Project Investigators: Dr. Kumar Sridharan, Dr. Todd Allen, Dr. Mark Anderson (University of Wisconsin) and Dr. Mike Simpson (Idaho National Laboratory) Post-Doctoral Research Associates: Dr. Luke Olson Graduate Students: Mr. Mehran Mohammadian and Mr. Sean Martin Undergradua te Students: Mr. Jacob Sager and Mr. Aidan Boyle University of Wisconsin -Madison NEUP Final Report Project Contact: Dr. Kumar Sridharan [email protected] Date: November 30th, 2012 Table of Contents 1. Introductory Narrative ............................................................................................................................. 1 2. Motivation ................................................................................................................................................ 2 3. Experimental Setup ................................................................................................................................... 3 3.1 Molten Salt Electrochemistry.............................................................................................................. 3 3.1.1 LiCl-KCl Eutectic Salt ................................................................................................................
    [Show full text]
  • Fissile, Fertile and Dual-Use Structural Materials Involved in Nuclear Reactors
    ARTICLE NUCLEAR MATERIALS – FISSILE, FERTILE AND DUAL-USE STRUCTURAL MATERIALS INVOLVED IN NUCLEAR REACTORS N. R. DAS* The article presents a brief account of nuclear materials, with special emphasis on fissile, fertile and some important dual-use structural materials generally involved in nuclear reactors. The dual- use structural materials utilized in nuclear reactors have got important applications in both nuclear and non-nuclear fields. In the hostile environment, the important phenomena such as interactions between fission products and the surrounding elemental species, radiation-induced effects, corrosion, generation of gases, swelling and so forth, become increasingly complex and the performance of a nuclear reactor system thus becomes very much dependent on the physicochemical stability and nuclear compatibility of the dual-use structural materials used in fuel sub-assembly towards the fuel elements. In this context, the dual-use structural materials like stainless steel, zirconium alloys, etc. as claddings; water, liquid sodium or gases, etc. as coolants and water, boron, etc. as moderators, having good reliability and appropriate nuclear compatibility with the fuels, are of prime importance in reactor technology. In advanced designed reactors, development of novel fuels coupled with efficient dual-use structural materials may mitigate the challenges involved in optimizing the efficiency of the power reactors under specific experimental conditions. Introduction presented as 14N(, p)17O. In 1932, J. Chadwick discovered the fundamental particle, neutron, by bombardment of boron ince the discovery of X-rays by Wilhelm C. with alpha particle through the nuclear reaction,10B( , Roentgen and ‘Becquerel rays’ or ‘Uranic rays’ by n)13N. Subsequently, Irene and Frederic Joliot Curie in Henri Becquerel in the last decade of the nineteenth S 1934 discovered artificial radioactivity and Otto Hahn and century, conscientious activities in nuclear science have F.
    [Show full text]
  • Engineering Database of Liquid Salt Thermophysical and Thermochemical Properties
    INL/EXT-10-18297 Rev. 1 Engineering Database of Liquid Salt Thermophysical and Thermochemical Properties Manohar S. Sohal Matthias A. Ebner Piyush Sabharwall Phil Sharpe June 2013 The INL is a U.S. Department of Energy National Laboratory operated by Battelle Energy Alliance INL/EXT-10-18297 Rev. 1 Engineering Database of Liquid Salt Thermophysical and Thermochemical Properties Manohar S. Sohal Matthias A. Ebner Piyush Sabharwall Phil Sharpe June 2013 Idaho National Laboratory Idaho Falls, Idaho 83415 http://www.inl.gov Prepared for the U.S. Department of Energy Office of Nuclear Energy Under DOE Idaho Operations Office Contract DE-AC07-05ID14517 DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed 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. References herein to any specific commercial product, process, or service by trade name, trade mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof. ABSTRACT The purpose of this report is to provide a review of thermophysical properties and thermochemical characteristics of candidate molten salt coolants, which may be used as a primary coolant within a nuclear reactor or heat transport medium from the Very High Temperature Reactor (VHTR) to a processing plant; for example, a hydrogen-production plant.
