CENTRAL RESEARCH LIBRARY DOCUMENT COLLECTION LIBRARY LOAN COPY PO NOT TRANSFER TO ANOTHER PERSON se¢ {o see this document, send in neme wilth docupiont and the {ihrary will arrangs a foan, i g o :o <4 ) ¥} " <) ORNL- This document consists of 214 pages. copy {pp of 222. , ‘ Contract No. W-7405-eng-26 AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 10, 1952 R. C. Briant, Director A. J. Miller, Assistant Director Edited by: W. B. Cottrell DATE ISSUED CMAY 7 {959 DAK RIDGE NATIONAL LABORATORY . operated by CARBIDE AND CARBON CHEMICALS COMPANY A Division of Union Carbide and Carbon Corporation Post Office Box P 0ak Ridge, Tennessee 1227 Series A. 1 | T g o p— = L cciiant oo T T T T TR T P ik e e i1 INTERNAL DISTRIBUTION G. M. Adamson . J. Barton D. S. Billington N P. Blizard Brasunas CiBriant yuce Bug Calljhan Cardwell Center j Cisar N Clewett % Clifford W\ Cottrell N Cowen ® Eister Ergen Fraas Gall A Gr aham & Grimes y ollaender 4 Householdgr Humes (Ku25) Jones fif Kellhqftz QLM NEOOPErrE0EFO=E=UI00SOTFETE R M ERHY RN I I TN NP ERE =0 E D Its trans Felbeck (C&CCC) X defxned 1nt%h¢'F““ 82-91, < Enérgy Act-. o isclosure of its i afi?“manne to“?fi?fifine_- person is proh:b:ted ‘ ORNL - 1227 ~f3gress . Lavers_;fill2) 35. W, 36. R. S. L1v1n;ston 37. R. (" 38. W. D. 39. W. B/ 40. J. L. 41. ;?jj. Miller 42. SK. Z. Morgan 43¢ E. J. Murphy fifi%l H. F. Poppendiek 745, P. M. Reyling 4 . H. W. Savage 47. R. W. Schroeder 48. E. D. Shipley 49. O. Sisman 50. A. H. Snell 51. F. L. Steahly 52. R. W. Stoughton 53. C. D. Susano 54, J. A. Swartout 55. E. H. Taylor 56. F. C. Uffelman 57. F., C. VonderLage 58. A. M. Weinberg 59. C. E. Winters 3 60. Biology Library ¥1-62. Chemistry Library & 63. Health Physics Library m"64. Metallurgy Library $5. Physics Library 66-6%. Training School Library 68-77% ANP Library 78-81.% Central Files '%Eentral Files (0.P.) orized 7 e T TR T PR Pt 4 g x T e X RS I R TS e e s e e —— ! : _ s e .. = RN PN ol T dFECYRE LRI H SN ; EXTERNAL DISTRIBUTION f <4 G i R ; 92-101% ~Argonne National Laboratory £ IOSQKHArmed Forces Special Weapons Panect (Sandia) 104-111. "Agomic Energy Commission, Wa 1ngton 112. Battelle Memorial Inst1tute 113-117. Bro;%haven National Labor ry 118, Burea§§nf Aeronautics ‘ 119. Bureau of\Ships / 120-125. Carbide andsCarbon CM€micals Company (Y-12) 126, Chicago Patent Grop 127. Chief of Naval:@- Ffearch 128-132. duPont Company ¢ %\ 133-157. General Electg 5' any, Oak Ridge 158-161. General Ele f;r1c Comg y, Richland 162. H. K. Ferd uson Company % 163. Hanford f-eratlons Offlc 164-167. Idaho @ferations Office 168. Toway/ gfflte College - 169-170. Lo_f fand Area Office, AEC % 171-174. Krb?“s Atomic Power Laboratory\ - 175-177. ;bm'Alamos fii ¥ 178. /Mfssachusetts Institute of Technology (Kaufmann) 179~180_'v,ound Laboratory % 181-18é&_* ational Advisory Committee for Aerfihautlcs, Cleveland 65." National Advisory Committee for Aero%@ptlcs, Washington 186£4187. New York Operations Office . 18§ 189. 'North American Aviation, Inc. /" 190. Patent Branch, Washington £ 191. Savannah River Operations Office s £192-193. University of California Radiation Laboratobhy 4 194-197. Westinghouse Electric Corporation ' 3 198-207. Wright Air Development Center 208-222. Technical Information Service, Oak Ridge < o o a}#ifT;éu;£ €%chtg%i§§E§estr1cted Data4ga {3 111 TR T T iv Reports previously issued ORNL-528 ORNL-629 ORNL-768 ORNL- 858 ORNL-919 ANP-60 ANP-65 ORNL- 1154 ORNL-1170 Period Period Period Period Period Period Period Period Period in this series are as follows: Ending November 30, 1949 Ending February 28, 1950 Ending May 31, 1950 Ending August 31, 1950 Ending December 10, 1950 Ending March 10, 1951 Ending June 10, 1951 Ending September 10, 1951 Ending December 10, 1951 Aetencenane kb b e o i x) * TABLE OF CONTENTS FOREWORD 1. PART I. REACTOR THEORY AND DESIGN 'SUMMARY AND INTRODUCTION CIRCULATING-FUEL AIRCRAFT REACTOR Reactor with Tandem Heat Exchangers Reactor with Annular Heat Exchangers Reactor Shield Designs Design procedure Activation of the secondary coolant Shield weights and specifications CIRCULATING-FUEL ATRCRAFT REACTOR EXPERIMENT Core Design Primary coolant circuit Secondary coolant circuit Core temperature distribution External Fluid Circuit Pumps Heat exchangérs Primary coolant syétem Secondary coolant system Monitoring circuit Preheating System Reactor assembly Piping ', co ;“ K ,. -def1ned in SHE A ”;“;}magjrgy Act of 1946. ,7,-~V:s“f:¢:”'*'? sclosure of its ”“L o any unauthorized =D WO D ] N1 U W e Pt et — 13 13 15 15 16 16 16 18 18 18 18 18 19 19 o e B L e vi Heat exchangers Electrical Power Circuits Control System Shim control Regulating and safety rods High-temperature fission chamber Control console and panel Reactor dynamic computer Instrumentation Building EXPERIMENTAL REACTOR ENGINEERING - Seals and Closures Frozen sodium seal Frozen fluoride seal o , Bellows type of face seal for ARE pump Stuffing-box seals for molten fluorides Lubrication of seals and shafts Pumps ARE centrifugal pump Laboratory frozen-fluoride-sealed centrifugal pump Worthite frozen-sodium-sealed pump Modified Durco centrifugal pumps Forced-convection-qooled sodium-sealed pumps Canned-rotor pumps Valves Packing-glandiseal test eguipment This document- corfalis K defined in.-the At Its transmittal”or’#led™ contents in g fhas person is gf o1l 19 19 20 20 20 21 21 21 21 22 23 23 24 24 24 25 26 26 26 27 27 27 29 29 29 29 w2 4.“ " Frozen fluorides valve Ball check valves Valve seat test Heat Exchangers Aircraft type of radiator Sodium-to-air radiator 'Fluoride-to-fluoride heat exchanger NaK-to-NaK heat exchanger ‘Heat transfer in circulating fluoride loops Instrumentation Flow measurement Pressure measurement Temperature meaSfirement 7 | Level controls and level indicators B Heatlng and Cooling of High-Temperature Systems External heatlng systems Induction heating Resistance heating Insulation testing ARE core preheating Technology of High-Temperature Liquids Fluoride preparation and handling Sampllng and analyz1ng techniques D1ffu51V1ty of he11um through stainless steel Cleanlng and 1nspect10n technlques REACTOR PHYSICS Clrculatlng-Fuel Alrcraft Beactor | | 30 30 30 30 30 31 31 32 32 33 33 34 34 34 35 35 36 36 36 37 37 37 39 39 39 41 41 vii o B, Rl B g . g e by heb, o, o, o, e R i d Eale e ¢ o o i Gk o e el e B, i . bl e, o S B pan B i e e d i i MW“aMmMmm&mmwmmMmm.mMMkm s Bt e ak Oscillations » Slow kinetic effects Critical mass Neutron leakage spectra Alkali Hydroxide Moderated Aircraft.Reactor Survey Calculations of the Circulating-Fuel ARE ARE core design Critical mass and total uranium investment Reactivity coefficients | Power distribution Neutron flux and leakage spectra Statics of ARE Controls Shim control requirements Regulator rod Safety rods Specific Design Problems of the Cifculating—Fuél ARE 5. CRITICAL EXPERIMENTS - Direct-Cycle Reactor j - , - Control rod calibration ’ - / Temperature effects 1 Reflector studies | »Graphite Reactor 3,»-:7‘&%’%"‘“"“*’ Circulating-Fuel Reactor Correlation of Theory and Critical Experiments Criticality with nonhydrogenous moderators Criticality with hydrogenous moderatops By CT) DOATA 5 . 4 estriéte&fbaifl;aifiww;r brgy Act of 1946, egsure of its viii 42 44 44 46 48 48 48 50 50 51 52 54 54 54 55 57 59 59 59 61 61 62 62 63 63 65 . ¥ A { STAWARY AND INTRODUCTION. 6 Foil exposures Danger coefficients Rod sensitivity - Gap experiment " PART IT. 'SHIELDING RESEARCH 'BULK SHIELDING REACTOR "Mockup of the Divided Shield "Reactor Calibration DUCT TESTS Theoretical Treatment of Duct Transmission Measurement of Air-Filled Ducfis in Watér Straight ducts Ducts with bends Comparison with theory TOWER SHIELDING FACILITY Tower Facility Design Experimental Program NUCLEAR MEASUREMENTS .Measurements with theIS-Mev Van de Graaff Accelerator Total cross section of iron Inelastic scattering levels in iron Time-of-Flight Neutron Spectrometer Background measurements Indium resonances OPRexgy Act of 1946, losure of its ‘ unauthorized defined in Lhe” AtQuiN - Its transsise€al”. " contents; WM y~ha " person’ isTprG R 65 66 66 68 69 71 73 73 76 79 79 80 80 80 80 89 89 89 93 93 93 93 03 93 94 1X e e Sk s e kg kb SECUR PART I1I. MATERIALS RESEARCH SUMMARY AND INTRODUCTION - 10. 1. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS | Low-Melting-Fluoride Fuel Systems NaF-Ber-UF4 KF-Ber-UF4 BbF-Ber—UF4 LiF-NaF-Ber-UF4 Fuel containing zirconium fluoride Simulated Fuel Mixture for Cold Critical Experiment ~Ionic Species in Fused Fluorides Preparation of Standard Fuel Sahples Preparation of Pure Hydroxides Coolant Development LiF-ZrF4 | NaF-ZrF, KF-—_ZI'F4 RbF-ZrF, NaF-KF-ZrF4 NaF-RbF—ZrF4 NaF-_KF-LiF-ZrF4 Preparation of Pure Fluorides Fuel preparation equipment Fuel handling equipment CORROSION RESEARCH Static Corrosion by Fluorides . - o RRBE = trlcted Date as kgy Act of 1946. gt losure of its ally unauthorized This document Q) defined in the A% Its transmitgel-g 95 97 99 99 100 101 101 101 101 102 102 103 104 104 105 105 105 105 105 105 106 106 106 107 1109 110 SECURILY_INTORMATION Effect of pretreatment of fuel Corrosion of structural metals Corrosion by fluorides with various additives Melting point of fluorides after corrosion tests Static Corrosion by Sodium Hydroxide Corrosion of special alloys Corrosion by sodium hydroxide with various additives Corrosion of refractory materials Static Corrosion by L1qu1d Metals Corrosion by low meltlng point alloys Corrosion by sodium-lead alloy - Facilities for Dynamic Corrosion Testing Thermal convection loops Seesaw corrosion tests Differential temperature tests Rotating dynamic corrosion tests Forced convection loops Dynamic Corrosion by Fluorides Corrosion by fluorides in thermal convection loops wh Corrosion by fluorides in a seesaw furnace Standpipe tests of fluoride corrosion Dynamic Corrosion by Hydroxides Corrosion by hydroxides in thermal convection loops Corrosion by sodium hydroxide in seesaw tests Standplpe tests of sod1um hydrox1de corrosion Fundamental Corros1on Besearch | Poss1b1e equ111br1a among_fluorldes and metals ted Data as | Thls docume.n " "Rt ' 4b~ 3 *‘rgy Act of 1946. - defined in Je- . “-Its trans} isclosure of its unauthorized 110 110 112 113 114 114 115 116 116 116 118 119 119 120 120 120 122 122 122 124 124 127 127 129 129 129 132 X1 o ] 1 ,’ | é i FMF measurements in fused fluorides . , _ 135 ; | Electrode potentials in fused sodium hydroxide | 135 Polarography of sodium hydrox1de in silver and plat1num ' 136 , Magnetic susceptibility of stainless steel exposed to fluorides 136 12. METALLURGY AND CERAMICS ’ - - 139 § Fabrica;ion of Reactor Elements - . - -”f:'139 § ~ YCold drawing of tubular solid fuel elements - =7l 7140 : ARE control rod | 140 Cone-Arc Welding 142 | Equipment} | % 142 _ Principle of operation ' | 142 i ,'t - Experimental procedure | _ 142 | Brazing 143 : Flow tests . | 144 z Corrosion of brazing alloys 144 ; Mechanical Testing of Materials | 147 Inconel creep and stress data | 147 ; Tube-burst tests | ' 148 § Operation of creep and stress-rupture equipment 148 | Ceramics Laboratory ' ‘ ' ' 151 ] Ceramic appliéations to reactors | 151 Coatings for the radiator - 151 é Ceramic laboratory equipment 152 % Microscopic examination of fluorides ‘ . _ 152 13. HEAT TRANSFER AND PHYSICAI PROPERTIES RESEARCH - ._ - 153 - - Viscosity of Fluoride Mix3 153 ~ ._?ustrlcted Data as 0 kpergy Act of 1946. shendesclosure of its Y any unauthorlzed '"“Thls document _,. 'w;deflned in th*“ A : _r.: contentsf7;;jnV]:¥ person - 1s*propt T Viscosity of NaF-KF-LiF Viscosity of NaF-KF-LiF-UF, Modifications of viscosity apparatus Thermal Conductivity of Liquids and Solids L3 Heat Capacities Vapor Pressure of Liquid Fuels Physical Property Data Natural Con?ection in Confined Spaces with Internal Heat Generation Analysis_of Heat Transfer in a Circulating-Fuel System Heat Transfer Coefficients Heat transfer in molten lithium, Heat transfer to fused salts and hydroxides Entrance region heat transfer in a sodium system Heat Transfer of Boiling Liquid Metals 14. RADIATION DAMAGE Irradiation of Fused Materials Pile irradiation of fuel Cyclotron irradiation of fuel i " Inpile Circulating Loops Creep Under Irradiation Radiation Effects on Thermal Conductivity PART 1V. APPENDIXES | .SUMMARY AND INTRODUCTION %/15. THE' SUPERCRITICAL-WATER REAC st ':Thxs.document , "f ; '~f”£icted Data as - | defined in E AT MR Act of 1946. : ' Its™ trans”;_“y'wg@a; - losure of its conternts- A ;ie “man -; unauthorized ) 153 154 154 154 155 155 156 156 159 161 161 161 161 162 163 163 164 165 165 165 166 167 169 171 X111 i b i i, Al dinatia i i Description of Reactor Conclusion of the NDA Study Recommendations of the NDA Study 16. ANALYTICAL CHEMISTRY .Studies‘of Diatomaceous Earth Analytical Studies of Fluoride Eutectics Uranium Beryllium Total alkali metals Total fluoride Nickel Chromium Manganese Slllcon Solublllty of Borlc ‘Acid in Water | Clarlty of Borated Water. Studles of Alkali and Alkaline Earth Hydroxide Coolants Analytical Services 17. LIST OF REPORTS ISSUED 18. DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS AT ORNL Reactor and Component Design Shleldlng Research Materlals Research Technical Administration . -~ . . . 2, This document zk.:=;-- 2 “iété&5bfiiafaq_ "+ . defined 1n_tr- X1V 171 171 172 175 175 175 176 176 176 176 177 177 177 177 1T 178 178 179 183 183 185 186 191 © oy FIGURE O O A N ot 8 W BN st o 11 12 13 14 15 - 16 LIST OF FIGURES TITLE Tandem Heat Exchanger Arrangement for Circulating-Fuel Reactor Annular Heat Exchanger Arrangement for Circulating-Fuel Reactor Circulating-Fuel ARE Core Design Arrangement of External Fluid Circuit Equipfient Frozen Flfioride'Seal Tester “ ARE Centrifugal Pump Ball Check Valve Schematic Diagram of Circulating-?uel‘Aircraft Reactor Leakage Spectrum Through the Reflector of the Circulating- Fuel .Aircraft Reactor Leakage Spectrum Around the Fuel Pipes of the Circulating-Fuel Aircraft Reactor : : Critical Mass vs. Core Diameter for Hydroxide Reactors with Thick Reflectors of the Same Comp051t10n Power Distribution in the Core of the Circulating-Fuel ARE Leakage Spectrum from the Reflector of the Circulating-Fuel ARE Leakage Spectrum from the Open Ends of the Circulating-Fuel ARE Transverse Neutron Flux Spectrum for Three Sections Through the Core of the Circulating-Fuel ARE ‘ Radial Neutron Flux’ Dlstrlbutlon in the Core of the Clrculatlng- Fuel ABE ' Loadlng Chart of Crltlcal Assembly of Direct-Cycle Reactor :Beact1v1ty as a Funct1on of Control Rod Position Reactivity vs. Temperetfiref'“""“, . This document co,;& >3 f;;fi“1cted Data as : def1ned in "the AR Sger gy - At of 1946. e e Le'fiJ“ B{sclosure of its - contents 13 Q)Y unauthor1zed o person is T prohthi PAGE 10 14 17 25 28 30 45 46 47 49 52 53 53 54 55 60 61 61 XV B 3 4 A o e i B i i . R o e i . b, s s FIGURE 20 91 22 23 24 25 26 “a1 .28 29 30 31 39 33 34 35 36 XVi TITLE Relétive Position of Reactof, Divided Shield, and Gamma-Ray Spectrometer in Bulk Shielding Facility | Gamma-BRay Spectra at 130Acm from the Water-Reflected Reactor Fuel Assembly Arrangement of Bulk Shielding Facility Reactor Center Line Measurements of Neutron Transmission Through Cylindrical Ducts in Water Traverse Measurements of Neutrons in Water Beyond Cyllndrlcal Ducts Center Line Measurements of Neutron Transmission in Water Through Cylindrical Ducts with Variable Bends Traverse Measurements of Neutrons in Water Beyond Cylindrical Ducts with Variable Bends Comparlson of Relatlve Source Strengths (n ) from Varlous Ducts Comparlson of Relative Source Strengths (n ) from Various Geometries of the 3- 1n.Duct 'Proposed 300-ft Tower Shleldlng Facility X- 6 Aircraft Shield Mockup on Tower Shielding Fac1l1ty Total Cross Section of Iron Corrosion of Inconel in a Fluorlde Fuel [(NaF-KF- UF ) + 2% Zrl for 100 hr at 816°C 'Intergranular Attack of Type-310 Stainless Steel Tested in 44% Lead-56% Bismuth Alloy for 100 hr at 1500°F Alloying Attack of Inconel Tested in 43% Tin-57% Bismuth Alloy for 100 hr at 1500°F Seesaw Apparatus for Dynamic Corrosion Tests Sections of Hot and Cold Legs of an Inconel Convection Loop After Circulating the Fluoride Fuel (NaF-KF-LiF-UF,) for 500 Hours Thiezdoeomen\,_.--f ) defined -in thea:‘;‘~ Act of 1946, gsure of its PAGE 74 75 77 81 82 83 84 87 88 90 91 94 113 117 117 121 125 i T FIGURE 37 38 39 - 45 46 47 48 49 50 TITLE Sections of Hot and Cold Legs of a Type-316 Stainless ‘Steel Convection Loop After: Circulating the Fluoride Fuel (NaF-KF-LiF- UF ) for 123 Hours Nickel Thermal Convection Loop Operated for 117 hr with Sodium ‘Hydroxide Under an Air Atmosphere Nickel Thermal Convection Loop Operated.fdr 296 hr with Sodium Hydroxide Under a Hydrogen Atmosphere Inconel Thermal Convection Loop Operated for 135 hr with Potassium Hydroxide Under a Hydrogen Atmosphere Sectional View of L-Nickel Specimen after 117 hr (14,000 cycles) - with Molten Sodium Hydroxide in Seesaw Apparatus Transverse SectionsAThrough %-in. OD, Cold-Drawn, Tubular, Solid Fuel Elements_ o Corrosion Test of Nlcrobrazed Inconel Tube-to-Header Specimens Exposed to NaOH for 100 hr at 1500°F Corrosion Test of Nicrobrazed Inconel Tube-to-Header Joint Exposed to NaF-KF-LiF-UF, for 100 hr at 1500°F Corrosion of Inconel Tube-to-Beader Joint Brazed with a 60% Manganese-40% Nickel Alloy after 100 hr at 1500°F in NaOH Creep and Stress-Rupture Data for Fine-Grained Inconel Sheet Creep and Stress-Rupture Data for Coarse-Grained Inconel Sheet Dimensionless, Wall-Mixed Mean Temperature Difference as a Function of Reynold’s and Prandtl’s Moduli for a Heat Transfer System with Insulated Pipe Walls Heat Transfer Coefficients for Boiling Mercury Comparison of Bench and Inpile Creep Rates of Nickel . \ ':elbi ) Thi§~dobumg,v)wfifigfi-fliflRéstrlcted Data as defined ingth '“'?L RN gy Act of 1946. Its trap#h Y aclosure of its conten Q™ unauthorized person 1 PAGE 126 128 130 131 132 141 145 146 147 149 150 160 162 166 XVvii Y Y N T e T g T T e e ety s i A - 'y TABLE - b D ke O 13 14 15 16 17 18 O © N N B W B LIST OF TABLES TITLE Shields for Circulating-Fluoride-Fuel Reactors Fuel Temperature After Each Pass Through Core Performance of Fluoride-to-Fluoride Heat Exchanger Analysis of Heater Section (Model A-1) Cofiposition of Vérious Fluoridé Fuels and Coolants‘ Volume Fractions of the Circulating-Fuel ARE Core and Reflector Uranium Requirements of the ARE Reactivity Coefficients Shim Control Requirements CompariSon.of Control fioé Calibfafiions Beacti#ity'ChangelIntroduced by Substituting Beryllium for Air Reactivity Change Effected by Substituting Plastic and StalnleSa Steel for Beryllium in the Reflector Experimental Data on Critical Assemblies 1 and 4 Calculated Results for Critical Assemblies 1 and 4 Compafison of Experimental and Calculated Values of the Ages of Thermal Neutrons in Reryllium and Graphite Expéfiméntal and Calculated Values for the Cadmium and Cadmium- - Indium Ratios ' Expérimental and Calculated Values for the Totai Loss in keff upon Introduction of Various Materials into Assembly 4 Comparison of Calculated Effectlve Source Strength for Stralght Cylindrical Ducts : ’/(_ This document c;rh defined in thf-; ol PAGE 12 16 32 33 38 50 50 51 56 61 62 62 64 64 64 66 67 85 X1X # % i g 4 o 3 ORMATION TABLE TITLE PAGE 19 Comparison of Calculated Effective Source Strength for Bent, o Cylindrical Ducts 86 20 Improvement of Count-to-Background Ratio w1th Aluminum and ' Berylllum Filters L 94 21 Comparison of Break Temperatures from Heat1ng and Coollng Curves of the System NaF-BeF, UF , 100 22 Break Temperatures from Coollng Cur§es of the System KF;BeF -UF, 101 23 . Break Temperatures from Coollng Curves of the System BbF BeF -UF,101 24 Break Temperatures from Heating and Cooling Curves of Zir- : conium-Bearing Fuel Mixtures 102 25 Disposition of Standard Fuel Samples | 103 26 Static Cofros1on of Various Meterlals in the Untreated Fluoride Fuel (NaF-KF-LiF- UF ) in 100 hr at 1500°F 111 27 - Static Corrosion of Inconel and Type- 309 Stalnless Steel by - the Fluoride Fuel (NaF-KF-LiF-UF,) with Magnesium and Zir- ~conium Additives in 100 hr at 1500 F _ 7 112 28 Melting Points of Samples of Fluoride from Corrosion Tests 114 29 " Static Corrosion of Structural Metals by Sodium Hydroxide ' with Various Additives in 100 hr at 1500°F , 115 30 Melting Points of Various Alloys ' 116 | 31 Corrosion of Types-310 and -317 Stainless Steel and Inconel by Various Low Melting Point Alloys in 100 hr at 1500°F 118 32 Corrosion Data from Inconel and Stainless Steel Thermal Con- ' vection Loops Operated with Various Fluorides 123 33 Free Energies and Equilibrifim Constants for Reactions of Metals with Alkali Fluorides 133 34 Free Energies and Equilibrium Constants for Reduction of | Uranlum Tetrafluoride with Metals 134 DATA e I D ;?fstrlcted Data as deflned in the’il-g;‘w gy Act of 1946. a1 6T sNosure of its contents runauthorized person j& :-'$ - e 1 (\i \MU (_} e A , o SECURIY TABLE 35 36 37 38 39 TITLE Viscosity of NaF-KF-LiF-UF, Mixtures as a Function of Uranium Tetrafluoride Concentration Physical Properties of Fluoride Salts Physical Properties of Miscellaneous Materials LITR Tests on Fused Fluoride Fuels in Inconel at 1500°F Summary of Service Analyses ‘qfigtr1cted Data as N Act of 1946, T o §o T ey, PAGE 154 157 158 164 178 XX1 L 3 * . . * " ¥ 3 ! . i A a v ' ‘ . e . a v * & % v g - . " e . . < = « .o . ' - . . - . < . - . ' -~ . » A W TR T I ” ey e Y ~ e A ey v ANP PROJECT QUARTERLY PROGRESS REPORT FOREWORD This is the quarterly progress report of the Aircraft Nuclear Propulsion Project at the Oak Ridge National Laboratory and summarizes the technical _ progress on the project during the period covered. It includes not only the work of the Laboratory under its own contract, W-7405-eng-26, but also the research for the national ANP program performed by Laboratory personnel. The reportis divided into four parts: I. Reactor Theory and Design; II. Shielding Research; III. Materials Research; and IV. Appendixes. Each part may be regarded as a separate entity and has a separate "Summary and Introduction.™ .~ This doéument',fif;?”"