Liquid Fluoride Salt Experiment Using a Small Natural Circulation Cell
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Molten-Salt Technology and Fission Product Handling
Molten-Salt Technology and Fission Product Handling Kirk Sorensen Flibe Energy, Inc. ORNL MSR Workshop October 4, 2018 2018-10-16 Hello, my name is Kirk Sorensen and I’d like to talk with you today about fission products and their handling in molten-salt reactors. One of the things that initially attracted me to molten-salt reactor technology was the array of options that it gave for the intelligent handling of fission products. It represented such a contrast to solid-fueled systems, which mixed fission products in with unburned nuclear fuel in a form that was difficult to separate, one from another. While my focus will be on our work on molten-salt reactor fission product handling, many of the principles are general to molten-salt reactors as a whole. Fundamental Nuclear Reactor Concept In its simplest form, a nuclear reactor generates thermal energy that is carried away by a coolant. That coolant heats the working fluid of a power conversion system, which generates electricity from part of the thermal energy and rejects the remainder to the environment. coolant working fluid fresh fuel electricity Power Nuclear Heat Conversion Reactor Exchanger System spent fuel heated water or air coolant working fluid The primary coolant chosen for a nuclear reactor determines, in large part, its size and manufacturability. The temperature of the coolant determines the efficiency of electrical generation. Fundamental Nuclear Reactor Concept In its simplest form, a nuclear reactor generates thermal energy that is carried away by a coolant. That coolant heats the working fluid of a power conversion system, which generates electricity from part of the thermal energy and rejects the remainder to the environment. -
A Comparison of Advanced Nuclear Technologies
A COMPARISON OF ADVANCED NUCLEAR TECHNOLOGIES Andrew C. Kadak, Ph.D MARCH 2017 B | CHAPTER NAME ABOUT THE CENTER ON GLOBAL ENERGY POLICY The Center on Global Energy Policy provides independent, balanced, data-driven analysis to help policymakers navigate the complex world of energy. We approach energy as an economic, security, and environmental concern. And we draw on the resources of a world-class institution, faculty with real-world experience, and a location in the world’s finance and media capital. Visit us at energypolicy.columbia.edu facebook.com/ColumbiaUEnergy twitter.com/ColumbiaUEnergy ABOUT THE SCHOOL OF INTERNATIONAL AND PUBLIC AFFAIRS SIPA’s mission is to empower people to serve the global public interest. Our goal is to foster economic growth, sustainable development, social progress, and democratic governance by educating public policy professionals, producing policy-related research, and conveying the results to the world. Based in New York City, with a student body that is 50 percent international and educational partners in cities around the world, SIPA is the most global of public policy schools. For more information, please visit www.sipa.columbia.edu A COMPARISON OF ADVANCED NUCLEAR TECHNOLOGIES Andrew C. Kadak, Ph.D* MARCH 2017 *Andrew C. Kadak is the former president of Yankee Atomic Electric Company and professor of the practice at the Massachusetts Institute of Technology. He continues to consult on nuclear operations, advanced nuclear power plants, and policy and regulatory matters in the United States. He also serves on senior nuclear safety oversight boards in China. He is a graduate of MIT from the Nuclear Science and Engineering Department. -
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 -
0409-TOFE-Elguebaly
