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2 cSS s, Ve-* ORNL-4728 1 i MOLTEN SALT REACTOR PROGRAM Semiannual ^Piiogftess ^epoftt ^Pcjkiod finding August 31.1971 ^2V OAK RIDGE NATIONAL .LABORATORY jP^A"' 6V SiOS 'A^'.? ^PQRAr*rs ziy^mM BLANK PAGE Printed in the United States of America. Available from National Technical Information Service US. Department of Commerce 5285 Port Royal Road. Springfield. Virginia 22151 Kxe: Printed Copy $3.00; Microfiche $0.95 This report was piepared as an account of work sponsored by the United States Government. NeroVr the United States nor the United States Atomic Energy Commission, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liaMity or responsibi'ity for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or lemeseius that its use would not infringe privately owned rights. OAK RIDGE NATIONAL LABORATORY OPERATED »Y UNION CAftftlDC CORPORAT10M HUCUAt DiVfSlON POST OFFICE ftOX X OAK RIDGE, TENNESSEE 3700 ORNL-4728 Errata MOLTEN-SALT REACTOR PROGRAM SEMIANNUAL PROGRESS REPORT FOR PERIOD ELDING AUGUST 31, 1973 The TID-4500 category, UC-80 — Reaccor Technology, was Inadvertently ositted from the title page. The following corrections should be aade to your copy (or copies) of the subject report: Page v, PART 3. MATERIALS DEVELOPMENT should start with Chapter 11. Page 1, PART 1. MSBR DESIGN AND DEVELOPMENT should have the names R. B. Briggs and P. N. Haubenreich. Page 41, Fig. 5.3 should be replaced with the corrected version. Page 45, PART 2. CHEMISTRY should have the name W. R. Grimes. The attache I page is printed on gummed stock and the corrections may be cut out and pasted directly on the copy. Pagev PART 3. MATERIALS DEVELOPMENT 11. EXAMINATION OF MSRE COMPONENTS 89 I I.I Examination of Hastelloy N Componenb Exposed to Fuel Salt in the MSRE 89 11.2 Examination of Components from In-Pile Loop 2 94 11.3 Observations of Grain Boundary Cracking in the MSRE 96 1' A Examination of a Graphite Moderator Element 106 11.5 Auger Analysis of the Surface Layer on Graphite Removed from the Core of the MSRE 107 12. GRAPHITE STUDIES Ill 12.1 The Irradiation Behavior of Graphite at 715°C S12 12.2 Procurement of Various Grades of Csrbcn and Graphite 114 123 X-Ray Studies i 16 12.4 Thermit Property Testing 117 12.5 Helium Permeability Measurements on Various Grades of Graphite 118 12.6 Reduction of Graphite Permeability by Pyrorytk Carbon Sealing 119 12.7 The Magnified Topography of Graphite Sealed with Pyroly tic Carbon 120 12.8 Reduction of Permeability by Fhnd Impregnation 121 12.9 Fundamental Studies of Radiation Damage Mechanisms in Graphite 122 Paget Part 1. MSBR Design and Development R.B.Briggj P. N. Haubenreich Page 41 0RNL-0WG 7I-I3OT {»»"*) • 1 (Mt. TM6 POMT E «2 1—•r' I 2 4 0 400 500 600 700 800 900 TEMPERATURE PC) Page 45 Part 2. Chemistry W. R. Grimes ORNL-4728 Contact No. W-740S«ne-26 MOLTEN-SALT REACTOR PROGRAM SEMIANNUAL PROGRESS REPORT For Period Endmg Aopm 31,1971 M. W. Rosenthal, Program Director R. B. Briggs, Associate Director P. N. Haubenreich, Associate Director -NOT7CE- FEBRUARY 1972 OAK RIDGE NATIONAL LABORATORY Oak Ridgt, TMMMMM 37830 operated by UNION CARBIDE CORPORATION for the U «. A fOMIC ENERGY COMMISSION MSTMNITION Of THIS DOCUMENT «l This report is one of a series of periodic feports in which we describe the progress of the program. Other reports issued in this series are fisted below. ORNL 2474 Period Ending January 31,1958 ORNL-2626 Period 1-nd J»g October 31,1958 ORNL-2684 Period Ending January 31,1959 ORNL-2723 Period Ending April 30,19S9 ORNL-2790 FSriod Ending J«!y 31,1959 O*NL-28<>0 Period Ending October 31,1959 ORNL-2973 Periods Lading Jamsry 31 and April 30,1960 ORNL-3014 Prriod Ending July 3i, 1960 ORNL 3122 Period Ending Februuy 28,1961 ORNL-3215 Period Esding August 31,1961 JRNL-3282 Period Ending F^i ^y 28,1962 ORNL-3369 Period Ending Augu* 31.19f 2 ORNL-3419 Period Ending January 31,1963 ORNL-3529 Pfcriod Ending Jury 31,1963 OPNL-3626 Period Ending January 31,1964 OKNL-3708 Period Ending Jury 31,1964 ORNL-3812 Period Ending February 28,1965 ORNL-3872 Period Ending August 31,1965 ORNL-3936 Period Ending February 28,1966 ORNL4037 H~iod Ending August 31,1966 0RNL4119 Period Endinf February 28,1967 ORNL-4191 Period Ending August 31,1967 ORNL4254 Period Ending February 29,1968 ORNL4344 Period fcwHng Augurt 31,1968 ORNL4396 Period Ending February 28,1969 0RNL4449 Period Ending August 31,1969 ORNL4548 Period Ending February 28,1970 ORNL4622 Period Ending August 31,1970 ORNL4676 Period Fading February 28,1971 Contents PART 1. MSBR DESIGN AND DEVELOPMENT 1. DESIGN 2 1.1 Motten-Sah Pewamiatiot Reactor Desiga Sttfaly 2 l.l.i General 2 1.1.2 ReactorCore 3 1.13 MSDROrT-Gas and Salt Pump-Back System 4 !.!.4 FJram Tar* CooSsg System 7 1.1.5 DrainV?JveCell 9 1.1.6 Dean C* Catch Pan 9 1.2 Some Consequssca of Tubing Failure in tir. If S8R Heat Exchanger 9 1J Side-Stream Processing cf the MSBR Primary Flow for Xenon and/or Iodine Removal 10 1.4 MSBE Design 11 1.4.1 Reactor Core Power and Neutron Pux Distribution 11 1.4.2 Reactor Core Heat Removal , 11 1.43 Maintenance Studies 14 1.5 MSBR Industrial Design Study 15 2. REACTOR PHYSICS 16 2.1 MSR EXPERIMENTAL PHYSICS 16 2.J.1 HTLTR Lattice Experiments 16 2.2 PHYSICS ANALYSIS OF MSBR 18 2.2.1 Matron Irradiation Effects Outside the MSBR Core 18 2.2.2 MSBE Nuclear Characteristics 20 2.23 Fixed-Moderator Molten-Salt Reactor 21 3. SYSTEMS AND COMPONENTS DEVELOPMENT 26 3.1 Gaseous Fission Product Removal 26 3.1.1 Gas Separator and Bubble Generator 26 3.1.2 Bubble Formation and Coalescence Test 27 3.2 Gas System Te*t Facility 27 3.3 Off-Gas Systems 28 33.1 Off-Gas System for Molten-Salt Reactors 28 3.3.2 Computer Design of Charcoal Beds for MSR Off-Gas Systems 29 3.4 MoUcn-Salt Steam Generator 29 3.4-1 Steam Generator Industrial Program 29 3.4.2 Molten-Salt Steam Generator Technology Facility 29 3.4.3 Molten-Salt Steam Generator Test Proposal*. 