-“éstr1cted Data as " defined in the XIS Edergy Act of 1946. Its trang ihsclosure of its conten i unauthorized persoil 4 oo e i pdcadide . S o o e s i b e DA, vzt kb s | ! \ e gt N ¥ ) 1 -y SUMMARY AND Analysis of the circulating-fuel aircraft reactor has been extended to systems incorporating intermediate heat exchangers, various secondary coolants, liquid moderators, and the use of heavier reactor shielding (sec. 1). All these systems utilize the fundamental advantage of the bi- functional fuel-coolant, and appear to be capable of supersonic nuclear propulsion. The location of the heat exchangers around the reactor results in lower shield weight, even with a larger shielded-volume diameter, than a tandem reactor and heat exchanger arrangement. In order to perform a limited amount of aircraft maintenance without special shielding, various modifications of the minimum divided shield specifications have been in- vestigated. Studies of the performance and design of the circulating-fuel air- craft reactor are sufficiently en- couraging that the first Aircraft Reactor Experiment (ARE) to be con- structed by the Oak Ridge National Laboratory will be of this type (sec. 2). The reactor core, as designed for the ARE, consists of a beryllium oxide moderator with a multipass fuel-coolant system. The core and a surrounding beryllium oxide reflector are contained in an Inconel pressure shell. Design of the reactor, fluid circuits, build- ing, and assoclated equipment are essentially complete. The reactor is expected to be in operation early in 1953, ‘ The developmental work in reactor plumbing and associated hardware has been primarily concerned with the technology of high-temperature fluoride mixtures, and a secondary effort has been the study of liquid metals (sec. INTRODUCTION 3). The techniques of the prepara- tion, purification, and handling of the fluoride mixtures have been developed so that 100-1b batches of the treated fluoride may be prepared and loaded 1in adequately cleaned test equipment. Techniques of pumping, sealing, and controlling the fluoride coolants and lubricating moving parts of the systems have been demonstrated at temperatures above 1300°F, and it is considered that these techniques are adequate for ARE application. A centrifugal-flow fluoride pump has operated for weeks with neither me- chanical failure nor leakage. Liquid sodium technology appears to be well in hand, since continued success has been experienced in the operation of sodium (or NaK) pumps, seals, and heat exchangers. The NaK-to-NaK heat ex- changer loop has now operated for 2300 hr with a maximum temperature of 1500°F. Gross heat transfer studies indicate that space-economical systems and components can be built to handle copious quantities of heat, as required by fluoride systems, at temperatures between 1200 and 1800°F. The reactor physics calculations, which have further defined the statics of the circulating-fuel ARE, have led to some general observations regarding the kinetics of both the circulating- fuel ARE and ANP reactor (sec. 4). Although the thrombosis effect is an important concern in the control of these reactors, the loss of the delayed neutrons may not be i1f the circulation of the fuel itself 1s as good a damping mechanism as now indicated. These kinetic difficulties are of less con- " cern to the ARE than to the ANP, since the circulation rate in the ARE is so slow that the control rods can cope with the thrombosis effect and a large fraction of the delayed neutrons are T g s e e b et ANP PROJECT QUARTERLY PROGRESS REPORT emitted into the active volume. The current ARE design has a critical mass of 22.3 1b, a total uranium investment of 74 1lb, 71% thermal fissions, and a leakage-to-absorption ratio of about 1 to 3. Brief studies of hydroxide moderated reactors (including KOH, LLiOH, NaOH, RbOH, and SrOH) show that, except for KOH, the hydroxide moderated reactors require low critical masses and small core volumes for minimum critical mass. Measurements on the critical ex- periment of the simulated General Electric direct-cycle reactor have been completed and the simulated cir- culating-fuel reactor is now being assembled (sec. 5). Evaluations, in terms of contributions to reactivity, have been made of several reflector modifications of the direct-cycle assembly. In addition, the data from the earlier graphite reactor assembly have been correlated with the data from theoretical calculations of the assembly. The correlation lacks precision but gives results that are at least consistent with the experi- mental facts. L] n % -y FOR PERIOD ENDING MARCH 10, 1952 1. CIRCULATING-FUEL AIRCRAFT REACTOR A. P. Fraas, ANP Division A circulating-fuel aircraft reactor system in which the fluid fuel circu- lates directly through the turbojet radiators was described in the last report.{!) Other circulating-fuel aircraft reactor systems, 1ncorporating such features as 1intermediate heat exchangers, various secondary coolants, the use of liquid moderators, and heavier reactor shielding, have been considered in an attempt to determine the most practical system for a func- tional supersonic aircraft, ~All these systems utilize the fundamental ad- vantage of the bifunctional fuel- coolant — the elimination of a heat transfer stage within the reactor core. ~ Two series of shielded, full-size, circulating-fuel aircraft designs have been studied; one involves an annular, or "wrap-around,"”" type of heat ex- changer, and the other is a tandem arrangement with the heat exchanger behind the reactor. Analyses show that the annular arrangement gives the lower shield weight. These design studies brought forth the important technique of lacing the heat exchanger matrix with about 8 vol % B4C to keep the radiation from sodium or NaK in the secondary circuit to tolerable values even with something ~approaching a unit shield. It also appears from these studies that 1t would be advantageous to use a liquid moderator such as water or fused hydroxides. Not only would the problem of cooling the moderator and reflector be greatly simplified, but by using perhaps a 12-in.-thick re- flector, the problem of heating of the ()R, W. Schroeder, *“Circulating-Fuel Aircraft Reactor,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 10, 1951, ORNL-1170, p. 7. pressure shell by gammas, neutron captures, and inelastic scattering would be greatly reduced. More de- ‘tailed studies of the practicability of design for cooling the structure with a liquid moderator have been initiated. REACTOR WITH TANDEM HEAT EXCHANGERS The first basic configuration con- sidered was one in which the reactor core and the intermediate heat ex- changer were placed in tandem. A longitudinal section of a tandem re- actor and heat exchanger arrangement is shown in Fig. 1. A basic premise was that by interposing the heat ex- changer between the pumps and the re- actor core, a uniform flowdistribution among the fuel tubes would be assured. The design required that the liquid fluoride fuel flow through the tubes in the reactor core at 11 ft/sec for operation at 400,000 kw with a tempera- ture rise through the reactor of 400°F. The basic layout is best adapted to a liquid moderator (i.e., H,0 or NaOH) from the standpoint of fabrication and of means for cooling the moderator and reflector. In this design, the fuel enters the reactor at the top rear, makes a com- plete loop through the fuel tubes in .the core, and discharges to the heat exchanger. The fuel tubes are of stainless steel with 1 1/2-in. ID and 0.015-1n. wall thickness. If water were the moderator, a double-walled construction could be used. It may be noted that if the inner tube (con- taining circulating fuel) ruptured or cracked, the only fuel lost would be the amount that filled up the space between tubes before it froze in contact with the cold outer tube. e T T T R p T e e o ™ Gl =“SECREF DWG, E-Y-F7-47T1R{ APPROX. 42 in. H0 ARCUND HEAT EXCHANGER AND PUMP HOUSINGS Y, -in. STEEL PLATE FOR REACTOR AND SHIELD SUPPORT, 2 REQUIRED H,0 MODERATOR HEAT EXCHANGER INLET PIPE T . ACTIVE ‘ LATTICE H,0 MODERATOR PiPE OUTLET SECONDARY COOLANT LINES FUEL PUMP HOUSING i-in. THICK SHELL CIRCULATING AIR AND THERMAL EXPANSION GAP Y4 -in. THICK FUEL EXPANSION TANK, 4 REQ'D ¥ -in. THICK ALUMINUM '/Z-in. MINIMUM H,0 LAYER AT 300-350 psi AROUND ENTIRE REACTOR AND HEAT EXCHANGER CASTINGS Fig. 1. Tandem Heat Exchanger Arrangement for Circulating-Fuel Reactor. e LH0da¥ SSTY90Ud ATALIVND IDAf0dd ANV The water (or hydroxide) moderator enters the active lattice around the periphery of the rear of the reactor . and flows toward the outlet at the forward end of the reactor. As the moderator flows in it is distributed by a baffle sheet that is orificed "to give the flow distribution required by the power distribution within the active lattice. The control system for this proposed circulating-fuel reactor design con- sists of two curtains of cadmium rods mounted on two endless tracks. The curtains are moved from the reflector into the active lattice by an endless ~chain-sprocket type of mechanism. Each cadmium rod is mounted upon two steel shafts with steel rollers attached to the ends of the shafts. These rollers are linked together to form an endless chain. The sprocket 1is driven by a worm-gear drive, which in turn is driven by a hydraulic pump mounted within the pressure shell. Each cadmium rod is a cylinder ap-- proximately 1/2 in. OD by 1/4 in, ID by 12 in. long, and there are about 50 rods in each control curtain. REACTOR WITH‘A‘NNULAR’ HEAT EXCHANGERS The second basic configuration considered was a reactor with an annular, or "wrap-around," type of heat exchanger such as shown sche- matically in Fig, 2. Radial webs are used both to separate the annular heat exchanger into sectors and to support the shell around the outside of the reflector. To permit assembly, the reactor pressure shell would have to be split axially, probably on the center line. The pressure shell and other structures are jacketed and cooled by NaOH at around 1200°F, A heat exchanger to cool the NaOH is provided in the region where the NaK enters the pressure shell. Although not an essential element in this general FOR PERIOD ENDING MARCH 10, 1952 !type of design, the concentric helical coil carrying the NaOH moderator through the reactor core is interest- ing because it eliminates a header problem. By using perhaps ten tube fittings, welds at the ends of these coiled tubes could be avoided so that a material such as molybdenum might be used. REACTOR SHIELD DESIGNS It is important to know the effects of various parameters on the weight of the aircraft reactor and shield, but it 1s difficult to determine these effects quantitatively since the pres- ent knowledge of shielding does not permit weight estimates closer than +5%, at best. However, engineering designs have been completed for the shields for the preceding reactor arrangements, and real weight differences that result in lighter weight shields for the annular arrangement than for the tandem arrangement have been found to exist. In the shield weights cal- culated the use of B,C in the heat exchanger and between the reactor and " heat exchanger has been found to sub- stantially reduce the activity of the secondary coolant. Design Procedure. The primary objective in the design of these shields was to minimize the activation of the secondary coolant by delayed neutrons from the fuel. This has been effected by (1) attenuating the neutron leakage from the active lattice to a level below that of the delayed neu- ‘trons in the heat exchanger, and (2) lacing the heat exchanger with enough neutron absorbing material (B,C) so that relatively few neutrons would be absorbed in the secondary coolant. The header sheet and a 5-in. B,C layer between core and heat ex- changer was sufficient to reduce the neutron leakage flux to the heat ex- changer, whereas the use of B,C in U sy L i i o i s o e i s i ok i ER T - N ‘. \ . .-—I o SECREE : DWG. E-Y-F7-2i3aR1 - NOK ]NLET SIS NN o VNN SN AN NQK OUTLET Y» B4C CANNED WITH 0080 STEEL . / Yo NaOH I/*/2 STEEL R NaOH -NoK HEAT EXCHANGER { ] S ; HEAT EXCHANGER %8 OD TUBES ON %» ¢ - ¢ % —_ \ OUTLET MANIFOLD SQ PITCH HELICAL GOIL GOOLANT TUBES ,,,,’\ 1.250 OD TUBING \ N Y STEEL 27.000 | \ Al 60 ST | | g\‘ 5 B4C o 0080 STEEL =1t} %\\ }y STEEL £ 0 0O00BOI 6\ Y% B,C CANNED IN g d “ 2 Bg D 0 q DD = B 0.060 STEEL T O ‘\ 1 A g ?\ N b\ b b‘q ¢ 2 PRESSURE SHELL b 3?*95? oL 9@9@3 o 1 Lok NAOH PUMP E \ Y SN L, 4q9qqdqaqadag SN O ¢ QaQqqoaag S AN Z ’\\ 9 - AN Y% B,C CANNED IN Q O 0.080 STEEL Py '/2 NaQH % STEEL NeOH OUTLET TO NoOH-NaK HEAT EXCHANGER 10% CIRCULATING NaOH REFLECTOR O S R~ = RN o = W, s WY = ) TR LY = R \ Y T~ \‘\ WALy B3 ey 3 v, = \ \ N B MY g \\\'\\\\ = = Y S S B = 6 ANNULAR TYPE HEAT HEAT EXCHANGER INLET MANIFOLD HEAT EXCHANGER OUTLET — LINE, 6 REQ'D 3 O\ = EXCHANGER Y% 0D TUBES = o B 5 _ = X > NaK INLET \” =M—--= o /32; & 3 PcH = %y NaK OUTLET NOTE: DIMENSIONS ARE IN INCHES Fig. 2. Annular Heat Exchanger Arrangement for Circulating-Fuel Reactor. 140d3¥ SSIY90uUd ATHALYVAO 1Daf 0d4d dNV every ninth tube in the heat exchanger served to supress the activity of the circulating fuel. The balance of the shield design work was straight-forward and followed the procedures outlined by the Shield- ing Board.(?? Both Lid Tank and Bulk Shielding Facility test data were used. The fission-product-decay gamma rays for the equilibrium fission-product concentration at full power were taken as being equivalent to three 1.0-Mev gamma rays and one-half of a 3.0-Mev gamma ray per fission. Activation of the Secondary Coolant. The degree of activation of the second- ary coolant is, of course, a function of the materials 1in that coolant. Since NaK (56% Na and 44% K) has many advantages as a high-temperature heat transfer medium, and since 1t would probably be about as bad from the activation standpoint as any coolant that could be used, an estimate of its activation was made with the heat ex- changer shielded and laced with B,C as outlined above. It was found that 1f - 25% of the NaK in the system were considered as concentrated at a point and the self-absorption of the sodium- decay gamma rays released in the air radiator or other similar component (Z)Report of the ANP Shielding Board for the Air- craft Nuclear Propulsion Program, NEPA-ORNL, ANP-353, October 16, 1950. FOR PERIOD ENDING MARCH 10, 1952 were assumed to be 50%, the dose at 1.5 meters (5 ft) from the center of the source would be 200 r/hr after long periods of operation at a reactor heat output of 400,000 kw. The activation of a quite different coolant, a fused salt consisting of approximately 60% LiCl, 30% MgCl,, and 10% KCl, was also estimated. Al- though with no B,C in the heat ex- changer the activity of this fused salt was only 5% of that of the Nak, the activity for the case with the B,C in the heat exchanger was 35% of that of the NaK. Thus, it appears that there is little to be gained from the shielding standpoint in the use of this molten chloride in place of Nak. Shield Weights and Specifications. Several engineering designs of shields have been completed for both the tandem and annular heat exchanger arrange- ments. For either reactor arrangement the divided shield resulted in the lowest shield weight. However, the desire to be able to carry out a limited amount of airplane maintenance work without special shielding argues in favor of more shielding around the reactor. The various shield designs for the two reactor and heat exchanger arrangements are given 1in Table 1. For comparable shields the annular heat exchanger and reactor assembly gives appreciably lower weight than the tandem arrangement. 11 ¢l T G s i g TABLE 1 -Shields;for,Circuiating~Fluoridé-Fuel'Reactors - TANDEM HEAT EXCHANGER PARTIALLY DIVIDED ANNULAR HEAT EXCHANGER PARTIALLY DIVIDED 1 2 3 4 1 2 Details in report Y-F15-10 Y-F15-10 Y-F15-10 Y-F15-10 Y-F15-10 Y-FI5-10 Date of design Jan. 1952 Jan. 1952 Jan. 1952 Jan. 1952 Feb. 1952 Feb. 1952 Reactor shield diameter (in.) 150 150 121 121 148 118 Crew shield weight (1b) 5,000 11,000 36,000 14,000 5,000 36,000 Weight of reactor, intermedi- 151,000 130,000 75,000 75,000 123, 000 62,000 ate heat exchanger, and reactor shield (lb Total weight of reactor, in- 156,000 141,000 111,000 89,000 128,000 98, 000 termediate heat exchanger, : and all shielding (including crew shield) (lb% ' _ Reactor power (kw) 400, 000 400,000 400,000 400;000 400, 000 400, 000 Diameter of reactor core (in.) 32 32 32 32 32 Liquids in primary and second- ary circuits Temperature loss in ingermedi- ate heat exchanger (°F Pressure loss in intermediate heat exchanger (psi) Crew shield size (ft) Beactor-crew separation distance (ft) Radiation inside crew com- partment (r/hr) Radiation 5 ft from center of reactor {(r/hr) Radiation 50 ft from center of reactor (r/hr) . Radiation 300 ft from center of reactor (r/hr) Fluoride-NaK 100 100 6% X T% x 12% 50 300 32 Fluoride-NaK 100 100 6% X T4 x 12% 50 2,400 36 Fluoride-NaK 100 100 6% X % x 12%° 50 1 350,000 | 5,600 156 Fluoride -NaK 100 100 5 X5 x 12% 120 1 380, 000 5,600 156 .Fluoride-NaK 100 50 6% X T% x 12% 50 300 Fluoride-NaK 100 30 6% X Th x 12% 50 380, 000 5,600 156 1Y0d3¥ SSAYO0uUd X THIALMVND lDHrOHd dNV 'FOR PERIOD ENDING MARCH 10, 1952 2. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT R. W. Schroeder, ANP Division E. S. Bettis, Reactor Projects Division - The studies of the circulating-fuel aircraft reactor using fused fluoride have culminated in the design of two aircraft reactor systems (one with and one without an intermediate heat ex- changer), both of which are potential power plant systems for supersonic nuclear propulsion, Furthermore, these aircraft reactor systems appear to permit higher performance than the sodium-cooled reactor, primarily be- cause of the bifunctional capacities of the primary coolant, which is the circulating fuel. Consequently, the ~ ARE to be constructed by the Oak Ridge National Laboratory at Qak Ridge will be a circulating-fuel type of reactor, The circulating fuel of the reactor will be a mixture of molten alkali metal fluorides plus uranium fluoride, - The structural metal of the core and container shell will be Inconel, since this metal has been on order for the ligquid-metal ARE and currently appears at least as corrosion resistant to the circulating fluorides as any of the stainless steelsor The reactor c¢an advantageously be moderated with beryllium oxide, which is also cur- rently available for the previously contemplated liguid-metal ARE, The power is generated within the circulating fuel as it is passed through the beryllium oxide moderated core. The core is provided with a beryllium oxide side reflector, which is housed with the core within an Inconel pressure shell. The reflector and pressure shell are cooled by a separate circuit using a mixture of nonuranium-bearing fluorides, Power is abstracted from the fuel by means of four fuel-to-helium heat exchangers through which the fuel is circulated. The heliumis cooled by passage through four helium-to-water heat exchangers and the water - the ultimate heat source — is discharged. Control of the reactor is provided in three forms, (1) shim control, (2) regulating rods, and (3) safety rods, Shim control is achieved by varying the uranium concentration in the circu- lating fuel. One boron rod, which passes through the core center lineand effects approximately 0.75% Ak/k, serves as the regulating rod. The boron safety rods, equally spaced about the.center of the core, have approximately 5% Ak/k per rod. CORE DESIGN The core design is illustrated by Fig. 3. It may be noted that the over-all assembly includes an Inconel pressure shell in which beryllium oxide moderator and reflector blocks are stacked and through which fuel tubes, reflector coolant tubes, and control assemblies pass. The innermost region of the lattice is the core, which is a cylinder 3 ft in diameter and 3 ft long. The core is divided into six 60-degree sectors, each of which in- cludes 13 vertical stacks of hexagonal beryllium oxide blocks. FEach sector includes one serpentine, fuel-tube coil, which passes through the 13 moderator stacks in series, as 1l- lustrated., The six serpentine coils are connected in parallel by means of external manifolds. A reflector with a nominal thick- ness of 6 in. 1is located between the core and the pressure shell on the cylindrical surface only. The re- flector consists of beryllium oxide i’ "‘:’ 13 Wl g Y e & :6}' £y e PENE .. EhE s gdst: gl b mpldh Bk So ot e podt e G LSl SER R G i ESArr iR i e L byl e e e s s e g il oo i o s i b R ‘__. . fiSREfl‘ = _ DWG. E-Y-F7-202 Ri WELD BACK-UP RING BOTTOM g?SE c%Fr,aFéORT : CORE AND REFECTOR COOLANT INLET (SHOWN ROTATED 30 DEGREES - FROM TRUE POSITION ) REINFORGING PLATES e FUEL OUTLET s MANIFOLD o ASSEMBLY ~ £ CONTROL ROD ‘ BLY ASSEM THERMAL INSULATION CORE AND REFECTOR COOLANT QUTLET PIPE { ROTATED 30 DEGREES FROM TRUE POSITION FUEL INLET MANIFOLD 1Yg FUEL TUBE WELDING JOINT FUEL TUBES 1.235 0D, 0.060 WALL BOTTOM TUBE . SHEET, 4 THICK TOP HEAD TOP TUBE SHEET CORE AND REFLECTOR CAN, Y1g THICK COOLANT ANNULUS, REFLECTOR COOLANT TUBE g THICK 0500 0D, 0.020 WALL NOTE: ALL DIMENSIONS IN INCHES 35.250 ACTIVE LATTICE 2 7.500 34.500 - 1 | 48.500 Fig. 3. (Circulating-Fuel ARE Core Design. Riin e S ol oo o i e sl ity bERE { ;o 140434 SSAYI0Nd ATHALUVND IDAC0Ud dNV wd .