BBenenefitsefits ooff RRadadialial BBuuildild MinMinimimizatioizationn anandd RReqequuirirememenentsts ImImposedposed onon AARRIEIESS CComompapacctt SStetellallararatotorr DDeesigsignn Laila El-Guebaly (UW), R. Raffray (UCSD), S. Malang (Germany), J. Lyon (ORNL), L.P. Ku (PPPL) and the ARIES Team 16th TOFE Meeting September 14 - 16, 2004 Madison, WI Objectives • Define radial builds for proposed blanket concepts. • Propose innovative shielding approach that minimizes radial standoff. • Assess implications of new approach on: – Radial build – Tritium breeding – Machine size – Complexity – Safety – Economics. 2 Background • Minimum radial standoff controls COE, unique feature for stellarators. • Compact radial build means smaller R and lower Bmax fi smaller machine and lower cost. • All components provide shielding function: – Blanket protects shield Magnet Shield FW / Blanket – Blanket & shield protect VV Vessel Vacuum – Blanket, shield & VV protect magnets Permanent Components • Blanket offers less shielding performance than shield. • Could design tolerate shield-only at Dmin (no blanket)? • What would be the impact on T breeding, overall size, and economics? 3 New Approach for Blanket & Shield Arrangement Magnet Shield/VV Shield/VV Blanket Plasma Blanket Plasma 3 FP Configuration WC-Shield Dmin Magnet Xn through nominal Xn at Dmin blanket & shield (magnet moves closer to plasma) 4 Shield-only Zone Covers ~8% of FW Area 3 FP Configuration Beginning of Field Period f = 0 f = 60 Middle of Field Period 5 Breeding Blanket Concepts Breeder Multiplier Structure FW/Blanket Shield VV Coolant Coolant Coolant ARIES-CS: Internal VV: Flibe Be FS Flibe Flibe H2O LiPb – SiC LiPb LiPb H2O * LiPb – FS He/LiPb He H2O Li4SiO4 Be FS He He H2O External VV: * LiPb – FS He/LiPb He or H2O He Li – FS He/Li He He SPPS: External VV: Li – V Li Li He _________________________ * With or without SiC inserts. -
Medical Isotope Production in Liquid-Fluoride Reactors
Medical Isotope Production in Liquid-Fluoride Reactors Kirk Sorensen Flibe Energy Huntsville, Alabama kirk.sorensen@flibe-energy.com 256 679 9985 Flibe Energy was formed in order to develop liquid-fluoride reactor technology and to supply the world with affordable and sustainable energy, water and fuel. Liquid-Fluoride Reactor Concept Reactor Containment Boundary Turbine Coolant LiF-BeF2-UF4 LiF-BeF2 outlet Primary HX Gas Heater Gas Cooler Generator Reactor core Compressor Coolant inlet Drain Freeze valve Tank Electrical Warm Recompressor Main Compressor Turbine Generator Saturated Air Cool Dry Air Gas Heater High-Temp Low-Temp Gas Cooler Recuperator Recuperator Cooling Water Accum Surge Short-Term Gas Holdup Cryogenic Long-Term Gas Holdup Storage Accum Surge ea Fluor Decay Scrub Decay Tank KOH Bi(Th) Bi(Pa,U) Isotopic Quench metallic Th feed H2 ulFluorinator Fuel 2Reduction H2 HF Electro Bi(Th,FP) Bi(Li) Cell metallic HDLi feed F2 Water Water Coolant Coolant Torus Torus Drain Tank Waste Tank 7 Fuel Salt ( LiF-BeF2-UF4) Fresh Offgas UF6-F2 200-bar CO2 7 Blanket Salt ( LiF-ThF4-BeF2) 1-day Offgas F2 77-bar CO2 7 Coolant salt ( LiF-BeF2) 3-day Offgas HF-H2 Water 7 Decay Salt ( LiF-BeF2-(Th,Pa)F4) 90-day Offgas H2 Waste Salt (LiF-CaF2-(FP)F3) Helium Bismuth The Molten-Salt Reactor Experiment was an experimental reactor system that demonstrated key technologies. Lanthanide Fission Products Alkali- and Alkaline-Earth Fission Product Fluorides c ORNL-TM-3884 THE MIGRATION OF A CLASS OF FISSION PRODUCTS (NOBLE METALS) IN THE MOLTEN-SALT REACTOR EXPERIMENT R. -
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 ................................................................................................................ -
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. -
System Studies of Fission-Fusion Hybrid Molten Salt Reactors
University of Tennessee, Knoxville TRACE: Tennessee Research and Creative Exchange Doctoral Dissertations Graduate School 12-2013 SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS Robert D. Woolley University of Tennessee - Knoxville, [email protected] Follow this and additional works at: https://trace.tennessee.edu/utk_graddiss Part of the Nuclear Engineering Commons Recommended Citation Woolley, Robert D., "SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS. " PhD diss., University of Tennessee, 2013. https://trace.tennessee.edu/utk_graddiss/2628 This Dissertation is brought to you for free and open access by the Graduate School at TRACE: Tennessee Research and Creative Exchange. It has been accepted for inclusion in Doctoral Dissertations by an authorized administrator of TRACE: Tennessee Research and Creative Exchange. For more information, please contact [email protected]. To the Graduate Council: I am submitting herewith a dissertation written by Robert D. Woolley entitled "SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS." I have examined the final electronic copy of this dissertation