30 11*. IV 3-5 Sodium Ftuoroborate Test Loop 31 3.5.1 Inspection of FKP Pump Rots-y Element 31 3 6 Coobnt-Salt Technology Facility 34 3.7 MSBR Pumps 35 3.7.1 Sah Pumps for IISRP Technology Facffities 35 3.7.2 ALPHA Pump 35 3.73 Drain Tank Jet Pump System for MSDR 36 4. niSTRUMENTATION AND CONTROLS 38 4.1 Transient aid Control Studte of the MSBR Sysm 38 4.2 MSRE Design and Operations Report, Part !!B, Nucicar as<J Pioce» Imtnuneniaoon 38 43 Further Discussion of Intlrumentation and Controls Development Needed for the Mokce£alt Breeder Re*, tor 38 5. HEAT AND MASS TRANSFER AND THERMOPHYS1CAL PROPERTIES 39 5.1 Heat Transfer 39 5.2 Thermophysical Properties 41 53 Mass Transfer to Circulating Bubbles 41 PART 2. CHEMISTRY 6. POSTOPERAHONAL EXAMINATION OF MSRE 46 6.1 Examination of Moderator Graphite from MSRE 46 6.1.1 ResuteofVralFjtanmiation 46 6.1.2 Segmenting of Graphite Stringer 46 6.13 Examination of Surface Samples by X-ray Diffraction 47 6.1.4 Examination with a Gamma Spectrometer 47 6.1.5 Miffing of Surface Graphite Samples 49 6.1.6 Radiochemical and Chemical Analyses of MSRE Graphite 49 6.2 Cesium Isotope Migration in MSRE Graphite 51 63 Fission Product Concentrations on MSRE Surfaces 54 6.4 Metal Transfer in MSRE Salt Circuits 55 7. HYDROGEN AND TRITIUM 3EHAVIOR IN MOLTEN SALT 56 2 7.1 The Solubility of Hydrogen in Molten «JF-BeF2 56 7.2 Permeation of Metals by Hydrogen 57 7.3 Tritium Control in an MSBR 59 73.1 MassSpectrometric Examination of Stability of NaBF3OH 59 7.3.2 The Thermal Stability of Nitrate-Nitrite Mixtures 59 7.4 Dissociating-Gas Heat Transfer Scheme and Tritium Control in Molten-Salt Power Systems 60 8. PROCESSING CHEMISTRY 62 8.1 The Oxide Chemistry of Pa** in Molten LiF-BeF2-ThF4 62 5 8.2 The Oxide Chemistry of Pa * in Uranium-Containing Molten LiF-BeF2 -ThF4 64 8.3 Reductive Extraction Distributions of Barium and Thorium Between Bismuth-Lead Eutectic and Breeder Fuel Solvent 67 8.4 Removal of Cerium from Lithium Chloride by Zeolite 67 V 9. DEVELOPMENT AND EVALUATION OF ANALYTICAL METHODS FOX MOLTEN SALT REACTORS 69 9.1 In-line Chemical Analysis of Molten Fluoride Salt Streams 69 9.2 Slotted Fr Jbe lor In-Une Spectral Measurements 70 9.3 In-line Determination of Hydrolysis Products in NaBF« Cover Gas 71 9.4 Determination of Hydrogen in Fluoroborate Salts 73 9.5 Voltammetric and Electrolysis Studies of Hydroxide Ion in Molten NaBF4 74 9.6 Ekctroanarytkal Studies in the NaBF4 Coolant Salt 75 9.7 FJectroandytical Studies of NKJI)in Molten Fluoride Fuel Solvent 75 10. OTHER FLUORIDE RESEARCHES 77 10.1 Absorption Spectroscopy of Molten Fluorides 77 10.1.1 The Disproportionarjon Equilibrium of UF3 Solutions 77 10.1.2 Estimation of the Available F ~ Concentrations in Melts by Measurement of UF4 Coordination Equilibria 77 10.2 Solubility of BF3 in Fluoride Melts 78 10.3 Fluorides and Oxyfluorides of Molybdenum and Niobium 80 10.3.1 Mass Spectroscopy of Molybdenum Fluorides 80 10.3.2 Mass Spectroscopy of Niobium Fluorides and Oxyfluorides 80 10.4 Electrical Conductivity of Molten and Supercooled Mixtures of NaF-BeF2 .
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  • System Studies of Fission-Fusion Hybrid Molten Salt Reactors

    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.