} 8 blocks similar to the moderator blocks but with}-in, holes for the passage of reflector coolant, The reflector coolant enters at one end of the pressure shell, passes through the reflector, bathes the pressure. shell and fills the moderator interstices, and exits at the other end of the pressure shell, Primary Coolant Circuit, Hydro- dynamic and thermodynamic considera- tions have influenced the design to such an extent that a discussion of these problems appears to be desirable prior to description of the primary coolant circuit itself, In a circu- lating-fuel type of reactor, heat 1is generated in the flowing fuel. The particle heat generation is dependent on the particle residence time within the reactor., As the flow proceeds through the tubes, the velocities adjacent to the walls are lower than the bulk mean velocities, so that the fuel particles adjacent to the walls are subjected to higher temperature rises., 1Tlhis effect is mitigated by convection within the fuel stream, but a temperature difference will exist between the wall and the stream center line, Studies have indicated ~that this temperature difference is directly proportional to the fuel power density, tube diameter, Prandtl’s ~number, and Reynold’s number. Appli- cation of these studies to various postulated ARE configurations have established that excessive temperature differences are encountered in the laminar flow region (Reynold’s number below 5000) but that small temperature differences are encountered in the turbulent flow region. Consequently, the reactor core was designed for a Reynold’'s number of 10,000, thereby allowing for some deviation of flow rate or fluid physical properties. The design for a straight-through flow core arrangement was revised because the calculated Reynold’'s FOR PERIOD ENDING MARCH 10, 1952 number was less than 1000, Con- sideration of the parameters for in- creasing Reynold’s number indicated that the desired correction would entail increasing tube diameter, in- creasing the number of tubes in series, increasing the volumetric flow rate, or decreasing the volume of fuel with- in the core, or combinations thereof. Investigation of these variables led to the observation that a few series passes would be required even with maximum feasible exploitation of the other variables. Since this was the case, and since the design problems involved in the use of many series passes appeared to be similar to those associated with the use of a lesser number of series passes, it was decided to use 13 series passes and avoid compromising the volumetric flow rate or the core fuel volume. Secondary Coolant Circuit. The use of a liquid circuit other than the fuel circuit within the pressure shell is desired for the following reasons: (1) to maintain the pressure shell essentially isothermal at a tempera- ture of approximately 1150°F; (2) to cool the beryllium oxide reflector with a fluid other thanm circulating fuel:; and (3) to fill the beryllium "oxide moderator interstices and other internal voids with a liquid maintained at a pressure higher than the fuel pressure to prevent a fuel tube leak from adding reactive material to the core, The ligquid to be used has to be compatible with the structural material under dynamic conditions and should cause no undesirable reaction with the fuel in the eveht of a leak in the separating wall. At the present time it is believed that nonuranium-bearing fluorides, perhaps the fuel carrier without the UF,, will best meet these specifications, The possibility of cooling the moderator by positive flow of this fluid through the moderator interstices 15 ANP PROJECT QUARTERLY PROGRESS REPORT has been studied. Yt has been found, however, that the mass flow through a gapof constant periphery varies as the third power of the gap width (in the laminar flow region, with fixed pressure drop), Temperature analyses - with postulated gap width distributions indicated that gross temperature inequalities could exist throughout . the moderator because of tolerance accumulations affecting interstice widths. Accordingly, it has been decided to obstruct flow through the moderator so as to render it virtually stagnant. The moderator heat then is conducted to the fuel stream, and thus a relatively accurate analytical determination of moderator tempera- tures can be obtained. Core Temperature Distribution. The circulating fuel in the ARE core flows in parallel through six serpentine “tubes, each tube traversing the core 13 times. In passing through the core 13 times, the fuel is heated from 1150 to 1500°F., The calculated mixed mean fuel temperatures at various stations in the reactor are given in Table 2. | ' | TABLE 2 o Fuei'Temperaturé after each Pass Through Core _ MIXED FUEL STATION TEMPERATURE (°F) Entering reactor 1150 Leaving pass No. 1 1171 | ' 2 1191 3 1211 4 1233 -5 1257 6 1281 7 1305 8 1335 9 : 1365 10 1392 11 1426 12 1462 13 1500 16 There are two sources of heat for the fuel circulating through the reactor: 1nternal heat generation (from fission and gamma-ray absorption) and heat transferred to the fuel from the remainder of the core. The heat transferred from the remainder of the core 1s produced in the following manner. Heat is generated in the moderator and the parasitic core material as a result of gamma-ray absorption, and additional heat 1is generated in the moderator by neutron slowing down. All this heat is trans- ferred to the circulating fuel and results in a temperature gradient across the fuel, moderator, and parasitic material, EXTERNAL FLUID CIRCUIYT The ARE fluid circuit, intended to handle toxic and corrosive fluids at 1500°F, requires a considerable amount of design and developmental effort. ‘Where possible, commercially available components are used, but pumps, heat exchangers, and certain other com- ponents are being constructed es- pecially for the ARE. The location of the major system components, including the reactor pits, the heat exchanger room, the reactor, and the two heat disposal loops, is 1llustrated in Fig. 4. It may be noted that the heat exchanger room is shielded from the reactor pits to permit servicing after fuel drainage and flushing. The "shield was designed to minimize activation of fluid circuit structure by reactor neutrons during power operation and to attenuate reactor post- shutdown gamma rays to a level permitting access to fluid circuit components after shutdown., Pumps. The pumps to be used in the fuel and moderator coolant circuits are vertical-shaft, tangential-dis- "charge, centrifugal pumps in which a B T T LT UNCL ASSIFIED DWG. D-Y-F7207R4 ; ‘.:".‘ .:. :. - : Y .... 5 K . . i :'_ it ._.. S _., .: ° T W INERT SALT COOLING SYSTEM NO. 4 He TO H,0 EXCHANGER PRIMARY COOLENT SYSTEM NO. 1 SOURCE —— - — . — g —— S— EXCH. » TEST PIT | } } £ < HEAT = He TO H,0 EXCHANGER l w ROOM 4 o { ! i | | it | s I | | | COOLANT TO He EXCHANGER PRIMARY COOLENT SYSTEM NO. 2 LEGEND — e o e COOL ANT FROM SOURCE COOLANT TO SOQURCE —— ———— 1 — //== INERT SALT FROM SOURCE o % /=== INERT SALT TC SOURCE S 00 00— HELIUM FROM RODS S o O == HELIUM TO RODS @ — W ——— WATER IN a ——wo wo ~— WATER OUT v —_——— OIL N S —— — — — 0L OuT o ROD COOLING BLCWERS INERT SALT COOLING SYSTEM NO. 2 - Fig., 4. Arrangement of External Fluid Circuit Equipment. ‘0T HOYVI HNIGNA doIydd Hod cS61 T N P T R T R T Ty T T TR ET R - ANP PROJECT QUARTERLY PROGRESS REPORT gas seal 1s used and the liquid level is maintained. The gas seal is formed by a floating graphite ring between the stationary nose on a bellows seal and the hardened rotating nose on the shaft. The primary pump seal carries no appreciable pressure differential and, consequently, requires that the oil-circulating system above the seal be pressurized to balance that of the circulating system below the seal. The ‘circulating oil serves to cool the shaft, bearing, and lower seal. In addition, internal cooling of the structure is obtained by the circu- lation of helium. Another feature of the pump is that the entire rotating assembly can be put together outside the pumping casing, and in event of a failuqe it can be installed with comparative ease, fieat Exbhangérs. The fuel-to- ‘helium heat exchangers are on order with the Griscom-Russell Company; they were designed by Griscom-Russell to ORNL specifications. The heat ex- changers are of the cross-flow type with the fuel flowing through 1-in.-0D, 0.109-in. wall Inconel tubing. The tubing is finned on the gas side with 'stainless steel, helically wound " strips. Each heat exchanger includes five parallel tube banks; each bank consists of seven to nine horizontal tubes in series with integral return ~bends. The fuel inlet and outlet headers are constructed of 4-in. pipe to which the five tube banks are Weldedf - The helium-to-water heat exchangers also are finned-tube, cross-flow ex- changers, but steel tubes with copper fins are used. Primary Coolant System. Fuel flows from the six core tubes to a common header, and a common trunk line conveys the fuel into the heat exchanger room. The fuel then divides into two com- pletely independent loops, each 18 capable of dissipating 1500°F kw, Within each loop the fuel divides into two parallel 750-kw fuel-to-helium heat exchangers and then passes through a centrifugal pump to a common return trunk (Fig. 4). In an individual loop, helium flows from a centrifugal blower through a fuel-to-helium exchanger in which the temperature of the helium is increased from 250 to 750°F, The helium is then cooled to 250°F in a helium-to-water exchanger from which it passes through another heating and cooling cycle and is returned to the blower. This so- called "double-sandwich" arrangement permits two helium heat transports per cycle and thereby halves the helium flow rate required through the blower and ducting for a specified power and temperature rise, ' ' Secondary Coolant System. The reflector and pressure shell coolant system, as 1in the case of the fuel system, is divided into two heat disposal loops, each capable of handling the load associated with 1500-kw reactor power, Heat 1is conveyed from the primary coolant to helium and then to the water sink, Monitoring Circuit. All lines and components containing fuel or reflector cotlant are double-jacketed, with helium passing through the annulus, The helium pumping head 1is maintained by drawing helium from the system, cooling 1t, and admitting it to rotary- type compressors at various poilnts in the system as indicated. Monitoring for fluoride leaks into the helium is achieved by passing helium samples through halogen detectors, PREHEATING SYSTEM The relatively high melting points of the circulating fuel and secondary coolant (around 752 to 932°F, depending T I T TR T o At S o TR T [y F L) N ” -1y upon the particular fuel composition) require that all equipment within which these coolants are to be main- tained be preheated to permit loading and unloading, "Reactor Assembly. The insulation heat leakage at operating temperature has been calculated to be approximately 15 kw, and the heat capacity of the assembly has been established as approximately 4000 Btu per °F. After having studied various preheating methods, including pumping hot fluids through the fuel passages or through the moderator interstices, it was decided that the simplest method entailed the application of electrical resistance heaters to the outside of the pressure shell, withheat transport to the inner regions by conduction. With a constant heat input of 14 kw, in addition to the insulation losses, it is estimated that the assembly can be heated from room temperature to 1200°F in approximately four days. With this heat input rate, certain calculated temperature gradients are as follows: 1. Across pressure shell, 20°F, 2. Across helium gap between shell inside diameter and reflector outside - diameter, approximately 100°F, 3. From reflector outside diameter to core center line, approximately 250°F. Ppiping. The heat capacity of the ~ piping is small relative to the piping ~insulation leakage. Consequently, the power required to preheat the piping "is substantially the same as the power required to maintain it at the ultimate temperature. If the liquid-bearing pipes are empty, as is the case during preheating, the pipe resistance to axial heat flowis high, and the appli- cation of heat to specific points on FOR PERIOD ENDING MARCH 10, 1952 the pipe surface tends to cause large temperature valleys to exist between the points of heat application. By attaching electrical resistance heaters to the outer pipe of the double- ‘walled piping, however, the flowing helium acts as a thermal conveyor and facilitates axial heat transport, Accordingly, substantially isothermal helium contacts the inner pipe, and it is thus possible to heat the pipe uniformly. Heat Exchangers. The fuel-to- helium heat exchangers require pre- heating prior to filling and the addition of heat to compensate for thermal leakage during filling or at any time after filling when the fuel pumps are inoperative. These heat exchangers have a total of approxi- mately 40 ft? of exposed free-flow area (both upstream and downstream), which, if allowed to radiate to the adjacent cold structure, would dissi- pate approximately 40 kw when the tubes are at 1325°F. Accordingly, it appears necessary to include radiation barriers in the form of gates that can be lowered during warm-up or zero- or low-power operation, When these radiation barriers are in position, the fuel-to-helium heat exchangers are enclosed and may be preheated by means of hot helium supplied by the external piping annuli. ELECTRICAL POWER CIRCUITS Because of the extreme incon- veniences associated with forced reactor shutdown, fuel drainage, re- filling, and restarting, the objective has been to design the system so that no one failure will necessitate a forced shutdown. Critical pumps, blowers, etc. are duplicated, and each set is served by an independent d-c circuit that includes independent 19 T ANP PROJECT QUARTERLY PROGRESS REPORT buses, switches, and a-c-d-c motor- generator sets, The use of direct cur- rent for these circuits permits con- venient speed control where required and the use of battery sets floating on the line to safeguard against outside power failures. Alternating-current instruments are fed from the d-c circuits via d-c-a-c motor-generator sets so that these instruments will derive their power from the batteries in an emergency. - The heat capacity of the system 1is large relative to the heat leakage rate, so that no large temperature loss will be incurred for about 6 hr in an emergency. Consequently, it 1is not necessary to use battery power for system heat addition. System heaters, therefore, are connected directly to the 440-v a-c¢ circuit. Similarly, the building crane and certain other items will not need to be operative during any short period in which it is neces- saTy to use battery power, so these components are supplied by the 440-v a-c system. ‘ ' CONTROL SYSTEM ‘The ARE is controlled by three essentially separate systems: (1) regulating, (2) shim, and (3) safety. A mechanical regulating rod is pro- vided, since the time-constants of the self-stabilizing effects of the fuel expansion 1in the ARE are too long to provide stiff control of the reactor. Shim control is conveniently effected by the addition of higher-uranium concentration fuel to the circulating- fuel volume. Specifications for the control console and panel have been released with a specified delivery date of October 1, 1952, The high-temperature fission chamber has operated at 1292°F, and an MIR-type servo for the regulat- ing rod has been ordered., The kinetic 20 behavior is currently being analyzed on the analogue computor. Shim control.. Fuel enrichment will be accomplished by adding enriched molten fuel at the surge tank. The molten concentrate will be introduced through a small (%-in.) line, whach, by means of a video pickup device, can be observed from the control room while fuel additions are being made, The fuel-enrichment mechanism has been laid out in elementary form. The principles of operation have been established, but the tanks, weighing devices, valves, heaters, etc. have not been detailed. This system 1is much simpler than was first thought possible and it makes possible a safe and relatively easy technique of bringing the reactor to critical after the initial loading. Regulating and Safety Rods. Because of the low power density of the circu- lating-fuel ARE, the temperature coefficient of the fuel must be supplemented by a servo-actuated regulating rod if transient conditions are to be controlled, This servo will operate in exactly the same manner as the fuel temperature coefficient. The error signal that actuates the servo will be a mixed signal of temperature and neutron flux; it 1s given by the equation ' €= (8, -6, ;) *58.5(p-py, T where € = error signal, O, -6, ¢~ reactor inlet error in OF, (p - pO) = reactor power error in megawatts (the factor 58.5 is in °F per mega- watt).' : i) When € < 0 the servo withdraws the rod, and when € > 0 the servo inserts the rod.” This servo equation is different from that previously reported as a result of a change in reactor design. A change has been made in the location of the drive mechanisms for the regulating and safety rods. They will not be located over the reactor pit, and they will have straight linkages to the rods. This change will allow considerable simplification of the over-all mechanical control system, | An MTR-type servo has been ordered for the regulating rod. The regulat- ing rod and safety rod designs have been determined and the fabrication of " these rods has been turned over to the Metallurgy Division. After fabri- cation, the rods will be run in the "cold" critical assembly to determine their worth, Calculations indicate that the safety rods are worth approxi- mately 5% Ak/k, and the regulating rod will be loaded so as to be worth about 0.75% Ak/k. One safety rod operating mechanism has been finished by the machine shop. This assembly has not been tested, but the design seems to be satisfactory. High-Temperature Fission Chamber. The fission chamber has been operated at 700°F but no insulator material has been found that will withstand higher temperatures. A design is in progress that eliminates the insulator in the high-temperature region. This chamber has not been tested and there is no certainty that 1t will function properly. Because of the uncertainty of high-temperature operation of the fission chambers, sufficient cooling is being provided to maintain the chambers at 400°F, where it 1is known they will function. FOR PERIOD ENDING MARCH 10, 1952 Control Console and Panel. Complete specifications for the components of the control room have been released, including detailed construction draw- ings and fabricational and material specifications., Sets of these specifi- cations have been mailed to ten pros- pective bidders, and the bids are to received by March 1. The instructions to bidders specified that the complete order is to be filled by October 1, 1952, The items covered in these drawings and specifications include an operating console, instrument racks, relay racks, recorders for nuclear measurements, amplifiers, power supplies, and assorted mounting hardware, Items not covered in this outside order are nuclear chambers, some preamplifiers, servos, and process instruments. The order takes care of from 90 to 95% of the electronic equipment needed for the ARE. Reactor Dynamic Computer. The entire circuit of the ARE has been put on the analogue computer, and the kinetic behavior of the system 1is being analyzed., The analysis is not yet complete, but the work 1is pro- gressing satisfactorily and indi- cations are that a fairly compre- hensive study will result from this work, This computer analysis has received the full attention of two engineers for the past eight months., INSTRUMENTATION A basic purpose of the ARE is the acquisition of experimental data, so the importance of complete and reliable instrumentation cannot be over- emphasized, Most ARE process in- strumentation is intended to observe and record rotational speeds, flow rates, temperatures, pressures, oOr ligquid levels. 1In the low-temperature 21 -~ e e g T ANP PROJECT QUARTERLY PROGRESS REPORT loops, this equipment is sufficiently conventional to obviate the need for detailed descriptions., The high- temperature loops involve special sensory problems, however, which may merit some discussion, In some instances several alternate sensory principles are under development con- currently, No attempt will be made to describe each of these alternate methods in detail in this report; “however, the various technigues that are being used or developed are out- lined in the section on "Experimental Reactor Engineering" (sec., 3). 22 BUILDING The building to house the ARE is proceeding on schedule., Additional contracts have been negotiated with the contractor to complete work not specified in the original contract, and this new work is to be completed so that the building can be released to OBNL by June 1, 1952. The auxiliary power specifications have been com- pleted and orders for the equipment are being released. An elementary electrical diagram for this equipment has been drawn, Y FOR PERIOD ENDING MARCH 10, 1952 3. EXPERIMENTAL REACTOR ENGINEERING H. W. Savage, ANP Division Liquid metals, hydroxides, and mixed fluorides are being investigated as heat transfer media and fuels for aircraft reactor experimentation at temperatures of 1200 to 1800°F. Some developmental effort on the use of sodium and sodium-potassium alloy continues, although the effort on the technology of molten fluoride mixtures, as required by the circulating-fuel reactor, predominates. The develop- ment of fluoride systems 1s limited by the corrosion problem (see sec. 11 "Corrosion Research”). Chemical and _physical treatment of fluoride com- ponents to eliminate contaminants and improving handling, storing, and transfer techniques to avoid reintro- duction of contaminants are being studied as means of limiting corrosion. The applicability of known methods of pumping, sealing, controlling, and measuring properties and quantities of these high-temperature coolants and fuels is being investigated. The high- temperature reactor systems, however, place unique restrictions on materials, lubricants, leakage, and performance of mechanical devices and associated equipment. Pumping has been accom- plished with conventional hydraulic designs, but alleviation of thermal distortions and stresses, cooling of bearings, and the development of liquid- and gas-tight seals have been required. Alleviation of thermal dis- tortions and the development of liquid- -tight seals have also been required for valves. 1In addition, the valves must contain seal materials that will not interdiffuse in the presence of the coolant at high temperatures. Lubrication of the moving parts of these devices at high temperatures has been accomplished. The cumbersome, and at times massive, geometry of these devices has made it necessary to provide auxiliary heating to avoid freezing of the coolant. Heating and cooling of liquid metals, hydroxides, and mixed fluorides are being investigated from room temperature to 1800°F. The high melting temperatures of hydroxides and fluorides have introduced pre- heating, insulation, and operational complexities, since 1t is desirable to avolid freezing of the coolant and possible bursting of containers upon remelting. Heat transfer studies at temperatures at which the materials are molten thus far appear to be straightforward, and the technology advances at the rate at which the controlling physical properties are defined. Equipment performance 1s improving markedly, and a number of mechanical and other devices for fluorides have been operated for periods exceeding 1000 hr in some cases without visible equipment damage and without leakage. Heat exchangers are being designed for aircraft and laboratory applica- tion, and the NaK-to-NaK heat exchanger has now operated for over 2000 hr at 1500°F. Flow, volume, pressure, and temperature control appear to be straightforward 1f care 1s exercised at the high temperatures involved. SEALS AND CLOSURES Frozen-sodium-sealed pumps have operated over extended periods, and frozen fluorides that have been used for sealing shafts show considerable promise of giving satisfactory service. The metallic braid with self-contained packing lubricant appears to be highly 23 T i g st ANP PROJECT QUARTERLY PROGRESS REPORT satisfactory for use with molten fluorides. Controlled isolation of two sections of a fused salt system has been accomplished by the use of a "freeze valve." Welded or metal- gasketed joints are used in plping systems. Improved weld designs and welding techniques that allow full penetration of weld metal have made possible the operation of equipment without failure for extended periods. Flanged joint seals with oval-ring gaskets have proved satisfactory for operation at temperatures to 1300°F. Frozen Sodium Seal. The frozen- sodium-sealed pump regorted in the last quarterly report‘!’ has operated "approximately 1500 hr during this quarter without failure of the frozen sodium seal. A modified Durco cen- trifugal pump has been constructed that has a finned sleeve for forming a frozen sodium seal by means of con- vective cooling. Frozen Fluoride Seal (W. B. McDonald and P. W. Smith, ANP Division). An additional 160 hr of testing was ac- complished during this period with the frozen fluoride seal previously re- ported(?? (Fig. 5). 1Initial tests were conducted to determine whether a frozen fluoride seal 1s feasible, and additional tests were conducted to determine (1) the effect of fluorides on the Stellite-coated shaft around which the seal is formed; (2) the maximum pressure that can be sealed without leakage; (3) the problems encountered in starting the shaft after it has been stopped and the fluorides permitted to freeze around the shaft; and (4) the determination of design and operational parameters with this (1)H. W. Savage, ‘“Experimental Reactor Engineering,”” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 10, 1951, ORNL-1170, p. 42. (2)W. B. McDeonald, ¢‘‘Seal Testsfi; op. cit., ORNL-1170, p. 43. ' 24 type of seal. These tests 1indicate that greater clearance is desirable between shaft and cooling sleeve than was used in the frozen sodium seal. A build-up of magnetic, metallic material, from 1 to 2 mils thick, was found on the shaft and cooling sleeve. Several shallow surface scratches were found on the shaft; however, these were no more severe than those normally encountered with stuffing-box packing materials. A significant difference was found between a frozen fluoride seal and a frozen sodium seal; when rotation 1s stopped and fluorides are permitted to freeze around the shaft, it is necessary to apply heat to the shaft in order to free it for continued operation. Maximum pressure limits for this seal were not determined; however, the seal was operated agalnst 60-ps1i pressure, which 1is satisfactory for the operation of the ARE. Bellows Type of Face Seal for ARE Pump. Sealing of the ARE pump 1is accomplished by two bellows type of face seals installed in series. The seal below the bearing space consists of a graphite ring floating between two hardened, metallic sealing faces, one on the nose of the bellows and the other attached to and rotating with the shaft. Sealing against the atmosphere 1is accomplished by the conventional lapped face seal above the bearings. This arrangement permits the lower seal, which must operate at higher tempera- tures, to operate with no pressure differential across it and thus lengthens the life of the seal. This arrangement would also permit complete isolation of the system from the " ambient atmosphere in the pump room if leakage should occur in the primary seal. The temperature of the upper high- pressure seal can be easily controlled &7 A 1) - and should present no problem. Any leakage of bearing lubricant into the system would be trapped immediately below the seal and could be easily removed from the system. SHAFT (316 STAINLESS STEEL, STELLITE COATED) FOR PERIOD ENDING MARCH 10, 1952 Stuffing-Box Seals for Molten Fluorides (H. R. Johnson, ANP Divi- sion). Thin-walled bellows have thus far been unsatisfactory for sealing against high-temperature flucrides UNCLASSIFIED DWG.14424 SUPPORTED BY OUTBOARD BEARINGS AND DRIVEN BY V-BELT THERMOGOUPLE THERMOGOUPLE THERMOCOUPLE THERMOCOUPLE — LEVEL INDICATOR ¥4 THERMOCOUPLE N — — — U WELL — 0.060-in. CLEARANGE FINNED SLEEVE, TYPE-346 STAINLESS STEEL SPACER GAS PRESSURE THERMOCOUPLE GAS PRESSURE APPLIED TO THIS SURFACE POT, TYPE—316 STAINLESS STEEL " Fig. 5. Frozen Fluoride Seal Tester. 25 T T T ST T WY ANP PROJECT QUARTERLY PROGRESS REPORT owing to metal and weld failures. A stuffing-box seal for a valve stem has been developed that will furnish a positive seal against fluorides at 1500°F. Such a seal has operated satisfactorily and has sealed against fluid pressures to 50 psig. The seal consists of a conventional type of stuffing box in which successive layers of Inconel braid, graphite, nickel - ‘powder, and another layer of Inconel braid are packed under compression around a shaft. A conventional stain- less steel globe valve with the stem packed in this manner has operated during this period for more than 700 hr at 1500°F and 30-psi pressure and is still in operation, " The molten fluorides pump described in the following is sealed with a con- ventional stuffing-box arrangement by using four rings of commercial-graphite- impregnated asbestos packing separated by Teflon washers. The washers are machined from Teflon bar stock, since it has been determined that sheet Teflon formed by extrusion or rolling processes has a "memory" for its pre- extruded shape and at elevated temper- atures tends to return to this shape. ~ The stuffing box is surrounded by a cooling coil through which refrigerated ethylene glycol is circulated. Lubrication of Seals and Shafts (H. R. Johnson, ANP Division). Tests indicate that when the valve stem seal described above is lubricated periodi- cally (a few drops each week) with tricresyl phosphate (a compound used for lubricating wire-drawing dies), friction is reduced to the point where this high-temperature valve can be operated with as little friction as a standard valve operated at room tem- perature. Further tests show that when such a gland is packed with Inconel braid and finely powdered UF, is substituted for the nickel and graphite powders, a 26 shaft can be rotated at 1000 rpm with the entire assembly heated to 1500°F with no damage to the shaft. According to the Celanese Corpora- tion of America, extensive tests indi- cate that tricresyl phosphate has approximately six times the film strength of a lubricating oil of ap- proximately the same viscosity. Con- tinuous heating at well over 200°F in in the presence of air is required to bring about decomposition. "The lubri- cating effect of tricresyl phosphate on iron is believed to be due to forma- tion of iron phosphide Fe,P."(37 A valve stem lubricated periodically (a few drops once a week) with tricresyl .phosphate has operated 660 hr, and it remalns very easy to turn. PUMPS Development of a mechanical pump has progressed to the point where reliable operation can be expected in forced-circulation systems containing either liquid metals or molten fluo- rides with fluid temperatures in the pump as high as 1300°F. Designs have been completed, and fabrication of pumps, which are expected to operate reliably at 1500°F or above, is under way. All pumps tested to date are of laboratory size; however, pumps de- signed to meet ARE flow and head re- - quirements have been designed and are to be fabricated. ARE Centrifugal Pump (W. G. Cobb, G. F. Wislicenus, J. F. Haines, and A. G.Grindell). The ARE fluid circuit consists of two pumps in the circu- lating-fuel system and two pumps in the moderator coolant system., The head and flow requirements for each of (3)3. J. Bikerman, Surface Chemistry for Industrial Research, p. 369, Academic Press, 1948. . » these pumps are 1dentical. Each must deliver approximately 45 gpm with a head of approximately 45 ft of fluid ‘pumped. Two models have been proposed and designed for ARE service. The ARE centrifugal pump, Model D, requires a gas seal, whereas ARE cen- trifugal pump, Model F, has a frozen fluoride seal. The design features of Model D were described in the last quarterly report. The assembly of this pump is shown in Fig. 6. The Y-12 shops are presently in the process of fabricating three Model D pumps from type-316 stainless steel. The Model F frozen-fluoride-sealed pump is almost identical to the Model D pump, with the exception of the inclusion of a gas-cooled section in which the fluo- rides surrounding the shaft will be frozen. The principle of the frozen fluoride seal has been described in the section on "Seals and Closures" as well as in the two previous quarterly ~ reports. Laboratory Frozen-Fluoride-Sealed Centrifugal Pump (W, G. Cobb and P, W. ‘Taylor). The laboratory size cen- trifugal pump described previously(*’ ‘has been used to circulate molten fluorides in an isothermal loop op- erating at temperatures up to 1300°F for more than 500 hours. Although some shutdowns have occurred because of leaks in the plumbing system, no indi- cation of failure has been encountered in the pump itself. This pump has been modified to incorporate a Graphi- tar, rotary face seal in place of the graphite-impregnated-asbestos stuffing- box seal, and the parting faces of the pump casing have been moved to a point above the liquid level in the pump. These modifications are expected to make it possible for this pump to (4. AircraftNuclear Propulsion Project Quarterly Progress Report for Period Ending June 10, 1951, ANP-65, p. 168. - FOR PERIOD ENDING MARCH 10, 1952 circulate fluids at temperatures 1in excess of 1500°F. Worthite Frozen-Sodium-Sealed Pump (W. B. McDonald and W. R. Huntley). The Worthite centrifugal pump modified to incorporate the frozen-sodium seal, testing of which was described in the last report,‘'’ has been transferred from an isothermal loop to operation in a figure-8 loop. This pump has operated for a total of approximately 1500 hr with no seal failure. Occa- sional shutdowns for repair have been caused by the mass transfer of copper from the gasket used to seal the part- ing faces of the pump to the shaft in the cold section of the sealing annulus. This resulted in a build-up on the sleeve, which filled the annulus and froze the shaft. The copper gasket has been replaced by a soft nickel gasket and the mass transfer condition has apparently been remedied. Maxi- mum sodium temperature in this loop is 1500°F and the sodium temperature at the pump is between 1000 and 1100°F. During the total operating time no difficulty has been experienced 1in starting the cold pump from stand-by conditions. As soon as molten sodium is lifted from the sump to the pump the shaft rotates freely. Some bear- ing failures have been encountered owing to the proximity of the bearing housing to the high-temperature pump casing. Further modifications have been designed for moving the bearing housing away from the pump and also for furnishing better lubrication to the bearings. Such a pump 1s expected to operate at 1500°F. Modified Durco Centrifugal Pumps. Three Durco centrifugal pumps are currently being modified for the forced circulation of fluorides or sodium. The sodium pump will 1ncorporate a frozen sodium seal, and the two fluo- ride pumps will be modified, one to a frozen fluoride seal and the other to a stuffing-box seal. 27 T o e 8¢ Shaft, type-316 stainless steel. Shaft liner spacer, any 300-series stain- less steel. "O" ring, material to be recommended by seal manufacturer. Impeller, type-316 corrosion-resistant steel casting. _ Pressure sleeve, any 300-series stainless steel. Seal face ring, Ontario steel. Bearing housing, type-316 corrosion-re- sistant steel. Fig. 6. ARE 10. 11. 12. 13. 14. 15. UNCLASSIFIED DWG. ESK—15329aR Shaft liner, any 300-series stainless steel. Single-row radial ball bearings. Slinger, type-316 stainless steel. Special graphite ring, alternate with seal. arrangement. Gasket, copper. Pump body assembly. Impeller casing assembly. Bellows shaft seal. ' Centrifugal Pump. 190434 SSAY90Ud ATHALYVAO 1JAf0ud dNV -y “ -l Construction is under way ona type- 316 stainless steel Durco centrifugal pump having a conventional stuffing box packed with Inconel braid and a lubricant similar to that described in the first part of this section. The portion of the shaft passing through the stuffing box is hard-faced with a layer of Stellite 1/16-1in, thick. A Durco centrifugal pump (Model H '34MDVX-80) made of type-316 stainless steel has been modified to incorporate a frozen sodium seal. The seal is cooled by forced convection. The lubricating oil for the bearings 1is circulated through a cooling radiator. This pump will deliver 45 gpm with a head of 45 ft of fluid and is expected to operate over a temperature range of 1000 to 1300°F. The modified pump is being assembled in an isothermal loop that has temperature measurement, flow measurement, and control equip- ment. Forced-Convection-Cooled Sodium- Sealed Pump. A type-316 stainless steel Durco pump similar to that described above has been modified to incorporate a copper-finned-sleeve frozen sodium seal that will be cooled by forced-air convection. This pump 1s designed to be used for supplying sodium at approximately 1500°F to a fluid-to-air radiator for the opera- tion of a Boeing turbojet engine. The pump will operate at 3000 rpm and deliver 50 gpm of sodium at 20 psig. The temperature of the sodium at the pump will be approximately 1100°F, Cénnédéfidtdr.Pumfi (M. Richardson, Reactor Experimental Engineering Divi- sion). The new, 3/4-hp, canned-rotor "pump for the NaK loop has been in- stalled and operated for about 8 hours. At present the pump motors. are of two different types - a standard wound motor and a motor with G-E Class H insulation for 500°F operation. The FOR PERIOD ENDING MARCH 10, 1952 1000°F, Bentley-Harris, insulated wire has arrived and is in the shop to be wound. It appears that the No. 22 wire will fit in the existing stator slots without enlarging the slots. The loop was started with con- siderable difficulty because of a partial plug in the suction-line valve seat. However, the plug was partly cleared and the loop checked to tem- peratures of 500 to 650°F by using the newly installed cooler that feeds cold metal back through the motors. Only a slight rise above normal operating temperatures for the motors was noted. The pump was shut down when a leak occurred at the discharge flange. VALVES In the construction of a plumbing system for the forced circulation of high-temperature fluids such as will be encountered in the ARE, 1t will be necessary at several points to 1in- corporate valves that will operate reliably in any emergency. Some of these valves may be normally open or normally closed and required to open only once, whereas others may be re- quired to open at more frequent inter- vals. It 1is an absolute requirement, however, that each valve incorporated in the system perform its designated function quickly and reliably at any desired time during the entire Jlife of the system. During this period empha- sis has been placed on the development of valves for such application. Packing-Gland Seal Test Equipment. Equipment has been designed and 1s now being constructed for extensive testing of the packing-gland type of seal described above (see "Seals and Closures™). Although initial tests have produced a valve that appears to be reliable, the optimum design of such a seal, as well as the most 29 TR nor o T TR e S i satisfactory packing materials and lubricants, should be determined. . Frozen Fluorides Valve (W. G. Cobb). It has been determined in actual tests that two sections of a high-tempera- ture forced-circulation fluoride sys- tem can be isolated from each other by means of freezing a short section of the connecting pipe. This frozen section can be melted and operation resumed at any desired time. Such a valve, however, 1s not quick-acting. Several minutes are required to either freeze or melt the fluoride. It 1is thought that a valve of this type used in series with the conventional valve described above would provide ade- quate protection against leakage past a valve seat. Such an application would be the isolation of the dump tanks from the ARE system. ' Ball Check Valves (D. R. Ward). Specially designed ball check valves (Fig. 7) have been operated to trans- fer fluoride at 1400°F. The balls were constructed of type-440 (passi- vated) stainless steel and the housing andlfittings were of type-316 stainless steel. There was some slight evidence of the balls sticking in the open position, but in all cases increased pressure or slight tapping of the valve corrected this condition. Valve Seat Test (A. P. Fraas and G. Petersen). Zirconium and molybdenum ‘were recommended and tested as valve seat materials because neither element diffuses readily at high temperatures, and zirconium does not readily alloy with other metals. Both materials " operated satisfactorily as valve seats for 175 hr in sodium at temperatures up to 1500°F. Metallurgical examina- tion following this test indicated that the zirconium was slightly at- tacked but there was no evidence of attack on the molybdenum. 30 ANP PROJECT QUARTERLY PROGRESS REPORT UNCLASSIFIED ASK DWG.15287 Rf SWAGELOK FITTING TYPE- 300-1, TYPE- 316 STAINLESS STEEL, MODIFIED BY DRILL- ING CLEAR THROUGH FOR _~ /15710.-0D TUBING -~ ~ g NOTCH TO PERMIT LIQUID FLOW UPWARD AFTER BALL HAS WSEN\\\\\\\\ "4-in. DIA BALL STAINLESS 7 STEEL {400 SERIES)~___ N5 Yg-in. IPS PIPE COUPLING, 4-in. LONG, TYPE - 316 STAINLESS STEEL END OF TUBING IS FLARED DRgZ. WITH FLARING TOOL AFTER /b TUBING 1S PASSED THR- OUGH SWAGELOK FITTING FIRST WELD "SWAGELOK FITTING TYPE- 300-1, TYPE-316 STAIN- LESS STEEL, MODIFIED BY DRILLING CLEAR THR- OUGH FOR 3/g-in.- 0D TUBING Fig. 7. Ball Check valve. HEAT EXCHANGERS Heat transfer studies of the fluo- rides and liquid metals indicate that essentially conventional types of heat exchangers will be applicable to these systems once the corrosion prob- lems with these fluids are surmounted. The NaK-to-NaK heat exchanger has now’ operated for well over 2000 hr with no apparent loss in performance. Other heat exchangers have been designed for aircraft application as well as for use 1in other laboratory tests. Aircraft Type of Radiator (J. F. Bailey and W. C. Tunnell, ANP Divi- sion). A literature survey of the present fluid-to-gas heat exchangers T T e T, ) uty indicated that some aircraft and loco- motive radiators with similar require- ments for space, weight, and efficiency have been operated with the gas passing through the tubes and the fluid flowing through the interstices between the closely stacked tubes. This has re- sulted in a design proposed for a fluoride-to-air radiator. Preliminary calculations on the basis of arbitrarily chosen dimensions indicate that 1/2-in.-0D by 0.010-in. - wall Inconel tubes flattened to a rectangular cross section and stacked in close contact with each other may result in a desirable heat exchanger configuration. Air will flow through the flattened tubes and fluorides around the tubes in the interstices. . The advantages of the design appear to be: (1) parallel (essentially) coun- “ter-flowing fluids, (2) no serious thermal expansion problems, (3) low air pressure drop across the exchanger, (4) low specific volume (in.® per megawatt of heat transferred) of the heat exchanger, (5) simplified fabri- cation because all welds are exposed for easy inspection and repair, and (6) versatile tube spacing. Sodium-to-Air Radiator (A. P. Fraas and G. D. Whitman). The testing of the high-performance, air-radiator, core element described in the last report(®’ has not yet been accomplished because of delays encountered in Nicrobrazing the parts. The contractor who is performing the Nicrobrazing operation has promised delivery by March 10. The forced-circulation sodium loop for testing the performance of this heat exchanger has been completed. The ‘S)A. P. Fraas and M. E. LaVerne, “Heat Exchanger Tests,”” op. cit., ORNL-1170, p. 47. FOR PERIOD ENDING MARCH 10, 1952 loop is heated by I?R losses in the piping. An air flow test of the core element, which was made prior to the Nicrobrazing operation, indicates that the air pressure drop is approximately 30% lower than had been estimated for the "cold" condition. Fluoride-to-Fluoride Heat Exchanger (D. Salmon). The performance of a small shell-and-tube heat exchanger operating with fluoride mixtures 1in pure counterflow in a single-fluid figure-8 system was investigated. Tubing chosen for the study was 0.10- in.-ID by 0.025-in.-wall Inconel, and the parameters btaken were flow rate, number of tubes, and length of ex- changer. The results of the investi- gation were plotted over a range of - flows from 1 to 1000 1b/hr/tube and for lengths from 1 to 10 feet. Considering the limitations of available heat sources and pumps, a heat exchanger containing 52 tubes and having an effective length of 1.5 ft was analyzed using the above data. Its performance 1s shown in Table 3 for a constant heat input to the system of 70,000 Btu/hr. For obtaining heat transfer it would be desirable to have the Reynold’s number range extend well into the turbulent region, and it is seen from the tabulated data that the turbulent region would be barely penetrated at the highest flow for the heat exchanger analyzed. The main value of such a heat ex- changer system would be for endurance testing to determine the effect of high-temperature dynamic corrosion and possible deterioration of heat transfer performance for a heat exchanger 1in which small tubes and delicate fabri- cating techniques are used. 31 T B ST ANP PROJECT QUARTERLY PROGRESS REPORT - TABLE 3 Performance of Fluoride-to-Fluoride Heat Exchanger SYSTEM HEAT TRANSFER |PRESSURE DROP | REYNOLDS NO. INLET COLD TEMP. FLOW RATE | IN EXCHANGER | IN EXCHANGER IN FOR INLET HOT TEMP. (gpm) (Btu/hr) (psi) EXCHANGER OF 1500°F (°F) - 0.52 37,500 0.027 104 970 - 1.56 30,000 0.26 312 1336 5.2 23,500 2.5 1040 1456 10.4 20,000 10.0 2080 1478 15.6 18,500 21.0 3120 1486 NaK-to-NaK Heat Exchanger (A, P. Fraas). The previously reported(®’ NaK-to-NaK heat exchanger, core ele- ment consisting of 52 tubes 1/8 in. OD has continued in operation. The total running time has reached 2230 hr, over 2000 hr of which have been with a heat exchanger inlet temperature of 1500°F - and an outlet temperature of 1000°F for the hot stream. The cold stream 1is about 50°F lower at inlet and outlet. - No apparent losses in heat transfer - performance or increase in pressure drop have occurred. The only diffi- culty experienced after the 1initial "shake-down" of the loop has been that the accumulation of oxide in the header tank became great enough after 2000 hr to interfere with the operation of the level control. The header tank and filter have been removed, cleaned out, reinstalled, and operation resumed. Heat Transfer in Circulating Fluo- ride Loops. A series of loops are being constructed for circulating (6)rbid., p. 45. 