    [Show full text]
  • Corrosion in Very High-Temperature Molten Salt for Next Generation CSP Systems
    Corrosion in Very High-Temperature Molten Salt for Next Generation CSP Systems Brenda Garcia Diaz (PI), Josh Gray (Co-PI), Luke Olson, Michael Martinez-Rodriguez, Roderick Fuentes (SRNL) Ramana Reddy, John Van Zee (University of Alabama) Project Identifier: Garcia-Diaz_A Project Start: Q1 - FY2013 1 April 1st , 2013 Outline • Motivation • Innovation • Goals and Objectives • Background • Molten Salt and Material Selection • Molten Salt Corrosion Review • Current status • Corrosion measurements/Inhibitor studies • Thermodynamic modeling • Electrochemical Kinetics and CFD Corrosion Modeling • Future plans 2 April 1st , 2013 Motivation: Enable efficient, long-term high temperature heat transfer using molten salts • SunShot program seeks major price reductions • Higher temperature energy conversion allows for better thermodynamic efficiencies • Temp. Range for Study: 800 – 1000 °C • Limited number of options for very high temperature heat transfer – Steam or other gases • Large pipe diameters needed • Large pumping power required – Molten Glass • Viscous • Low thermal conductivities – Liquid metals • Limited operating temperature range • Alloying with piping materials • Vapor pressures – Molten Salts • Heat transfer behavior at 700 °C similar to water at 20 °C Image Source: NREL, available: http://earthobservatory.nasa.gov/Features/RenewableEnergy/Images/solar_two.jpg Table Source: SAND2011-2419 3 April 1st , 2013 Goals and Objectives • Goal #1 – Identify molten salt / material combinations that have long service life and meet performance
    [Show full text]
  • Fluoride-Salt-Cooled High Temperature Reactor (FHR) Materials, Fuels and Components White Paper
    FHR Materials, Fuels and Components White Paper Integrated Research Project Workshop 3 Fluoride-Salt-Cooled High Temperature Reactor (FHR) Materials, Fuels and Components White Paper Panel Members Todd Allen (University of Wisconsin, Madison) Gordon Kohse (Massachusetts Institute of Technology) Mark Anderson (University of Wisconsin, Madison) Regis Matzie (ret. Westinghouse Electric Company) Ron Ballinger (Massachusetts Institute of Technology) Matthew Memmott (Westinghouse Electric Company) Tim Burchell (Oak Ridge National Laboratory) Jim Nestell (MPR Associates) Edward Blandford (University of New Mexico) Michael Patterson (Idaho National Laboratory) Denis Clark (Idaho National Laboratory) Scott Penfield (Technology Insights) William Corwin (Department of Energy) Per Peterson (University of California, Berkeley) George Flanagan (Oak Ridge National Laboratory) Sam Sham (Oak Ridge National Laboratory) Charles Forsberg (Massachusetts Institute of Technology) Lance Snead (Oak Ridge National Laboratory) Peter Hosemann (University of California, Berkeley) Kumar Sridharan (University of Wisconsin, Madison) David Holcomb (Oak Ridge National Laboratory) Carl Stoots (Idaho National Laboratory) Lin-Wen Hu (Massachusetts Institute of Technology) Richard Vollmer (Westinghouse Electric Company) John Hunn (Oak Ridge National Laboratory) David Williams (Oak Ridge National Laboratory) Victor Ignatiev (Kurchatov Institute) Dane Wilson (Oak Ridge National Laboratory) UW Madison Facilitators MIT Facilitators UC Berkeley Facilitators Guoping Cao David Carpenter
    [Show full text]
  • Experimental and Computational Study of Static Solidification of Molten Fluoride Salts for Reactor Coolant Application
    EXPERIMENTAL AND COMPUTATIONAL STUDY OF STATIC SOLIDIFICATION OF MOLTEN FLUORIDE SALTS FOR REACTOR COOLANT APPLICATION by Louis J. Chapdelaine A thesis submitted in partial fulfillment of the requirements for the degree of Master of Science (Nuclear Engineering and Engineering Physics) at the UNIVERSITY OF WISCONSIN-MADISON 2017 i 1 Abstract A series of experiments were conducted to measure various FLiBe and FLiNaK thermophysical properties, quantify supercooling behavior, and provide solidification modeling validation data. Both fluoride salt purities and compositions were characterized using ICP-MS elemental analysis. DSC analysis was conducted to measure specific heat capacity, latent heat of fusion, and solidification temperature for both FLiBe and FLiNaK using ASTM standard methods. The specific heat capacity measurement was unsuccessful based on the non- physical negative heat flow measurements. The latent heat measurements for FLiBe and FLiNaK were 424.7±6.2 kJ/kg and 412.5±3.3 kJ/kg. The solidification temperature measurements for FLiBe and FLiNaK were 460.8±0.1°C and 465.9±0.1°C respectively. The Density Measurement Experiment was conducted to measure the temperature-dependent density of molten FLiBe and FLiNaK using the Archimedean principle of buoyancy. The density correlation for FLiBe from solidification (~460°C) to 700°C was determined to be 휌 = 2241.6 − 0.42938 ∗ 푇 in kg/m3 with an error of ±0.22%. The density correlation for FLiNaK from solidification (~460°C) to 700°C was determined to be 휌 = 2485.8 − 0.69451 ∗ 푇 in kg/m3 with an error of ±2.7%. The Static Solidification Experiment was designed to characterize FLiBe and FLiNaK supercooling behavior, measure the solidification front propagation, and collect solidification model validation data.