for form and content and recommend that it be accepted in partial fulfillment of the equirr ements for the degree of Doctor of Philosophy, with a major in Nuclear Engineering. Laurence F. Miller, Major Professor We have read this dissertation and recommend its acceptance: Ronald E. Pevey, Arthur E. Ruggles, Robert M. Counce Accepted for the Council: Carolyn R. Hodges Vice Provost and Dean of the Graduate School (Original signatures are on file with official studentecor r ds.) SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS A Dissertation Presented for the Doctor of Philosophy Degree The University of Tennessee, Knoxville Robert D. -
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. -
Effect of Flibe Thermal Neutron Scattering on Reactivity of Molten Salt Re- Actor
EPJ Web of Conferences 239, 14008 (2020) https://doi.org/10.1051/epjconf/202023914008 ND2019 Effect of FLiBe thermal neutron scattering on reactivity of molten salt re- actor 1 1, 1, 1 1 1,2 1 Yafen Liu , Wenjiang Li ∗, Rui Yan ∗∗, Yang Zou , Shihe Yu , Bo Zhou , and Xiangzhou Cai 1Shanghai Institute of Applied Physics, CAS, Shanghai, 201800, China 2University of Chinese Academy of Sciences, Beijing, 100049, China Abstract. Thermal neutron scattering data has an important influence on the calculation and design of reactor with a thermal spectrum. However, as the only liquid fuel in the Gen-IV reactor candidates, the research on the thermal neutron scattering effect of coolant and somewhat moderator FLiBe has not been carried out sufficiently either experimentally or theoretically. The effect of FLiBe thermal neutron scattering on reactivity of TMSR- LF (thorium molten salt reactor - liquid fuel), TMSR-SF (thorium molten salt reactor - solid fuel) and MSRE (molten salt reactor experiment) were investigated and compared. Results show that the effect of FLiBe thermal neutron scattering on reactivity depends to some extent on the fuel-graphite volume ratio of core. Calculations indicate that FLiBe thermal neutron scattering of MSRE (with the hardest spectrum) has the minimum effect of 41 pcm on reactivity, and FLiBe thermal neutron scattering of TMSR-SF (with the softest spectrum) has the maximum effect of -94 pcm on reactivity, and FLiBe thermal neutron scattering of TMSR-LF has an effect of -61 pcm on reactivity at 900 K. 1 Introduction tionally [5]. When neutrons are moderated to thermal, the binding of the scattering nucleus in a liquid moderator ma- In a thermal neutron reactor, when neutron energy exceeds terial affects the scattering cross section and the energy and 4 eV, the binding energy within the molecule is not impor- angular distribution of secondary neutrons. -
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. -
“Advanced” Isn't Always Better
SERIES TITLE OPTIONAL “Advanced” Isn’t Always Better Assessing the Safety, Security, and Environmental Impacts of Non-Light-Water Nuclear Reactors “Advanced” Isn’t Always Better Assessing the Safety, Security, and Environmental Impacts of Non-Light-Water Nuclear Reactors Edwin Lyman March 2021 © 2021 Union of Concerned Scientists All Rights Reserved Edwin Lyman is the director of nuclear power safety in the UCS Climate and Energy Program. The Union of Concerned Scientists puts rigorous, independent science to work to solve our planet’s most pressing problems. Joining with people across the country, we combine technical analysis and effective advocacy to create innovative, practical solutions for a healthy, safe, and sustainable future. This report is available online (in PDF format) at www.ucsusa.org/resources/ advanced-isnt-always-better and https:// doi.org/10.47923/2021.14000 Designed by: David Gerratt, Acton, MA www.NonprofitDesign.com Cover photo: Argonne National Laboratory/Creative Commons (Flickr) Printed on recycled paper. ii union of concerned scientists [ contents ] vi Figures, Tables, and Boxes vii Acknowledgments executive summary 2 Key Questions for Assessing NLWR Technologies 2 Non-Light Water Reactor Technologies 4 Evaluation Criteria 5 Assessments of NLWR Types 8 Safely Commercializing NLWRs: Timelines and Costs 9 The Future of the LWR 9 Conclusions of the Assessment 11 Recommendations 12 Endnotes chapter 1 13 Nuclear Power: Present and Future 13 Slower Growth, Cost and Safety Concerns 14 Can Non-Light-Water Reactors