32 fluorides at various rates to obtain thermodynamic, hydraulic, and other engineering data. The first loop uses convection circulation and has been in operation for 500 hours. The loops are made of 1/2-in.-ips Inconel pipe, heated by electric tube furnace ele- ments, and cooled by natural convective air over a coiled pipe section. By means of a heat balance on the heater section, flow rates, velocities, and heat transfer coefficients are cal- culated. Data derived from the first loop, designated Model A-1, are given in Table 4. A natural circulation loop (desig- nated Model B) has been designed to maintain 600°F temperature difference between hot and cold legs. The di- ameter of the tubing between the hot and cold legs 1s reduced to achieve velocities of the order of 5 ft per second. This loop, fabricated from l1-in.,-ips Inconel pipe with a finned- tube cooling section, will be heated by electric tube furnace elements. The loop is expected to operate with a maximum hot-leg temperature of 1500 to 1600°F and a cold-leg temperature of 900 to 1000°F. o ¥ o iR ¥ » Ah FOR PERIOD ENDING MARCH 10, 1952 TABLE 4 Analysis of Heater Section (Model A-1) RUN NUMBER 1 2 3 4 Heat input (Btu/hr) 36,460 36,660 33,600 29,600 Over-all heat transfer coefficient (Btu/hr+ft?:°F) 43.0 34.0 34.8 37.5 Flow rate (lb/sec) 0.09 0.08 0.07 0.03 Velocity (ft/sec) 0.35 0.30 0.29 0.13 Inside heat transfer _ coefficient (Btu/hr-ft?:°F) 132 82 108 129 Heater outlet temperature {(°F) 1,450 1,275 1,343 1,580 Heater inlet temperature (°F) 1,172 950 1,029 9717 Flow rate from hydraulic analysis (1lb/sec) 0.03 INSTRUMENTATION extended periods. The degree of ac- Operation of a high-temperature forced-circulation system with either liquid metals or molten fluorides necessitates maintenance of level in- dication and control, accurate de- termination of flow rates, and reliable pressure measurements at several points in the system. Since the use of most commercially available equipment for performing these functions is limited to temperatures considerably lower than the minimum operating tempera- ture, an extensive developmental pro- gram has been carried out to either produce new instruments or remove the temperature limitations placed on existing instrumentation. Development has now progressed to the point where levels can be controlled in both high- temperature fluoride and sodium systems with great reliability for periods extending more than 2000 hours. Al- though equipment 1is still lacking for the calibration of flow measuring devices, means are now available for curacy eventually attainable will depend upon the procurement of ac- curate calibration equipment. Pres- sures are reliably measured at many points in fluoride and sodium forced- circulation systems with both com- mercial and locally developed instru- mentation. Flow Measurement. Electromagnetic flowmeters have proved to be very reliable when used with systems cir- culating sodium or other liquid metals having good electrical conductivity. With liquid metals it is possible to measure toas high a degree of accuracy as 1s permitted by the calibrating equipment. In the case of fluorides, which have very low electrical con- ductivity, the prospects of using an electromagnetic flowmeter are nil. A venturi section with associated pres- sure sensing instruments appears to be the most attractive device for measur- ing flow in a high-temperature fluo- ride system because of the low head 33 T T - —rve loss. The extrapolation of water- calibrated venturies to fluorides has been sufficiently accurate. Pressure Measurement. No pressure- sensitive instrument has been developed at ORNL or found commercially availa- ble that will operate at the tempera- tures encountered in the forced-circu- lation systems. The highest tempera- ture rating of any known commercial instrument is 450°F. Consequently, all pressure measuring devices have used trapped gas, which results in the pressure at the gas-liquid interface being measured instead of that of the liguid stream. However, this error in pressure measurement can be minimized by calibrating each set of instruments -and associated gas traps with accurate Bourdon tube pressure gages are ‘sufficiently reliable for rough pres- - sure measurements but are not desirable if a high degree of accuracy is to be obtained. Bourdon tube gages, which indicate_the”differential pressures directly, are commercially available. . Th? null balanéé typé of preésure gages indicate differential pressure on a single gage to the nearest hun- - dredth of a pound per square inch. Although use of this instrument 1is limited by temperature, it was connected to the hot liquid stream by gas lines and was operated successfully with the gage at room temperature. Manometers have also been used to ‘measure differential pressfire. Al- though they provide a reasonably ac- curate measurement of the differential pressure, these devices are more dif- ficult to seal than the other systems described and are therefore not recom- mended for use with a trapped gas sys- ‘tem. Temperature Measurement.- Tempera- tures up to 2000°F are measured by 34 ANP PROJECT QUARTERLY PROGRESS REPORT means of Chromel-Alumel thermocouples and are recorded on multipoint Brown temperature recorders. It has been found that the accuracy of temperature measurement depends to a great extent on the fabrication and installation ~of the thermocouples. Two methods of thermocouple attachment are being used: (1) pulse-welding each thermocouple wire to the outside surface of pipes or containers and (2) forming a beaded junction of the two thermocouple wires. In each case the thermocouple is given additional rigidity by tying it down with nichrome wire. With either of these methods a temperature error 1is introduced if the thermocouple is not insulated from ambient air or other adjacent material. Borax has been found to form an airtight coating that prevents further oxidation. Tests conducted with thermocouples placed in deep wells centered in the flowing stream and other couples directly opposite on the external surface of the pipe have revealed that there is no more than a 15°F temperature drop between the internal thermocouple and that placed on the outside. Thermocouples made by beading the wires are accurate within +1°F. The Brown temperature recorders, with careful adjustment, will record temperatures accurately within *1°F over the full range from 0 to 2000°F. They do however show a tendency to drift after extended periods of opera- tion, and cause errors of the order of 5°F. This necessitates frequent cali- bration of the instruments; therefore a program for routine inspection has been set up, which should reveal any instrumentation errors soon after their occurrence. Level Controls and Level Indicators (A. L. Southern, ANP Division). Four level indicating devices have proved ry ¥ A satisfactory for use 1in high-tempera- ture liquids. The conventional probe- type of level control operates a relay or solenoid when the level of the con- ducting fluid rises to short out the probe. Such a level control has proved most satisfactory both with liquid metals and molten fluorides, although in the case of liquid metals some difficulty was encountered because condensation on the spark plug in- sulation shorted out the plug. Another level indicator using the principle of a resonant cavity has been designed, and simulated experi- ments indicate that this instrument may give satisfactory performance. The variable inductance level in- dicator consists of a small coil wound on the outside of the tank in which the level is controlled and a tapered iron core mounted on a float 1i1nside the tank. The change in level raises and lowers the tapered core inside the coil and alters the inductance. Tests made with the iron core at room temperature demonstrate that it gives a linear response that can be directly correlated to a level inside the system. The fourth level indicator de- veloped consists of a small tube ex- tending to the bottom of the tank in which the level is to be controlled. Gas flowing through this tube bubbles through the liquid. The pressure re- quired to maintain a constant flow of gas through the ligquid is directly related to the height of the liquid level above the end of the tube and is measured by a manometer connected between the gas tube and the gas space above the free surface of the liquid. Such an instrument can be used for indicating and controlling liquid levels. FOR PERIOD ENDING MARCH 10, 1952 HEATING AND COOLING OF HIGH- TEMPERATURE SYSTEMS Heating and cooling of liquid metals, hydroxides, and mixed fluo- rides are being investigated from room temperature to 1800°F. Laboratory methods of heating, some of which are applicable to reactor preheating, have 1included a variety of electrical means, and gas heating ovens are being installed. Cooling has been accom- plished by liquid-to-liquid and liquid- to-air heat exchange. Knowledge of heat transfer properties, film coef- ficients, and the effects of turbulence is increasing. Initial attempts at preheating systems to be used in handling fused fluoride mixtures melting between 800 ‘and 1000°F resulted in an epidenic of heater failures. Heating methods used with sodium were studied and improved to avoid overheating of heater elements, and the insulation, which had been chosen for inertness in contact with sodium, was improved to reduce heat losses. The heating techniques de- scribed below, which were developed for use with fluorides, are equally applicable to hydroxides and liquid metals. External Heating Systems. Several varieties of external heater units have been tested in fluoride systems. The different heat units included an integral heater-insulation assembly that can be prefabricated in the de- sired shape, a flexible heater cable that can be wrapped around the object to be heated, and calrod heaters that are specially constructed to have greater contact with the surface to which they are attached. The limiting temperature of these heaters is de- termined by the heater elements and varies from 1250 to 1800°F, depending upon the desired lifetime. 35 LiEtolid akid oo ‘ment temperature of 1250°F. ANP PROJECT QUARTERLY PROGRESS REPORT The integral heater-insulation assembly, which can be prefabricated in a variety of sizes, consists of a conventional clamshell tube furnace element inserted into and attached to preformed insulation by means of clamps. This assembly has resulted in a substantial saving of installation time as well as in reduced heat loss to the ambient air. The heaters have operated more than 500 hr with " filament temperatures at 1800°F without becoming detached from the insulation. Heaters of this type are preferred when assembly time is at a premium. A fléxiblévheater cable that can be tightly wound on surfaces of com- .plex geometry has been used to obtain contact heating and reduction of heat-up time. The heater, consisting of nichrome wire covered with a double . layer of fiberglass tubular insulation, has been used successfully with fila- Nichrome wire commercially covered with this insulation has been procured and 1is being tested. This thin insulation is . superior to other types of insulation, ‘such as ceramic beads, since the wire can be more closely wound on the surface to be heated and the temperature drop from heater wire to heated surface is considerably lower. Dies have been constructed for flattening the tubular calrod heaters to give greater contact with the pipe to which they are attached. As modi- fied, one side of the heater 1s shaped to conform to the curvature of the pipe for nearly the full length of . the heater. 4 Care must be exercised in such shaping since it decreases the volume of the heater that contains the magnesium oxide insulation and results in heater filament relocation and distortion and subsequent hot spots in operation. 36 Induction Heating (A. L. Southern, ANP Division). High-frequency heating is advantageous and availlable as a means for rapid heating of small com- ponents. Megatherms producing 20 kw of power at 220 kc have been modified with water-cooled coi1ls to obtain temperatures over 1500°C. This method of heating is uniform when the ma- terial to which heat is applied is of symmetrical geometry. On complex surfaces, however, that section closest to the coil is heated as much as 200°C higher than the rest of the material. The power producing equipment 1is readily adaptable toheating components containable in a space up to 24 1in. in diameter and length. Resistance Heating (R. G. Affel, ANP Division). The generally appli- cable method for rapid heating of equipment appears to be direct resist- ance heating. This type of heating is superior to other types, in cases where applicable, since the heat 1is generated internally at the point where it is desired and heat conduc- tion by means of air or other gas 1is eliminated. Direct resistance heating has been used successfully for heating sections of pipe where freezing would be critical and for preheating thermal- convection loops 1n preparation for the filling operation. Sections of l-in.-ips pipe 6 ft long have been heated to 1500°F and convection loops to 2000°F, for the purpose of cleaning with hydrogen-firing. Heating rates of the order of 100°F/min may be ob- tained. The uniform heating obtained and the elimination of heater failures make this method of heat application attractive,. Insulation Testing (R. G. Affel, ANP Division). The most satisfactory insulation for use with sodium and sodium-potassium alloy has proved to T} i be the lead-mill slag-wool type be- cause of its relative inertness when in contact with these liquids at high temperatures. For fluorides, however, the chemicals themselves are rela- tively inert and the insulation 1is only required to be nonoxidizing and nonexplosive when suddenly placed in intimate contact with the high-tempera- ture liquid. However, with fluorides it was found necessary to reduce heat losses through insulation to alleviate overheating of heater elements, and Johns-Manville "Superex" insulation, rated for 1900°F service, was used since it 1is available in preformed shapes to cover standard pipe sizes. This new insulation has reduced the insulation surface temperature on 1500°F thermal-convection and forced- circulation loops from 400°F for the lead-mill slag to an average of 200°F for the Johns-Manville "Superex" in- sulation. ARE Core Preheating (D. F. Salmon, ANP Division). The effectiveness of various methods of preheating the ARE core 1s being examined analytically and by test. The most promising pre- heating methods are believed to be: 1. Electric resistance elements ap- plied directly to the outside of the pressure shell, 2. Circulation of a high-temperature gas or liquid through the coolant tubes or core 1interstices, 3. Cyclic charging and dumping of the actual reactor coolant, heating the fluid between cycles, and utilizing a portion of the tempera- ture excess above its freezing point. ' The test vehicle wofild be a fhll-éize 60-deg 7 segment of the reactor. The methods of preheating may be tried FOR PERIOD ENDING MARCH 10, 1952 singly or in combination and core tem- perature vs. time for varying quanti- ties of heat input recorded, in order to determine actual requirements and to galn operating experience. In the course of the investigation a graphical solution of the transient heating of a hollow cylinder of beryl- lium oxide with a constant flow of helium at 1500°F inside the cylinder was made. A thermal diffusivity con- stant of 0.186 ft?/hr for beryllium oxide was assumed, and an effective internal heat transfer coefficient of 25 Btu/hr ft?+°F was used. The solu- tion indicated that the time required to reach an outer beryllium oxide surface temperature of 1200°F with the above conditions is approximately 40 minutes. The low heat capacity of helium gas would obviously be the con- trolling factor for preheating time and would greatly increase the time required. TECHNOLOGY OF HIGH-TEMPERATURE LIQUIDS The successful operation of equip- ment containing liquid metals, fluo- rides, and hydroxides at high tempera- tures 1s largely dependent upon purity of the fluid sample and cleanliness of the ultimate container. Attainment of the desired purity of the fluid will require close monitoring of each fluid batch from the time it 1is prepared until 1t 1s sealed in its container and the development of special equip- ment for the preparation, storage,and transfer of the various fuel mixtures. Similar precautions must be taken to ascertain the cleanliness of fluid containers, and degreasing, pickling, hydrogen-firing, and electrolytic cleaning may be involved. Fluoride Preparation and Handling V(B. C. MacPherson, H. P. Kackenmester, .. A. Mann, J. C. White, E. Wischhusen, ANP Division). Pilot-plant-scale 37 T YT I T T e Lo - o g s e b, i s i it i i ki ANP PROJECT QUARTERLY PROGRESS REPORT preparation and handling of several eutectic fluoride mixtures, Table 5, ‘have been accomplished by using tech- "niques specified by the Materials Chemistry Division. For most of these batches, the mixed components were liquified in an 1nert atmosphere in the presence of stainless steel and Inconel chips at 1800°F to remove some of the objectionable contaminants. They were subsequently stored solid or transferred for use in the molten state under helium or argon gas pres- simply forcing the fluid through a frangible diaphragm located between the two chambers. The pretreatment consists of heating the fluorides to 1800°F for several hours in the presence of stainless steel and Inconel chips before filtering. Storage and handling require that the fluorides are not recontaminated and that they are contained in simple equipment in suitable physical condi- tion for handy use. These require- sure. ments are somewhat contradictory, and TABLE 5 - Composition of Various Fluoride Fuels and Coolants 7 FLINAK FULINAK* FUNAK FUBENA ~ * Mixture Number - 12 14 2 17 :C0m£dsiti0h'(wt %) UF, 7.8 71.4 12.6 NaF 11.7 10.8 16.1 39.5 KF _ , 59.1 54.5 12.5 LiF 29.2 26.9 BeF, " 47.9 ‘Cohpoéifion (mole %) UF, | 1.1 27.5 2.0 ~"NaF | 11.5 10.9 46.5 47.0 KF | 42.0 43.5 26.0 LiF | 46.5 44.5 BeF, | 51.0 *Made by adding UF, to Flinak The equipment for fluoride melting and treatment has been redesigned and simplified in construction and opera- tion. The mixing chamber, storage chamber, filter, and transfer lines can be enclosed within a single furnace to assure more adequate heating, and filtration can be accomplished by 38 for the present all fluorides are stored in the solid state under puri- fied inert gas when received from the melting and pretreatment equipment. Handling is accomplished by connecting the storage containers to test equip- ment and making a pressure transfer after remelting the fluorides. g T T T U SR . i ) » wi Division). H _ ferred for an inert gas blanket be- Since fluoride fuels in other than solid and liquid form may be needed ultimately, preliminary experiments, which were quite successful, have been carried out on pelleting the eutectic salts. Sampling and Analyzing Techniques (J. P. Blakely, Materials Chemistry Division). A molten fuel sampler has been designed, constructed, and used successfully to obtain a sample of molten fuel during the filling of a convection loop. The sampler con- sists of a heavy-walled, 100-cc, graphite crucible held between two metal plates with necessary transfer and gas lines attached. The sampler can be used either independently of or in conjunction with the actual filling of a loop. Another sampler has been designed that will make possible repeated “sampling of a system that is in opera- tion. When in use, molten fuel will continuously drip through the sampler and the sample container can be moved as desired to intercept the drops. Analysis of the individual fluo- rides and eutectic mixtures of NaF- KF-LiF-UF, (taken in this manner) have shown traces of oxygen, hydrochloric acid, sulphur dioxide, silicon, silicon tetrafluoride, sodium, potassium, boron trifluoride, and lead chloride. Lithium and uranium have not been ob- served.(7> . Diffusivity of Helium Through Stainless Steel (E. Wischhusen, ANP Helium is generally pre- cause o6f its low density and relative mobility. Recent tests have demon- strated that helium, under 54-psi constant pressure, may be contained . (7)Letter from C. R. Baldock to R. C. Briant, Y-B16-3, Jan. 15, 1952. FOR‘PERIOD ENDING MARCH 10, 1952 in stainless steel and/or Inconel seamless tubing of 0.030-in. wall thickness at 1500°F+ for 150 hr with- out detectable diffusion through the walls. A Westinghouse mass spectro- graph type of helium leak detector, sensitive to 1 part in 3.5 million, was employed in this test.(3) Cleaning and Inspection Technigques (L. A. Mann and D. R. Ward, ANP Divi- sion). Metal components have been receilved 1in various conditions of cleanliness and soundness; therefore inspection and cleaning specifications for components and assemblies have been established to eliminate struc- tural defects such as crevices and pits and to remove oxides and other contaminants from metal surfaces. The cleaning techniques include degreas- ing, pickling, hydrogen-firing, and possibly electrolytic cleaning. In addition to these techniques, the use of additional flushing solutions 1s required when cleaning previously used equipment. The cleanliness of internal surfaces may be determined with a boroscope. The process used for degreasing metal parts depends upon the type of grease encountered, the completeness of degreasing required, and the tolerable amount of film or deposit that can be left onor in the degreased part. Tet- trachloroethylene appears to be the most promising of all degreasers for ~these purposes and trichloroethylene the next. Following the degreasing operation, pickling with HNO, has been established as the most satisfactory general clean- ing process fornickel, Inconel, Monel, and stainless steel equipment. This type of pickling leaves the metal (S)F. Wischhusen, Containment of Helium in Stainless Steel and Inconel at the 1500 F+ Range, ANP-72, Oct. 16, 1951. 