    [Show full text]
  • Fluoride Salt Coolant Properties for Nuclear Reactor Applications: a Review
    Fluoride salt coolant properties for nuclear reactor applications: A review The MIT Faculty has made this article openly available. Please share how this access benefits you. Your story matters. Citation Romatoski, R. R. and L.W. HU. "Fluoride salt coolant properties for nuclear reactor applications: A review." Annals of Nuclear Energy 109 (November 2017): 635-647 As Published http://dx.doi.org/10.1016/j.anucene.2017.05.036 Publisher Elsevier BV Version Author's final manuscript Citable link https://hdl.handle.net/1721.1/123988 Terms of Use Creative Commons Attribution-NonCommercial-NoDerivs License Detailed Terms http://creativecommons.org/licenses/by-nc-nd/4.0/ Fluoride salt coolant properties for nuclear reactor applications: A review R.R. Romatoskia and L.W. Hua,* a MASSACHUSETTS INSTITUTE OF TECHNOLOGY, NW12-220, 77 MASSACHUSETTS AVE., CAMBRIDGE, MA 02139 * Corresponding author E-mail addresses: [email protected] (R.R. Romatoski), [email protected] (L.W. Hu) ABSTRACT Fluoride salts are crucial to achieving the benefits of Fluoride-salt-cooled High-temperature Reactors (FHRs). Intensive studies and modeling are being performed in different countries aimed at FHR technology development. Better understanding of liquid fluoride salt coolant properties and their uncertainties are needed for design and analysis of nuclear facilities. The main objective of the present study is to perform a literature survey of the experimental data, numerical studies, reports, and other review compilations for the main thermophysical properties of liquid fluoride salts. The review recommends density, heat capacity, thermal conductivity, and viscosity properties for potential primary coolants LiF-BeF2 (flibe) and NaF-ZrF4 (nafzirf) and secondary coolant LiF-NaF-KF (flinak).
    [Show full text]
  • The LM-LS Experiment: Investigating Corrosion Control, in Liquid Fluoride Salts, by Liquid Alkali Metal
    Blandine Laurenty in completion of the requirements for the Master of Science degree in Nuclear Engineering of the University of California Berkeley The LM-LS experiment: investigating corrosion control, in Liquid Fluoride Salts, by Liquid alkali Metal Report UCBTH-06-002 Research Advisor: Professor Per F. Peterson Academic Advisor: Professor Brian D. Wirth January 2006 1 Introduction The conceptual association of liquid fluoride salts and nuclear power is as old as the civilian use of nuclear energy: in 1947 the Aircraft Reactor Experiment (ARE) development program proposed their use as a reactor fuel and coolant. However, in the 1970s, the Atomic Energy Commission abandoned the promising Molten Salt Breeder Reactor (MSBR) project at Oak Ridge National Laboratory to focus breeder reactor research efforts on the Liquid Metal Breeder (LMBR) concept. The molten salts’ attractive features were quickly forgotten, and they today are more famous for the severe corrosion issues which their use entails under certain conditions. Composed of the most electronegative element (fluorine), and combinations of some very electropositive elements (lithium, beryllium, sodium, potassium, rubidium …) the salts involved are extremely stable compounds. The pure salts do not have the strong corroding power of water, because the products of salt corrosion (metallic fluorides), are less stable than the beryllium and alkali fluorides of the salt, whereas metallic oxides from water corrosion can be very stable. Yet, purity is not a realistic state of operation since contaminants can be introduced in the system via a variety of means including an initially impure salt charge, fluxing of the oxide film that the piping will invariably have, and ingress of impurities such as moisture or air due to a failure in the salt containment.