39 Ty et O ANP PROJECT QUARTERLY PROGRESS REPORT covered with what is thought to be essentially a monomolecular layer of protective (dgainst many corrodents) oxide; otherwise, 1t 1s as clean as possible. Oxides may then be removed from the surface of these metals by contacting the metal (oxide) surface with very dry oxygen-free hydrogen gas at temperatures of from about 1500°F for nickel to above 1800°F for chromium. The hydrogen 1s dried to a -40°F dew point by first passing it over a palladium (or other) catalyst at room temperature and then passing it through activated alumina. ‘Electrolytic cleaning has also been examined as a possible technique, although inherently it has the diffi- " culty of requiring properly shaped and placed internal electrodes. Of the various electrolytic procedures ex- amined, only the anodic treatment in phosphoric acid solution evidenced the high degree of cleaning desired. " In used equipment most of the fluo- -~ rides may be drained out while molten. However, a coating will remain that 40 is dangerous 1f radiocactive or toxic fluorides are present. It has been determined that some molten salts and hydroxides will remove the fluorides by flushing at temperatures above the melting point of the fluorides, but such a treatment is difficult and could easily mask the surface condi- tion by additional corrosion. Of the several noncorrosive cleaning solu- tions tested, tap water and water con- taining 10 to 50% H,0, gave the best results, the latter being perhaps slightly more effective. However, the results obtained amounted only to softening the fluorides so that they could be easily removed mechanically by brushing or with high-velocity water. : Prior to assembly all metal parts are visually inspected for surface defects and cleanliness. Internal surfaces are examined by use of a boroscope that will reveal major in- ternal flaws, scale, slag, etc., but its effectiveness is limited by the patience and thoroughness of each individual inspector. ’ o o i »y &) [ 3 a4 . FOR PERIOD ENDING MARCH 10, 1952 4. REACTOR PHYSICS W. K. Ergen, ANP Division . The statics and uranium investment of the 350-megawatt circulating-fuel aircraft reactor were described in the- last report¢!? and at present are not a major concern. However, the kinetic behavior of a circulating-fuel reactor introduces several new considerations into a somewhat obscure kinetic picture. Malfunctions that brought an increase as large as 3% of uranium into the reacting zone would cause an average temperature rise of ~100°C if a thermal expansion coefficient of 3 X 10°* per °C is assumed for the fuel. (An example of this type of mal function 1s the thrombosis effect, in which a precipitation of uranium accumulates outside the active lattice and is suddenly transported into the reactor.) In addition to the thermal expansion of the fuel, direct nuclear effects could yield a temperature coefficient of reactivity whose sign is not now known. The loss of delayed neutrons from the active lattice, caused by the circulation of the fuel, has been previously viewed with con- cern because of the damping effect associated with the presence of these neutrons in static-fuel reactors. It now appears plausible, although not yet proved, that the circulation of ‘the fuel itself acts as a damping mechanism, and possibly as a powerful one. The kinetic difficulties are not very serious for the ARE. The flow velocities of the fuel are so low that the control rods can keep up with the possible entrance of excess fuel into the reacting zone or with the exit of poison from this zone. (I)N. M. Smith, Jr., “Reactor Physics,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 10, ORNL-1170, p. 13. Consequently, the negative temperature coefficient of reactivity will not be called upon to compensate for large reactivity changes. The residence time of the fuel in the reactor 1is comparable with the longest delayed- neutron periods; therefore a con- siderable fraction of the delayed neutrons 1s given off in the reacting volume and is available for the damping of oscillations. Since 1t can reason- ably be expected that the uranium will be returned to the U?3% stockpile, the uranium investment 1is, within limits, a minor problem, The work on the ARE was mainly concerned with specific design problems and culminated in a fairly detailed study of the specific design selected for the feasibility study. The critical mass of this design was 22.3 1b, which led to a total uranium i1nvestment in the system of 74 pounds. CIRCULATING-FUEL AIRCRAFT REACTOR The kinetics of the circulating- fuel aircraft reactors are currently under intense investigation. The problems are the thrombosis effect, direct nuclear effects, and fast oscillations. The considerable amount of fuel outside the reactor could cause mal functions to occur and bring an excesslive amount of fuel into (or out of) the reacting zone.. As presently conceived, the main mechanism that compensates for these reactivity changes before any control rods can act 1s the thermal expansion of the liquid fuel. For thermal volume expansion coefficients of 3 x 10°* per °C, an increase by 3% of the fissionable material in the reactor 41 ANP PROJECT QUARTERLY PROGRESS REPORT 3 would call for temperature rise of 100°C. Actually, the temperature rise would be somewhat larger because at short times after a sudden increase of fissionable material in the reactor a temperature overswing occurs, and at longer times the effective fuel expansion coefficient i1s decreased by the expansion of the fuel tubes. Furthermore, the temperature rise of 100°C refersto the average temperature, ~and local temperatures may increase somewhat more. All this limits the allowable change in fuel concentration "in the reactor to about 3% - a very stringent requirement. In addition to the thermal ex- pansion of the ligquid fuel, direct nuclear effects, such as adaptation of the neutron temperature to the fuel temperature or Doppler broadening of resonance lines, could yield a . temperaturé coefficient of reactivity. ~Should it develop that such a temper- ~ature coefficient is positive and of ' the same order of magnitude as the negative temperature coefficient resulting from fuel expansion, then the circulating-fuel ANP reactor would be almost impractical. However, it is hoped that the direct nuclear effect can be made to yield a negative temperature coefficient of reactivity once these effects are well under- stood. A negative temperature coef- ficient would aid the fuel expansion effect. A strenuous effort is being made to investigate the temperature coefficient of reactivity resulting from direct nuclear effects. This is a very difficult problem and in its full extent is somewhat novel. No reportable results have been obtained thus far, In conventional reactors, any oscillations of periods that were short compared with control rod action times would be very effectively damped out by delayed neutrons. The circu- lating-fuel reactor loses a large 42 fraction of 1ts delayed neutrons in the volume exterior to the reacting zone, and at the beginning of this quarter considerable concern existed as to the possibility of undamped or increasing oscillations. This concern has been greatly alleviated by the reasoning given below, in which 1t 1is shown that 1t is plausible that the circulation of the fuel 1tself will act as a damping mechanism, and possibly as a powerful one, 'The uranium investment of circu- lating-fuel reactors is at present not a major concern, However, an in- vestigation of the use of low-assay uranium (say 10% U%2%) is planned, but no work has been completed thus far. ' O0scillations. Work in reactor "kinetics was concentrated on the effect of the circulation of the fuel. It appears likely that the circulation of the fuel itself acts as a damping mechanism. It should be emphasized that the following réasoning does not constitute a rigid proof and that there is not yet any quantitative measure of the damping effect of the fuel circulation. Furthermore, antidamping as a result of coupling between mechanical and nuclear oscil- lations is still a possibility. The following simplifying assumptions were made for this analysis of the circulating-fuel reactor kinetics: 1. All particles of the fuel spend the same time, £, in the reactor. 2. The flux and power distri- butions are constant over the reactor so that the temperature rise of a fuel particle during a time interval dt is proportional to the total reactor power P(t) and the proportionality constant € 1s independent of spatial coordinates, and so that a temperature changeof an element of fuel influences the reactivity to an extent that is Iy FL S $ " 4 independent of the position of the element. 3. Delayed neutrons are neglected. 4. A change AT of the average fuel temperature, 1, causes an instantaneous change ~oAT in the reactivity. 5. The reactor inlet temperature i1s constant. The equations used are . a - P =-—PpPT _ (1) T . € T = €p “ P(o)do . (2) t-6 In addition to the symbols already specified, 7 is the prompt generation time, P and T are functions of the time t; and T is the deviation of the average fuel temperature from the value that gives zero excess re- activity. With these simplifying assumptions, Egq. 1 1s equivalent to the first of Egs. 1 in an HRP quarterly report.¢?’ Equation 2 can be understood by considering that T changes during a time element dt for two reasons: 1t increases by €Pdt as a result of the power input and it decreases because hot fuel is expelled from the reactor. (This explanation would be very simple 1f T were defined as the excess of the temperature over the constant reactor inlet temperature. Such a ‘definition of T would differ from the one used here only by an additive constant that disappears 1f the time (2)T. A. Welton, “Damping Produced by Delayed Neutrons,”' Homogeneous Reactor Project Quarterly Progress Report for Period Ending August 15, 1951, ORNL-1121, p. 99. . traction. FOR PERIOD ENDING MARCH 10, 1952 derivative T is formed.) The decrease is proportional to the fuel outlet temperature, which i1s € times the integral of the power over the time the expelled fuel spentin the reactor. The factor 1/6 is used because the temperature of the expelled fuel influences the average temperature T only to the extent of the ratio of the volume of the expelled fuel to the total volume of fuel in the re- actor; for the fuel expelled during dt this ratio is dt/6. By dividing Eq. 1 by P, differ- entiating twice with respect to ¢, substituting T from Eq. 2, and differ- entiating again with respect .to t, the following equation 1s obtained: ;3 5 log P+ =P =~ [P(t) ~P(: ~6)]=0. (3) at This third-order, non-linear, differential-difference equation has, evidently, a simple solution if P(t) happens to be periodic with the period &/n, where n is a whole number. This special case occurs, of course, only if € has a specific relation to the parameters determining the period of osciliation (a€/7, and the amplitude and average value of P). 1Ia the special case the solution turns out to be 1dentical with the well-known solution for constant power ex- (3) This is also physically easy to visualize: 1f each particle sees, on 1ts transit through the reactor, a whole number of cycles, then each particle attains the same temperature by the time 1t reaches the reactor outlet. With constant outlet temperature, the power ex- ~ traction 1s constant. (3)“Effect of Temperature Coefficient,” Homogeneous Reactor Experiment Report for the Quarter Ending February 28, 1950, ORNL-630, p. 23. 43 T T e ANP PROJECT QUARTERLY PROGRESS REPORT If the period is not &/n, the case of small oscillations can be con- sidered, that 1s, the linearized form of Eq. 3. The Nyquist theorem, or its equivalent, then shows that there are no antidamped oscillations. In the nonlinear case, the following reasoning may be used to make it plausible that all oscillations are damped: | Divide Eq. 1 by P, and mulfiiply by Eq. 2: ST - P(t)ft P(o)d " = ¢ 7 o, Los 8 o) do . t - Integrate from « to a *+ p, using partial integration on the last term on the right. a and ¢ + p are, so far, arbitrary times. times, this relation cannot be ful- filled for all particles, and the right side is really negative 1f the oscillations are undamped. Apparently this discrepancy between the vanishing left side and the nonvanishing, negative right side disappears 1f the oscilllations are damped, and hence P(o) 1s depressed as compared with Ploc - 0). Slow Kinetic Effects. (%’ Slow kinetic effects are those that occur in times of 1 sec or longer. These effects are not a major worry because of the possibility of compensating for them by control rods and servo mechanisms. Hence, the slow effects have not been investigated systemati - cally for the aircraft reactor. However, a study was made of the net positive reactivity coefficient re- sulting from the reduction 1n the Xxenon absorption cross section with t=atp t atp a € € - T? + eP - — log P(t)f P(o)do = "Ef log P(o) [P(o)-P(c=6)]do. 7 t-0 a t=aq " Now assume that fiheréis an undamped increased moderator temperature. The periodic oscillation, and identify p with the period. Then the left side of the equation vanishes and the right side is, by a known theorem(*) €0. Presumably the right side 1is equal to zero only if the transit time € is in a certain relation to the other parameters of the oscillation. Since, 1in reality, different particles " of the fuel have»different transit (4)This follows readily from theorem 378 of Inequalities by G. H. Hardy, J. E. Littlewood, and G. Polya, University Press, Cambridge, Eng., 1934, p. 278. In order to apply theorem 378 directly, log P has to be nonnegative, which always can be accomplished by choosing the units of power sufficiently small. The units of power are normally arbitrary. 44 study shows that for a 200-megawatt reactor, an introduction of an excess reactivity of 0.0091 causes a temper- ature increase of only 39°F in the first second. (For further specifi- cations of the assumptions, see ref. 6.) Critical Mass. Instead of the NaF-UF, coolant used in the corre- sponding calculations of the last quarterly report,(®) a NaF -BeF, -UF, (S)C. B. Mills, A Flux Transient Due to a Positive Reactivity Coefficient, Y-F10-63, Jan., 14, 1952. (G)C. B. Mills, **Circulating-Fuel Reactors,” ep. cit,, ORNL-1170, p. 14. — e e I e T Lo o L L 1952 FOR PERIOD ENDING MARCH 10, DWG. 14425 REFLECTOR PRESSURE SHELL 'Schematic Diagram of Circulating-Fuel Aircraft Reactor. Fig. 8. 45 i ] 4 g--‘ i 'O ANP PROJECT QUARTERLY PROGRESS REPORT 20 18 Fig. 9. Leakage Spectrum Through the craft Reactor. coolant was used.(’) Furthermore, the Inconel structure was assumed to occupy 2% of the reactor volume (instead of 1.5% as in the last re- port). The net effect of these two changes was a small decrease in critical mass. Using beryllium metal instead of beryllium oxide as a ‘moderator would reduce the critical mass by only 6.5%. The fuel self- shielding corresponds to changes Ak of 2.2 and 5.8% for 1-in.-ID and 2-in.-ID tubing, respectively. (T)C. B. Mills, The Circulating Fuel ANP Reactor with NaF-BeFQ—UFé Fuel-Coolant, Y-F10-83, Feb. 5, 1952. 46 (- oo - £ i ..?.- SFERET DWG. 14426 | ] * COMPOSITION BY VOLUME FRACTION — CORE o REFLECTOR FUEL-COOLANT ( NoF-BeF,-UF, ) 0.3407 - STRUCTURE (INCONEL) 0.0197 STRUCTURE 0.05 MODERATOR (BeO) 0.6076 MODERATOR 095 _| INERT SALT 0.0320 >.. £ 2 0.005 g O EE S 0.004 ‘ oW 0.04350 '—HJ-~L_1_ THERMAL S 0003 ESCAPE L D ; AT ¢ = 0002 y=18.6 T o Q W > 0.00 lej;____ 2 x ~7 < a0 — 6 14 12 o 8 6 4 2 0 LETHARGY (v) | Reflector of the Circulating-Fuel Air- + Neutron Leakage Spectra. As has " been described,¢®’ leakage spectra for 3 }.....-.-. an aircraft reactor were computed. The reactor is shown schematically in Fig. 8. It has two-pass fuel flow, and the core, which is aright cylinder with conical ends, 1s surrounded on all but the fuel entrance sides by a thick beryllium oxide reflector. Leakages through the reflected sides and end, through both the thin, un- reflected annulus around the fuel- coolant pipe, and through the pipe itself (cut off 6 in. from the core surface), are shown in Figs. 9 and 10. (S)C. B. Mills, Circulating Fuel-Coolant of the ARE of February 12, 1952, Y-F10-992 (to be issued). FOR PERIOD ENDING MARCH 10, 1952 SECRET DWG. 14427 ] l l | | 1 I I l 5.040 COMPOSITION BY VOLUME FRACTION - ‘ CORE REFLECTOR . 0035 FUEL-COOLANT (NaF-BeF,-UF, ) 0.3407 FUEL 0.84 _ § o "~ STRUCTURE (INCONEL) 0.0197 STRUCTURE 0.16 || : 9 MODERATOR (BeO) 0.6076 B £ 0030 = inert saT 0.0320 T - W@ £ 0.025 — z 4 0.00195 < THERMAL 5 £ 0020 ESCAPE o o= AT lé(’-! S 0 015 v =i86 ———h} ] g« | |—J Y a 0.010 : ] I |__r"J . 0.005 i 7 . i : dl (a) 20 18 16 14 12 10 8 6 4 2 0 LETHARGY (v) .040 1 i T 1 1 T T T 0 COMPOSITION BY VOLUME FRACTION , 0.035 CORE REFLECTOR § @ FUEL-COOLANT (NaF-BeF,-UF,) 0.3407 FUEL .00 I 8 STRUCTURE ( INCONEL) 0.0197 E & 0030 | b - MODERATOR ( BeO) 0.6076 - INERT SALT 0.0320 = 0.025 _ Zz J 0.00293 - g THERMAL M4 e > 0.020 ESCAPE ] - T 1 Jlles - l'l 8 = 0.015 ' ; — _ < Y @ 1 .JJ 5 & | | J_[ - 0.010 - | r__l_r'— ] = | | 0.005 L:"Hrr ] | L _ (b) 20 18 16 14 12 10 8 6 4 2 0O - LETHARGY (¢) Fig. 10. Leakage Spe'ctrum Around the Fuel Pipes of the Circulating-Fuel Air- u craft Reactor. (a) Through the annular area around the fuel pipes. (b) Through the fuel pipes. ' -y 47 ANP PROJECT QUARTERLY PROGRESS REPORT ALKALI HYDROXIDE MODERATED REACTOR(?) In continuation of the work on circulating-moderator reactors re- ported in the last quarterly report, (%) the critical masses plotted in Fig. 11 as functions of core diameter were computed by the bare-reactor multigroup method. A thick hydroxide reflector was assumed, the cross section of uranium and the hydroxides only were included (thus neglecting the poisoning by structures, etc.), and the following density values (in g/cc) were used: [LiOH = 1.21, NaCH = 1.61, KOH = 1.59, hbOH = 2.9, Sr(OH), = 3.4. All curves would show minimums 1f extended to lower reactor volumes. The large critical masses for potassium hy- droxide result from the large ab- sorption and small scattering cross section of potassium, and these cross sections also explain the moving of the minimum to large core volumes, Except with potassium hydroxide, the hydrogen moderation results in small critical masses. SURVEY CALCULATIONS OF TiHE 'CIRCULATING-FUEL ARE C.B. Mills, ANP Division At the beginning of this quarter, a survey calculation was carried out(!?? on the circulating-fuel ARE. The main variable was the vbyume fraction occupied by the fuel-coolant. Although this fraction varied from 6.85 to 22.4%, the total uranium investment (assuming 6.8 ft® of fuel external to the reactor) varies only from 41 to 43 1b and goes through a flat minimum of 40 1b around a value of 15% for (Q)C. B. Mills, Critical Masses of Sone Alkeli-Hydroxide Moderated Reactors, Y-Fl10-8¢, Jan. 21, 1952, (10)c. B. Mills, The Circulating Fuel ARE Core Series of December 13, 1951, Y-F10-76. 48 this fraction. The per cent thermal fission decreases from 83 to 67, as the fuel volume fraction increases over the above range. The most recent version of the ARE has a small (7.60) volume per cent of fuel-coolant circu- lating in a series-parallel piping arrangement. Critical mass, reactivity coefficients, and other numbers of design interest for the circulating- fuel ABE are presented. A specificARE design, now obsolete, was investigated('!) prior to the work on the ARE design presented here. ARE Core Design. The cross section of the circulating-fuel ARE core normal to the axis consists of 82 hexagons of beryllium oxide with central holes for 78 fuel tubes and 4 control rods. The fuel tubes are 1.235 OD and have 0.060-in. wall thickness. The core length 1s 35.25 in., and the ends of the core are bare. The reflector i1is composed of hexagonal blocks of beryllium oxide with 57 central cooling holes con- structed to build out the core plus reflector to a right cylinder with a diameter of 47.750 inches. A 2-in.- thick Inconel pressure shell with 48 in. ID, 52 in. OD, and 48.50 in. in length encloses the core and re- reflector assemblies. The fuel-coolant volume 1in the heat exchanger and the plumbing outside the reactor is 4.4 ft3 (does not include 0.2 ft® in tube bends at the reactor ends). The total core volume is 20.42 ft®, so there is 1.55 ft® of fuel in the core. The ratio of fuel in the core to total fuel is 0.26. Volume fractions of the core and reflector, as used in the design calculations, are given in Table 6. (11)c, B, Mills, The ARE with Circulating Fuel- Coolant, Y-F10-82, Jan. 11, 1952. 'FOR PERIOD ENDING MARCH 10, 1952 \/ SECRET DWG. 14428 CRITICAL MASS DETERMINED BY BARE REACTOR METHOD, USING A REFLECTOR SAVINGS VALUE OF L i '_ (THE DIFFUSION LENGTH) = | = 40 ‘ K OH SrOH 100 n RbOH wn <{ § . 4 30 75 g | NaOH. = - o 5 20 50 > & > LiOH 10 ' : , 25 - -0 { 2 | 3 4 . ' CORE DIAMETER (ft) fiig. 'li; 'C'r'i'tiéa'll‘mlass vs. Core Diameter for Hydroxide Reactors with Thick Reflectors of the Same Comp_o"s_ition._ , ‘ u [ S LS . 49 & B R R i ANP PROJECT QUARTERLY PROGRESS REPORT TABLE 6 Volume Fractions of the Circulating-Fuel ARE Core and Reflector* CORE REFLECTOR (vol %) (vol %) Fuel-coolant 7.6 Beryllium oxide 80.7 92.23 Inconel 1.96 0.69 Inert salt 8.26 6.34 Void 1. 40 0. 64 *Communicatedby R. W. Schroeder, February 8, 1952, Crltlcal Mass and Total Uranium Investment. The critical mass for the circulating-fuel ARE is obtained from bare(!?) and reflected(!®) reactor calculations of equivalent spherlcal core with the volume fractions as given in the preceding section. The critical mass is arrived at by a series of approximations in which the keff corresponding to an arbitrary uranium mass 1s calculated until the mass corresponding to the required keff 1s found. The %, from the bare and reflected calculations are weighted when determining the actual eff of the ARE according to the 'percentage surface area of the core, which 1s assumed to be either bare or reflected. The résulting'critical.mass of the ARE reactor described in the preceding ‘section 1is 22.3 1lb and will provide a (IZ)M. J. Nielsen, Bare Pile Adjoint Solution, Y-F10-18, Oct. 27, 1950. (13)D. K. Holmes, The Multigroup Method as Used by the ANP Physics Group, ANP-58, Feb, 15, 1951. 50 "to be bare. maximum k_.. of 1.034, which is dis- tributed among the various reactivity effects as follows: criticality 1.000 fiéSion-product override 0.005 excess for experiment 0.024 excess for delayed neutron loss 0.0054 The uranium 1nvestment 1n the circulating-fuel ARE is summarized in Table 7. TABLE 7 Uranium Requirements of the ARE Uranium reacting voiume‘ | 22,3 1b Tbtal.uranium inventory 85 lb'n> Assumed éfror rafiée ;riO% to 20% 94 to 68 1b A “ best’’ guess allowing.-4.5% for critical experiment correlation 74 1b There will be a small reduction in the uranium requirement as the result of reflection and fissioning in the end reflector region, which was assumed The reflected reactor has 70.73% thermal fissions and a leakage- to-absorption ratio of about 1 to 3. _ Reactivity Coefficients. The values of the reactivity coefficients, summarized in Table 8, were obtained by bare reactor calculation meth- ods.(12:13) These methods appear justified, since the corresponding bare reactor calculation for critical mass gave keff 1.054 for 15 1b of uranium, which 1s close to the value of keff = 1.0398 for the reflected reactor. However, the reflected reactor calculations for Ak/k per °F gave -0.64 x 10-%, so -that the actual reactor (whose surface 1s 13.92/41.39 bare and 27.47/41.39 reflected) would have a Ak/k per °F of -1.95 x 10°° H For the change Ak/k corresponding to a temperature change from 68 to 1283°F, the value calculated for a reflected reactor is very nearly equal to the above value of -0.0318 calculated for a bare reactor. "' FOR PERIOD ENDING MARCH 10, 1952 is that of fission fragments in the fuel-coolant. Figure 12 presents the results of the three separated power densities. The power density 1in the fuel-coolant is shown both as fis- sions/cc/sec and normalized to an TABLE 8 Reactivity Coefficients RANGE OF VARIABLE _ WITH CHANGE OF OF COLUMN 1 SYMBOL VALUE o A'k/k -7 Thermal base {reactor temperature) 1283 to 1672°F o -5.76 x 10 68 to 1283°F Dk /k -0.03180* | Lk/k | Uranium mass 11.75 to 15 1b — 0.404 (QM/m)U D/ k Coolant density 90 to 100% of ——— 0.0153 quoted density (Ap/p)coolant Dk/k Moderator density 95 to 100% of —_ 1 0.305 quoted density (AW/p)moderator _ . Ak/k Density of structure (Inconel) 100 to 140% of —_— | «0.174 quoted density (D;o/p)structure ) D/ k Core radius 100 to 101% of . _— 0.438 : o gquoted radius (AR/R)COre *Total change over range of column 2. Power Distribution. The power distribution in the ARE fuel-coolant has been evaluated by separating the total fission energy into three parts: fission fragment energy absorbed in the fuel-coolant; fission-neutron energy absorbed in the moderator; and gamma-ray energy from direct fission, fission products, and (n,?%¥) ab- sorptions. The large energy density average of 1 fission/cc/sec in the reactor core. lhe factor converting this to watts/cc from fission fragments in the fuel-coolant solution is given " on the graph, and to this must be added the gamma-ray heating, which is given on the graph directly 1n watts per cubic centimeter. Power density in the moderator as a result of heating by gamma rays 1s also given . e . Fo 1‘%} AT ™y Laadfp LoD 51 i | 4 | i s B cn et b oo S Ao M - bbb s 5 C POWER DENSITY IN THE CORE {NORMALIZED TO 1 IN THE RADIAL DIRECTION) on the graph in watts per cubic centimeter. The assumption 1is made ‘that gamma-ray heating is the same for moderator and coolant. A first approximation to the total power density in watts/cc in the moderator is obtained by adding gamma-ray heating to neutron heating. A refined calcu- lation will be made when the design is fixed. : ANP PROJECT QUARTERLY PROGRESS REPORT Neutron Flux and Leakage Spectra. - Neutron leakage from the surface of the reactor is shown in Figs. 13 and 14. Leakage {from the reflector surface in neutrons/cm?/sec and leakage from the ends are also presented in Figs. 13 and 14, and the relative importance of open ends on neutron flux out of the reactor core may be noted. . SFCREF DWG. 14429 — REACTOR 6.15% 1.00% CYLINDRICAL , BARE -END REACTOR, 35.25 in. IN LENGTH, 35.75 in. IN DIAMETER CORE VOLUME, 20.4 ft°, 7.6% FUEL - FISSIONS Zcc/sec = 1.61 x10'' (AVERAGE) POWER DISTRIBUTION 90.0% OF POWER IN FUEL, FROM FISSION FRAGMENTS 2.5% OF POWER IN MODERATOR, FROM NEUTRONS OF POWER IN CORE, FROM GAMMA-RAY HEATING OF POWER IN REFLECTOR, FROM GAMMA-RAY HEATING 035% OF POWER LEAKING FROM ENDS AS GAMMA RAYS THESE ARE APPROXIMATE FIGURES TO BE NORMALIZED TO 3 MEGAWATTS NOTE: 1.6 FOR GAMMA AND NEUTRON HEATING USE SCALE DIRECTLY (watts/ec). —— \ FOR FISSION FRAGMENTS MULTIPLY SCALE BY 92 FOR watts/cc IN 1.4 1.2 TRADIAL POWER 1.0 —=——=—GAMMA-RAY HEATING ™ T (APPROX. wofls/cc)\ 0.8 ~— AN : S —-\-szAL POWER 0.6 ™~ ~ \\\ +=——CORE SURFACE 0.4 - ON HEATING \\ ' t ! { 1 o NEUTR A e -(fi__ (APPROX . watts /cc) , 02 e T = 1 o AT T vREACTOR END i l B fi\ _ REFLECTOR SURFACE 0 10 20 30 40 50 60 DISTANCE FROM CENTER OF REACTOR (cm) Fig. 12. Power Distribution in the Core of the Circulating-Fuel ARE. 52 ' | G s FOR PERIOD ENDING MARCH 10, 1952 SEERET DWG.14430 0.009 0.008 0.007 0.006 - S 0.005 - 0.004 0.003 LEAKAGE IN NEUTRONS /cm?2/sec FOR 1 FISSION /cc/sec IN CORE 0.002 . 0.0431 M 0.00t THERMAL i LEAKAGE 'Ll- v=186 8 16 14 12 10 8 6 4 2 0O LETHARGY (¢) Fig. 13. Leakage Spectrum from the Reflector of the Circulating-Fuel ARE. ~SEERT DWG. 14434 0.4 - §§ 0.12 O ggg OJO. 2§ & 0.08 E 8 o~ 0.1046 %g 0.06 THERMAL Q LEAKAGE = =% u=186 ! - WT 0.04 ' = gt | | L < & | Wi 0.02 L. |_ 18 16 14 12 10 8 6 4 2 0 LETHARGY (v) Fig. 14. Leakage Spectrum from the Open Ends of the Circulating-Fuel ARE. { ’"”g . E ;; .5, ! LTy 53 ULy trons/sec/cm?. kil i i, o b . s, o it ANP PROJECT QUARTERLY PROGRESS REPORT Neutron flux normalized to 1 fission/cc/sec was computed to be 364 for the monthermal neutrons and 80 for the thermal neutrons in this reactor. At full power, 3 megawatts, the total integrated flux in the center of the reactor is 16.7 x 10'3 neutrons/sec/cm?. The thermal flux is 3.0 x 10'% neutrons/sec/cm?, and the fast flux 1is 13.7 x 10!'% neu- at three points in the reactor are given in Fig. 15. The important difference 1s that of amplitude. Note that high-energy neutron flux is relatively somewhat higher toward the reactor center, but the difference is small. Figure 16 shows the corre- .sponding plots of flux vs. radius for four energies. The importance of moderation by the reflector 1s quite apparent. TP DWS.14432 THERMAL VALUES AT 186 @ AT CORE CENTER AT '/, RADIUS I8 AT CORE BOUNDARY NEUTRON FLUX (NEUTRON /7 cm2 / sec . FOR AVERAGE OF 1 FISSION / cc / sec IN CORE AT CORE CENTER AT %, CORE RADIU AT CORE 0 2 4 6 8 10 2 14 ‘e . 18 20 LETHARGY (¢) _Fig; 15.. Transverse Neutron Flux Spectrum for Three Sections Through the Core of the Circulating-Fuel ARE. 54 ool U Neutron flux spectra ~reduces k STATICS OF ARE CONTROLS Static calculations of the various mechanisms for the regulation, safety, and shim control of the circulating- fuel ARE are given in terms of changes in reactivity. The one boron carbide regul ator rod, fully inserted, effects a net change 1in reéctivity_of 0.0075. The three safety rods each effect a net change in reactivity of 0.053. Details of the operation of these control mechanisms are given in sec. 2. Shim Control Regquirements, Esti- mated reactivity changes, Ak/k, f from room temperature to 1000°F andfto an assumed controlled reactor temper- ature of 1283°F are given in Table 9. The effect on reactivity of xenon at full power 1s to reduce the kegf Y 0of the clean reactor calculation Maximum transient xenon further by Ak = 0.0082. The temperature coefficient as a result of the xenon at full power is Ak = 0.0031. Ak/k ; — =+ 6.4 x 10-7 per °F AT | Xe in the vicinity of 1400°F. Regulator Rod. In the poison control system, there is one axial boron carbide regulator rod lying along the longitudinal axis of the cylindrical reactor. The permanent reactivity effect of the regulator structural material 1s included 1in the core volume fractions. Detailed calculations have been made for the sodium-cooled ARE reactor for this reactor control system design. The only important reactor characteristic that affects the control rod effective- ness 1s the neutron spectrum, and this spectrum is quite similar for ek o S a o ha Y NEUTRON FLUX PER UNIT LETHARGY FOR 1 FISSION /cc/sec AVERAGE FOR PERIOD ENDING MARCH 10, 1952 TSHECRET— DWG. 14433 60 55 50 | ~THERMAL NEUTRONS . \“ 45 I~ | \\ . 40 ‘ ~. \\ 35 ™~ 30 25 20 5 /l mev INEUTRONS e 500 ev NEUTRONS = 10 T E—— | e 04evNflflmfi§:1~4=-~~-l__ 5 0 SPACE POINT »n (r=2062n1n cm) Fig. 16. Radial Neutron?Flux Pistributionin the Core of the Circulating-Fuel ARE. the sodium-cooled reactor and the present ARE Hence, the control rod effect on react1v1ty 1s as sumed to be the ‘same. The net change in react1v1ty when the regulator rod is fully 1nserted is to be made 0 0075 , Safety Rods.4 Three safety rods are equally spaced on a 15-in. circle around the reactor axis. The spacing is sufficiently large to reduce the shadowing effect of each rod on the others to a relatively small value, and this effect is not considered. Ty § ¥ Wl These rods are 2-in. diameter cylinders " uranium requirement. of boron carbide with wall thickness of 0.335 in., and a beryllium oxide rod is attached to the end of each rod. The normal position of the safety rods is out, so that beryllium oxide will be added to core material with a resulting decrease in the | The net change in k, for each rod is thus the sum of two reactivity contributions: (1) from the removal of beryllium oxide moderating material and (2) from the insertion of a 2-in, -diameter boron carbide neutron-absorbing rod. Each boron carbide rod is worth -5.3% 1in S : ey x g, e o N Y TABLE 9 Shim Control Requirements Dk/k ;¢ FOR CHANGE | FROM 100 | FROM 1000| OF 1°F AT EFFECT ASSUMPTION EQUATION FOR (Lk/k ¢ () /AT T0 1000°F | TO 1283°F|OPER. TEMP. Expansion of liquid fuel |Volume expansion coefficients -0.0554 -0.0174 -6.16 x 107° o -4 7 per F; fuel, 1.67 x 10 AR/ & Ne/k (melting ignored), Inconel, 1.47 x 1074 0.20 x 10°%; difference, Bn/m)y B/ P gorant | 1.47 x 1074 Dimensional expansion Radial expansion determined Ne/ & 0.0024 0.00076 2.67 x 10°° with constant material by Inconel; axial, by BeO. 6.1 x 107° ————— (AR/B)core Linear expansion coefficients -0.0083 -0.0026 |-0.92 x 1075 per °F; Inconel, 6.7 x 10'6, BeO, 4.9 x 107%, weighted Dk k average expansion coefficient:| 18.3 x 10°° linear 6.1 x 10'6, volume (Ao/p)moderator 18.3 x 10°% per °F Change of density of Inconel volume expansion coef- Ne/k +0.0031 +.00098 |[+3.5 x 10°° Inconel in core ficient per °F, 20 x 10~ 20 x 10”6 (Ao/p)structure Change of density of Same as metal-cooled ARE -0.0054 -0.0017 -5.7 x 1078 reflector BeQ ' Change of cross sections -0.0271 -0.0047 -2 x 107° (except Xe) with reactor D/ k temperature °F Change of Xe cross section| 0.26 of decay products in the 4.6 x 1079 with reactor temperature | core, remainder in external Total Me/k [-0.0907 | -0.0247 |-7.2 x 1075 LY0d3d SSAYI0Ud ATHALYVNO 1DoAfoud dNV _‘ LR "Ak/k, and each beryllium oxide rod in the same position is worth +0.16%. The Inconel around the boron carbide rods is not an effective poison when the rods are inserted and results in a decrease in poison rod effect of 0.21%. The net effect of insertion of the three safety rods is thus 15.9% 1in k, This value is larger than that quoted for the metal - cooled reactor because there 1is no poison NaK to displace. The maximum effect of control rod motion is thus 16% in k_ .., corresponding to about 6 lb of uranium in the core. SPECIFIC DESIGN PROBLEMS OF THE CIRCULATING-FUEL ARE Dumping all the ARE fuel into a single tank does not result in a critical mass in the tank. The assumptions pertinent to this con- clusion are given in Ref. 14; the most i1mportant assumption 1is the absence of good moderators in or near the tank. ' Two ARE designs with volume frac- tions of fuel in the core of 6.85 and 22.5% were compared as to the sensi- tivity of their critical mass to the addition of potassium (which has a large absorption cross section) to the fuel mixture.¢ %) Of importance (14)C. B. Mills, A Note on Fuel Dumping fronm ARE No. 5, Y-F10-84, Jan. 18, 1952. (IS)C. B. Mills, Effect of Potassium in the Fuel-Cocolant Solution in Two ARE Reactors, Y-F10-85, Jan. 21, 1952. FOR PERIOD ENDING MARCH 10, 1952 during the discussion was whether the existing beryllium oxide blocks should be cut to accommodate the larger fuel percentage. For the 6.85% design, increaseof the potassium fluoride mole fraction in the fuel from 0 to 60% increased the critical mass by 55%. For the 22.5% design, the corresponding increase was 105%. Since the use of potassium may be necessary, the large sensitivity of the 22.5% design to potassium works to the disadvantage of this design. The power generation in a boron carbide curtain on one end of a 3 -megawatt ARE reactor was found to be 0.53 watts/cm?, 20% of which is developed in the first 0.05 centi- meter.¢1%) The problems of fuel-tube-wall corrosion may not be completely solved by the time the ARE is being built, and for this reason, as well as for reasons of fabrication, 1t 1s highly desirable to use thick tube walls. Under assumptions specified in the reference,{!”’? increasing the tube wall thickness from 20 to 40 or 60 mils increases the critical mass by 18.5 and 43%, respectively. (IG)C. B. Mills, The Power Generation in a 346 Curtein ARE on One End of the ARE No. 1 Reactor, Y-F10-87, Jan. 21, 1952. (17)C. B. Mills, Effect of Structure on Criticality of the ARE of Jaeruary 22, 1852, Y-F10-89. R . ~y st 8 W b e 57 T ™ TN T " i . T ™ " TR T T > " T T T T R TP SO Y TY R T R Ty oy ” o O L "reactor have been made. uranium, 5. CRITICAL EXPERIMENTS A, D, Callihan, The group responsible for studies of preliminary reactor assemblies has continued investigations during the past quarter with the mockup of the G-E, direct-cycle reactor described in the preceding report.{(1) Relative evaluations, in terms of contributions to reactivity, have been made of several reflector modifications., Data obtained several months ago from a critical assembly of uranium and graphite have been analyzed, and preliminary plans for experiments on the liquid-fuel-coolant aircraft The data from the graphite assembly have been corre- lated with the theoretical calculations of the assembly. The correlation lacks precision but gives results that are at least consistent with experi- mental fact. | " DIRECT-CYCLE REACTOR(?2) E. V. Haake and D. V. P. Williams Physics Division , W. G. Kennedy Pratt and Whitney Aircraft Division Dunlap Scott ‘ANP Division The preceding quarterly report(1) described briefly an assembly of ' beryllium, methacrylate plastic, and stainless steel that was designed to yield information of value to the General Electric Company (I)A. D, Callikan, “Critical Experiments,’”’ Aircreft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 10, 1951, ORNL-1170, p. 35. (Z)The critical assembly of this reactor will be discussed in greater detail in a report that is now being written to the General Electric Company. Physics Division in the development of the direct- cycle nuclear reactor for aircraft propulsion, A loading chart of this assembly is shown in Fig, 17. During this quarter temperature effects on reactivity have been studied, and some comparisons have been made of the effect on reactivity of the beryllium reflector and of varying relative thicknesses of stainless steel and plastic substituted for the beryllium. A control rod calibration was made, and although the data obtained are unique to the direct-cycle reactor, they are of value for other types of reactors, since they indicate the method and precision of evaluating reactivity changes introduced by changes in reflector, structural elements, and other components, Control Rod Calibration. The change in reactivity introduced by displacement of each control rod has been determined by two methods. In the first, the "period" method, a measure- ment 1s made of the period of the supercritical system resulting from the insertion of a control rod, and the corresponding change in reactivity is determined. In the second, the "rod-drop" method, a safety rod is evaluated from the transient occurring in the flux as the safety rod is rapidly removed from the assembly. Since each control rod is practically coaxial with a safety rod (Fig. 17), the over-all value of corresponding rods is taken to be equal. A comparison of the total rod calibrations obtained from the rod- drop method with the integrated value from the period measurements 1is given in Table 10. Changes in reactivity are expressed in "cents," where 100 cents ("one dollar™) is equivalent to oo TR fifi“ufié i '\l/, Tt Le I W 59 ANP PROJECT QUARTERLY PROGRESS REPORT SECRET - DWG.144324 ABCDEFGHI JKLMNOPQRSTUVW X 0w ~N O D W . - FE5] Be REFLECTOR, JACKET 18 in. LONG, EACH HALF BACKED BY 6in. OF kel GRAPHITE (END). FUEL ELEMENT, EACH {8in. LONG, EACH HALF BACKED BY 6in. OF GRAPHITE. TOP: 6 LAYERS OF STAINLESS STEEL, WITH 5 TO 40 - mil BY 3in. URANIUM DISKS, HORIZONTAL. BOTTOM: 1in. OF PLEXIGLAS, *STAINLESS STEEL, AIR, FUEL PLEXIGLAS LETTERS REPRESENT CONTROL ROD POSITIONS NUMBERS, SAFETY ROD POSITIONS. Fig. 17. Loading Chart of Critical AsSemny of Direct-Cyclé Reactor. 60 © o gsh a < § 0.00730 Ak/k. Tt is believed that the precision of the rod-drop method is no greater than 2 cents, so the agree- ment 1n some cases 1s fortuitous. TABLE 10 Comparison of Control Rod Calibrations REACTIVITY CHANGE o : : PERIOD, CONTROL | CORRESPONDING | INTEGRATED | ROD DROP ROD SAFETY ROD (cents) (cents) A 5 18.2 16 B 6 15.9 16 C 1 - 15.9 18 D 3 18.3 18 Y Data obtained in the incremental "calibration of control rod A by the period method are shown in Fig. 18, which gives the change in reactivity occurring when the rod is withdrawn from the reactor. ' FECRER DWG.14435 Ol.00131 Q.00141/ 000102 0.000876 4 000073 & 5 0.00584 ROD VALVE (cents) Q m b o O 0.000438 0000292 0.000146 . o 0 4 8 12 16 20 24 2 ROD POSITION FROM MiD-PLANE {in.) Fig. 18. ljkéébffifiifi'ashfiLFfifiction' of Control Rod Position. ’ Temperature Effects. During the operation of reactor assemblies an irreproducibility of varying degree has been observed occasionally in day-to-day locations of control rods FOR PERIOD ENDING MARCH 10, 1952 required for criticality under con- stant loading conditions. In the work reported here these variations exceeded the sensitivity required to detect the reactivity differences produced by some structural changes under study. Investigations have indicated the probable cause to be ambient tempera- ture changes. The concomitant re- activity differences, from the re- activity at 72.9°F, are plotted as functions of temperature in Fig., 19, SECRET DWG. 14436 10 |—— ALL VALUES REFERRED TO| // 0.00073 - = O cents AT 72.9°F: S t 100cents = 0.0073 <& ) g K . > I_ S //// '_.. g 0 /7 ° ! o ° z wl o z / I G40 Pa: ' -000073 63°F 729°F 80°F 1 0.9 1.0 14 12 1.3 1.4 MILLIVOLTS [IRON-CONSTANTAN) Fig. 19. Reactivity vs. Temperature. Reflector Studies. An experiment was initiated before the reactor was in final form to compare the reactivity value of a reflector element with that of a fuel element to ascertain the most satisfactory location of the safety rod. A section of the beryllium reflector 22 in., long (one-half of the 44-in. total reflector jacket length) was removed from positions M-21, V-11, and W-11 (Fig. 17), and the resulting change in reactivity was ascertained by a calibrated control rod. The resulting data (Table 11) give the reactivity change (with an estimated error of 1 cent) when a beryllium- filled tube is substituted for an air-filled tube,. Sk 075 S o il e ki e logiees sk cpilda Lo o sl el o TABLE 11 Reactifiity Change Introduced by ' Substituting Beryllium for Air . REACTIVITY CHANGE TEST CELL (cents) M-21 4.4 V 11 10.9 V-11 and W-11 27.5 In a second series of experiments, with final reactor geometry, the ~beryllium in a section 9 by 36 by 6 1in. thick(3®) was replaced by a type- 310 stainless steel section 9 by 36 by 3 ~in, thick., This change was made ‘separately at the side of the reactor, “at the top, and at the bottom. Results ‘giving the change in reactivity ‘effected by the substitution and the “changes in location of the reflector alterations are listed in Table 12 ‘(designations are the same as in Fig, 17). The system was more reactive ‘with the beryllium reflector. ANP PROJECT QUARTERLY PROGRESS REPORT GRAPHITE REACTOR E. L. Zimmerman, Physics Division. Limited effort has been directed to the analysis of data obtained from a uranium-graphite reactor and to cor- relation of the data with theory. This material will be presented when the analysis is complete. Such results as are now available are discussed 1in a subsequent paragraph on the "Corre- lation of Theory and Critical Experi- ments." CIRCULATINGfFUEL REACTOR Dunlap Scott, ANP Division ‘ A preliminary assembly is to be made of the projected aircraft pro- pulsion reactor by using beryllium oxide as a moderator and reflector and molten UF, -BeF,-NaF as a circulating fuel and coolant. This assembly 1s to be operated at room temperature and egsfintially zero power and will utilize TABLE 12 React1v1ty Change Effected by Substituting Plastic and Stainless Steel for Beryllium in the Reflector ~ LOCATION FINAL POSITION REACTIVITY CHANGE IN REACTOR | INITIAL POSITION PLASTIC IN STAINLESS STEEL IN (cents) Side V,W-11,12,13 v-11,12,13 W-11,12,13 ~56. 5 Top L,M,N-2,3 L,M,N-3 L,M,N-2 -41.3 Bottom L,M,N-21, 22 L,M,N-21 L,M,N-22 -33.0 v(3)1t must be noted that since the outside dimensions of the aluminum tubing are 3 by 3 in., the cross section of the units of the reflector (and the fuel) is 2 7/8 by 2 7/8 in. , and the remainder is aluminum (wall) and air-filled space, For convenience in discussion, the 3-in. dimension, and multiples thereof, will be used to deslgnate-the reflector alterations. In the quantitive presentation of the data actual thick- nesses of the material will be stated, with the understanding that the voids are present. 