    [Show full text]
  • HSC Chemistry References
    HSC Chemistry References Some publications where HSC Chemistry has been utilized SOURCES: http://scholar.google.com Recommended reference format to HSC: http://www.scopus.com A. Roine. HSC Chemistry® [Software], Metso Outotec, Pori 2021. Software available at www.mogroup.com/hsc Year Authors Title Source 2021 Chen, M., Avarmaa, K., Handling trace elements in WEEE recycling through Minerals Engineering, 173, art. No. 107189 Taskinen, P., Klemettinen, copper smelting - an experimental and L., Michallik, R., O'Brien, thermodynamic study H., Jokilaakso, A. 2021 Avarmaa, K., Taskinen, P., Ni–Fe–Co alloy – magnesia-iron-silicate slag Journal of Materials Research and Technology, 15, Klemettinen, L., O'Brien, equilibria and the behavior of minor elements Cu and pp. 719-730. H., Lindberg, D. P in nickel slag cleaning 2021 Teixeira, L.T., Braz, W.F., Sulfated and carboxylated nanocellulose for Co+2 Journal of Materials Research and Technology, 15, Correia de Siqueira, R.N., adsorption pp. 434-447. Pandoli, O.G., Geraldes, M.C. 2021 Makuza, B., Yu, D., Huang, Dry Grinding - Carbonated Ultrasound-Assisted Resources, Conservation and Recycling, 174, art. no. Z., Tian, Q., Guo, X. Water Leaching of Carbothermally Reduced Lithium- 105784 Ion Battery Black Mass Towards Enhanced Selective Extraction of Lithium and Recovery of High-Value Metals 2021 StaniþLü, I., Cañete Vela, I., Fate of lead, copper, zinc and antimony during Fuel, 302, art. no. 121147 Backman, R., Maric, J., chemical looping gasification of automotive shredder Cao, Y., Mattisson, T. residue 2021 Monteverde, F., Saraga, Compositional pathways and anisotropic thermal Journal of the European Ceramic Society, 41 (13), pp. F., Gaboardi, M., Plaisier, expansion of high-entropy transition metal diborides 6255-6266.
    [Show full text]
  • Influence of Molten Salt (Flinak) Thermophysical Properties on a Heated Tube Using Cfd Rans Turbulence Modeling of an Experimental Testbed
    INFLUENCE OF MOLTEN SALT (FLINAK) THERMOPHYSICAL PROPERTIES ON A HEATED TUBE USING CFD RANS TURBULENCE MODELING OF AN EXPERIMENTAL TESTBED. A Thesis by RAMIRO FREILE Submitted to the Office of Graduate and Professional Studies of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE Chair of Committee, Mark L. Kimber Committee Members, Pavel V. Tsvetkov Michael Pate Head of Department, Michael Nastasi August 2020 Major Subject: Nuclear Engineering Copyright 2020 Ramiro Freile ABSTRACT In a liquid fuel molten salt reactor key factor to consider upon its design is the strong coupling between different physics present such as neutronics, thermo-mechanics and thermal-hydraulics. Focusing on the thermal-hydraulics aspect, in support of potential reactor designs, it is required that the heat transfer is well characterized. For this purpose, turbulence models used for FLiNaK flow must be valid, and its thermophysical properties must be accurately described. In the literature, several expressions for each material property can be found, with differences that can be quite significant. The goal of this study is to demonstrate and quantify the impact that the uncertainty in ther- mophysical properties has on key metrics of thermal hydraulic importance for MSRs, in particular on the heat transfer coefficient. In order to achieve this, CFD simulations using the RANS k- ! SST model were compared to published experiment data on molten salt. Various correlations for FLiNaK’s thermophysical properties were used. It was observed that the spread in FLiNaK’s thermophysical properties lead to a significant variance in the heat coefficient. Motivated by this, additional CFD simulations were done to obtain sensitivity coefficients for each thermophysical property.
    [Show full text]