62 beryllium oxide and packed-powder fuel; beryllium fluoride will be omitted from the fuel for convenience in preparation, One possible experimental variable will be the uranium density of the fuel. In the first experiment the density will be that calculated for the molten fluoride mixture for the ARE. Also the powder mixture will have the same uranium-sodium atomic ratio as the design fuel. However, it will probably not be possible to achieve the over- all fuel density of the design fuel in this mockup. Design of fuel con- tainers and control and safety devices is under way. CORRELATION OF THEORY AND CRITICAL EXPERIMENTS D. K. Holmes, Physics Division An attempt has been made to de- termine how closely theory can be made to check with results from critical experiments., The several types of results that are subject to theoretical analysis include (1) criticality with a nonhydrogenous moderator and with a hydrogenous moderator, (2) foil ex- posures, (3) danger coefficients, (4) "rod sensitivity, and (5) the gap experiment, ’ Although criticality calculations for both hydrogenous and nonhydrogenous moderated reactors are acceptably consistent, the results could be improved in the case of a beryllium moderator by the existence of better cross-section data, Multigroup calcu- lations of hydrogen moderated reactors yield high multiplication constants. Danger coefficient data provide a further check of the multigroup method, and the cross sections upon which these calculations are based agree reasonably well, at least in the cases of iron and nickel. Foil measurements also show appreciable correlation with predicted values. However, elementary theory on the loss of reactivity 1in the gap experiment is consistently about 50% low. Criticality with Nonhydrogenous Moderators. Since considerable effort has been devoted to setting up the "multigroup" technique for numerical FOR PERIOD ENDING MARCH 10, 1952 integration of the age-diffusion equation,(*) a very definite attempt has been made to determine whether these calculations can predict experi- mental results. Two separate bare assemblies have been studied, one with beryllium as moderator and the other with graphite. Two assemblies are described in Table 13, The results of the multigroup calculations on these two reactcrs are given in Table 14, - The values of k_ listed in Table 14 encourage the belief that the method and the cross sections on which the calculation 1s based are accurate; however, these results are somewhat offset by the calculations of the ages of thermal neutrons in beryllium and in graphite. Since both these calcu- lations involve the same multigroup method and the same cross sections, the correlation with experimental data would be expected to be similar to that for the multiplication constants. Table 15 shows that whereas the calcu- lated value for graphite may be acceptable, the calculated age for beryllium is not acceptable, The following possible explanations for this discrepancy have been examined: 1. Incorrect experimental value for the age, 2., Incorrect values for the beryl- lium scattering cross section, 3. Neglect of p-scattering 1in beryllium. The measurements of the age in beryl- lium seem to be in good agreement. Changes in either the total scattering cross section or the p-scattering give the age, when contribution such as to correct value for the (4)M. J. Nielsen, Bare Pile Adjoint Solution, Y-F106-18, Oct. 27, 1950. ' oo . Lo Ui 63 ., i s, i o S g e, sl MG ANP PROJECT QUARTERLY PROGRESS REPORT applied to the criticality calculations, lead to values for keff of around 0.90. An additional complication is introduced by the possibility of the existence of a significant cross section for the (n,2n) reaction in beryllium.¢(5) Various experimental estimates for this cross section may lead to increases of as much as 20% in the calculated k for assembly 1. At present the best approach to match- ing the calculations with experimental data is probably that of allowing some p-scattering to raise the calculated TABLE 13 Experimental Data on Critical Assemblies 1 and 4 ASSEMBLY 1 ASSEMBLY 4 Critical dimgnsions Volume fractions bf constituent materials Fuel (93.4% U?3%) Béryl]ium Carbon Stainless steel Aluminum Total hass 93. 4% U235 21 x 21 x 23.22 in. 51 x 51 x 44.111 in. 0.00658 0.0015032 0.90202 0.88916 0.000826 0.0003127 0.03004 0.06073 19.36 kg 57.092" *This figure is the total mass of fuel which would be present if the loading had been carried uniformly to the outside of the assembly. Actually, this was not the case, but the difference is considered to be negligible for purposes of calculatjon since the affected region is only 3 in. wide (as compared with over-all dimension of 51 in.) and is at the very outside where the fuel has its lowest importance. TABLE 14 Calculatéd Results for Critical Assemblies 1 and4 ASSEMBLY 1 | ASSEMBLY 4 koL, . L9912+ eff 0.98 0.9912 Median energy for 1.1 ev 0.15 fission Fraction of fissions | 0.107 0.274 . that are thermal *These figures are based on uranium cross sections corrected for self-shielding owing to fuel lumping (10-mil disks), The removal of the self-shielding factors (calculations may be open to question) made only a I to 2% difference in eff’ 64 TABLE 15 Comparison of Experimental and Calculated vValues ot the Ages of Thermal Neutrons in Beryllium and Graphite BERYLLIUM GRAPHITE (cm?) (em?2) Measured age - 93 350 Calculated age 69 391 (S)W. K. Ergen, On the (n,2n) Reaction in Beryllium with Neutrons of a Polonium-Beryllium Source, Y-F20-12, Apr. 30, 1951. age and also some (n,2n) cross section to bring the k back to unity. In the absence of reliable experimental data on both points, such calculations would be purely speculative. In any case, 1t seems fair to conclude that the method of calculation gives results that are at least consistent with experimental fact. ' Criticality with Hydrogenous Moderators. A modified form of the multigroup method has been used to calculate the multiplication con- stants for four critical assemblies consisting of either bare or reflected cores of solutions of uranium hexa- fluoride in water. An extra term is added to the usual age-diffusion equation to take account of the relatively large energy losses of the neutrons because of scattering in hydrogenous materials. Since the reflectors, when used, were es- sentially infinite and of water, a simple, reflector-saving correction to the bare calculation was used. The results{(®) gave effective multipli- cation constants ranging from 0.97 to 0.988 for the critical assemblies, Since the reactors ranged from 41.5 to 94% thermal and a simple age-diffusion, multigroup calculation gave effective multiplication constants about 20% too high, it may be expected that the method of calculation will give acceptable results, in general, for hydrogenous reactors., Foil Exposures. It is possible to calculate the activity of an aluminum catcher foil placed against a fuel disk, either bare, cadmium covered, or cadmium-indium covered 1f the neutron flux in the reglon and the uranium and cadmium capture cross sections are known as functions of energy. By denoting the activations, bare cadmium ‘covered, and cadmium-indium covered by (6)C. B. Mills, Water Mederated Reactors, Y-F10-78, Jan. 7, 1952. ' comparison, - used, FOR PERIOD ENDING MARCH 10, 1952 AB, Ac, Agp, rTespectively, the follow- ing proportionalities are obtained: : 7 0 E th Ef(E) H(E)dE fc(E) L(E) ¢(E)dE ferE) Ly(E) $(E)dE th where E0 = some high neutron energy, say 10 Mev E,, = thermal energy, zf(E) = macroscopic fission cross section of the fuel disks, H(E) = neutron flux as a function of energy, and fC(E) and fCI(E) are the frac- tions of the incident neutrons of energy E. that reach the fuel disk. From these expressions it may be seen that the extent to which a calculation of foil activities checks experimental results 1s some measure of the accuracy of the calculation of the flux as a function of energy. For purposes of ‘ ratios of activations are The results are given in Table 16 (both calculation and experiment refer to assembly 4, the graphite assembly), The variation of the ratios over the reactor must be attributed to the loading in assembly 4 not continuing uniformly to the edge of the assembly so that the outer 3 in, gave a reflector-like effect; thus a higher fraction of the neutron flux 1s thermal near the edge of the reactor. The calculation represents Tl S bod 073 65 O e bR e e R i bt s - bl A b, | i i a . bk, ANP PROJECT QUARTERLY PROGRESS REPORT an average over the entire reactor. The calculated values reported in Table 16 do not take into account the relative depression of the low energy end of the flux spectrum near the foil upon introduction of the cadmium. Including this effect would tend to raise both calculated values. TABLE 16 Experimental and Calculated vValues for "the Cadmiumand Cadmium-Indium Ratios EXPERIMENTAL OUTSIDE EDGE OF ASSEMBLY | CALCULATED CENTER OF ASSEMBLY Cadmium ratio Ap = — 2.11 2.60 2.72 AC Cadmium-indium AB ratio = —— 3.40 3.90 2.98 ’ ACI ~ _Danger Coefficients. A further check of the multigroup method and the cross sections on which calculations ~are based is available in the danger coefficient measurements, Experi- mentally, the measurement consists of determining the loss in reactivity when a block of some material 1is placed in the center of the assembly. (The comparison is with a void of the same size and at the same position as the block of material; the experiments were performed on assembly 4.) There are two methods of calculation that may be used to check these experiments, In the "difference method," the re- activityis recomputed for the assembly using the multigroup method with the proper additional amount of the particular material added to the assembly., [Since the material 1is, by this procedure, essentially spread o 66 , Q- ‘relatively small, uniformly over the volume of the reactor, whereas in the experiment the material is concentrated at the center of the assembly, the calculated results are multiplied by (7/2)3 to properly weight the importance of the center of the reactor.] The new k . . eff is then compared with the old one obtained before the introduction of new material, Since the volume fractions of the added materials are this method involves taking small differences of large numbers; however, it is felt that the numerical methods used are guite adequate for this case, and the method has the advantage of including the scattering cross section and the & for the added material as well as 1its absorption cross section. The second method is a perturbation technique that involves the calculation of an "importance" function(!) (of energy) for the assembly. The perturbation method allows the material to be placed directly at the center 6f the reactor but, as used for the danger coefficient calculations, takes into account only the absorption cross section of the added material; thus any gain 1in moderation over the void is not included in the perturbation method., 1In Table 17 the total loss in k, upon introduction of a block of the material of the size listed into assembly 4 1s given. An additional effect owing to the lumping of the added material, which would reduce the effective absorption cross section, was not included in the calculation; such an effect would reduce the magnitudes of the numbers calculated. Rod Sensitivity. The control rods for assembly 4 were typical sections about 3 by 3 in. 1n cross section and extended from the center to the edge of the assembly., An experiment ,which is of interest froma theoretical point FOR PERIOD ENDING MARCH 10, 1952 TABLE 17 Experimental and Calculated Values for the Total Loss in keff upon Introduction of vVarious Materials into Assembly 4 Ne X eff 10 SIZE OF BLOCK CALCULATED MATERIAL (in.) EXPERIMENTAL DIFFERENCE PERTURBATION Sodium 3 x3 x1 0.066 0.002 Iron I x 3 x1 1. 44 1.42 2.23 Iron 3 x3 xY4 0.450 0.368 ¢.560 Nickel 3 x 3 x4 0.657 0.700 Mol ybdenum 3x3xY 1.31 0.600 0.960 of view, is the measurement of the incremental sensitivity of such a rod as a function of position in the assembly, 1i.e., 1 Ak “erf eff Ax evaluated when the control rod 1is withdrawn (from the center) a distance x. In the relatively simple case of a very thin poison rod (which leaves behind a negligible void as it 1is withdrawn) the data for assembly 4 fitted the theoretically expected importance function, i.e., 1 Dk, gg ~ cos? yx thin poison rod where x is measured from the center of the reactor, and v = /L, where L is the length of the reactor. However, the "typical-element™ control rod, which does leave a 3 by 3 in. void channel behind as it is withdrawn, showed an entirely different behavior. The sensitivity was fairly constant as a function of x until the rod had been withdrawn somewhat over three-quarters of the total rod length, at which point the sensitivity rose to a maximum (about 20% above its value at the center of the assembly) and then fell rapidly as the edge of the reactor was approached. It seems fairly certain that the removal of a block of moderator of width Ax at a point x in the reactor (leaving a thin void section behind) should give a change in k_ 7 that varies again as cos? Yx; tflus the “peculiar behavior of the rod sensitivity seems most likely to be the result of the void channel, which is of length (L/2) - x when the rod has been with-. drawn a distance x. Thus 1t 1s possible that a term provided by the transport of neutrons along the channel from near the center of the reactor to a region of much lower importance near the outside might account for the ~ T ey g 67 God 081 \J i s, Al 4 e okl where ANP PROJECT QUARTERLY PROGRESS REPORT experimental result. A typical form - ‘ -’ - for the theoretical estimate of this effect is: L Ok ¢y k eff void where P(x’ - x") is the probability that a neutron will make a flight through the void, leave the channel walls at x', and re-enter at x”. This term will depend upon the flux at x’ and the solid angle subtended at x' by a small area at x". I(x' — z") is the change in importance of a neutron while being transported from x' to x”. Since the integral is over all values of x' and x”, account is taken of the gain in importance owing to transport toward the center of the reactor, but this will be smaller than the loss of importance since the flux falls off from the center as cos yx. Numerical evaluation of the integrals shown indicates that the void contribution to the rod sensitivity rises from zero at the center of the assembly to a maximum near the three-quarter point and then falls off rapidly. It is possible to write k Ax total _ A[l Ak] . B{l Ak} | k Ax moderator k Ax void ] = cos? yx moderator ’ 68 has the values for the void effect given by the integration above, and to choose A and B so that an acceptable d X x Ax ~ Ejo“/ov I(x' = x") P(x! = x") dx' dx” fit to the experimental data 1is obtained, Gap Experiment. With calibrated control rods it 1is possible to measure the loss in reactivity upon separating the two halves of the reactor at the center plane, Such an experiment was performed on assembly 4 with sepa- rations up to 0.3 in. and losses in reactivity up to about 0.005. A calcu- lation has been made by using the mul tigroup method and allowing an apparent absorption cross section at each lethargy corresponding to the probability for loss of neutrons from the gap [with a weighting factor of (7/2)% to account for the fact that the losses are actually from the center of the reactor]. The leakage losses from the gap were calculated by using the results given in CP-3443.(7) The calculated losses in k& are uniformly about 50% lower than the experimental results. However, since the maximum separation of halves 1is only 0.3 in., whereas the dimension of a face of the reactor is 51 in., the entire experiment is in the range of "small gap" for which the results of CP-3443 are known to underestimate the leakage from the gap. An improvement of the method of CP-3443 has been made, and the losses are now being recalcu- lated on the new basis. (T)M. G. Goldberger, M. L. Goldberger, and J. E. Wilkins, Jr., The Effect of Gaps on Pile Reactivity, CP-3443, Feb. 20, 1946. Rt e RN [ P PP T R AT TR i) R N N -l o R Yo a ¢ . » o . oo w - M \ s q w @ n « 0 “ ' ’ * \ - ' “ .U e o T— il B I B b e e i o b i st Rt e R e i T T T T R T et o " Ty R ki | R i s SUMMARY AND INTRODUCTION E. P, Blizard, Physics Division The mockup of the divided shield is now being measured in the Bulk Shield- ing Facility (sec. 6). The angular and energy-dependent gamma-ray measure- ments that have been obtained are gratifyingly detailed., It is not certain that the neutron spectral and angular distributions will be as amenable to measurements; the in- struments for these measurements are sti]ll being developed, Research on ducts has included detailed measurement of the effect of duct geometry on neutron transmission, as well as the experimental corrobo- ration of a simplified theory of neutron transmission in ducts (sec. 7). The agreement between theory and experiment for duct transmission is within a factor of 2 for attenuations as high as 10°%., Duct parameters in- vestigated include diameter, length, and angles of a single bend, A comprehensive design of the Tower Shielding Facility, which will make possible full-scale (but not full- intensity) measurements of divided shields has been completed (sec. 8). The resulting configuration, basically a 300-ft tower with a 100-ft cross member for the reactor and crew shield, meets all requirements regarding freedom from spuriously scattered radiation and flexibility. Tt 1s estimated that this facility will cost about two million dollars and that it will be completed 1in the middle of 1953. Additional cross-section measure- ments have been obtained on the 5-Mev Van de Graaff and the time-of-flight neutron spectrometer for use in reactor cross-section measurements has been completed (sec., 9). The spectrometer has been installed and the counting rates optimized and tested on the LITR. Measurements of the total cross section of 1ron on the 5-Mev accelerator extend from 0.7 to 3.6 Mev. 71 a ¢ . » o . oo w - M \ s q w @ n « 0 “ ' ’ * \ - ' “ .U e o T— il B I B b e e i o b i st Rt e R e i T T T T R T et o " Ty R ki | R s oo v 6. BULK SHIELDING REACTOR . Meen » Cochran Haydon . Henry . Holland mRET G W= "TvoH H. E. Hfingerford E. B. Johnson J. K. Leslie F. C. Maienschein G. M. McCammon T. N. Roseberry Physics The divided shield mockup, supplied by the General Electric Company, has been installed in the Bulk Shielding Facility, For these measurements the reactor was reloaded to completely fill the lattice of the reactor and to minimize the effect of the borated water., The gamma-ray spectroscopy 1is well under way, but instrumentation required for the neutron spectroscopy- will not be completed until this summer. ‘ MOCKUP OF THE DIVIDED SHIELD The divided shield mockup consists of a tank cylindrical on the sides and roughly hemispherical in front (Fig. 20). A vertical slot that barely allows clearance for the reactor is cut along the length of the cylindrical section so that the reactor and its supporting bridge may be moved back out of the shield. Cylindrical air voids on the sides provide a region of no attenuation for neutrons and gamma rays. This has the effect of simulat- ing a reactor 4 ft in diameter and extending out to the walls of the air voids. Mounting brackets are provided to hold two, large, roughly hemispheri- cal, lead dishes that can be installed to mockup the lead shadow shield. These lead dishes are not being used for the present experiments. ' For the éxperiments now under way the shield has been filled with borated water (0.4 wt % boron), and all Division measurements are being taken along the center line out from the front face of the reactor, as follows: 1. Center line measurements of thermal-neutron flux, fast-neutron dosage, and gamma-ray dosage, such as were made on the unit-shield mockup, 2. Energy and angular distribution of gamma rays, 3. Energy and angular distribution of neutrons, ' Unit-shield measurements will pro- vide a temporary estimate of measure- ment 1, and measurement 2, the gamma- ray spectroscopy, is well under way. Assuming 140 cm to be a typical shield thickness, measurements have been made with the spectrometer(!’ at various angles with respect to the center line (see Fig. 20). The results have been described by Maienschein(?2?) and are shown in Fig, 21. These data supersede the preliminary spectradata previously reported. (3) (g, . Maienschein, Hultiple-Crystai Gamma- Ray Spectrometer, ORNL-1142 (in press). (Z)F. C. Maienschein; Gamma-Ray Spectral Measurements with the Divided Shield Mock-up, Part I, ORNL CF-52-3-1, Mar. 3, 1952. (3)Figure 5.1, ““Preliminary Gamma-Ray Spectrum at 130 cm from the Water-Reflected Reactor,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1951, ORNL-1154, p. 85. coa (85 73 122 i e il i e, s b, i Kot B i S i il G B L R e e e 22 i S5 A B 322 s MR, a5 it . s s R, e it i, e ~BEGRET DWG. 14237 STEEL SHELL STEEL SHELL -1 BORATED /‘fi AIR PACE REACTOR REACTOR ¢ _ 214.5¢cm AIR SPACE PROPOSED LEAD SHADOW SHIELDS BORATED WATER SECTION A—A DIVIDED SHIELD / PLAN VIEW d o —————————— e ——————————————————— Fig. 20. Relative Position of Reactor, Divided Shield, and Gamma-Ray Spectrometer in Bulk Shielding Facility. 140434 SSTU90Ud ATHALYVAD 1D03A[0¥d dNV ke Lk bl iEe - GAMMA—-RAY FLUX, I (y*s/_cmzléec/Mev/wafi/sterodicn) ‘ o no 104 N 10 Fig. FOR PERIOD ENDING MARCH 10, 1952 v SEGREF— - DWG. 14235 -0 2 4 6 8 10 {2 14 GAMMA—RAY ENERGY (Mev) | ( | j | 21. Gamma-Ray Spectra at 130 cm from the Water-Reflected Reactor. s i \ CER QT 15 S SHUo ke o s bl i ANP PROJECT QUARTERLY PROGRESS REPORT Measurements are now being made at various other distances from the reactor. The reactor will then be moved out of the shield into the open water, and a few of the measurements will be repeated for comparison with the spectra i1in the borated-water shield., Finally, an attempt will be made to measure the spectrum of gamma rays emerging from the face of the reactor, The third measurement, the energy and angular distribution of the neutrons, will be undertaken as soon as the necessary instruments have been developed, which now appears likely to be some time this summer. REACTOR CALIBRATION For the divided-shield experiments the reactor was loaded so as to com- pletely fill the lattice when rolled into the shield as shown in Fig. 20. Furthermore, since the shield con- tained borated water and duplicate measurements were to be made with the ~reactor in the open water behind the shield, the reactor was loaded so as to minimize the effect of the borated o) £ v 76 water and was surrounded on four sides with beryllium oxide reflector as shown in Fig, 22. Fuel elements were added in the interior of the lattice until " criticality was reached with 3.1 kg of U235, The complete interior could not be filled, but two positions (44 and 46) had extra beryllium oxide elements and two positions (43 and 47) were left filled with water. Repeating the critical experiment in the open water necessitated a slight rearrangement of the extra beryllium oxide elements, ~which amounted to exchanging the positions of the extra beryllium oxide elements (44 and 46 above) with that of the water (43 and 47). The final assembly required about 30 g less of fuel. . The power distribution of the two critical assemblies is being measured with gold foils, following the method of Meem and Johnson.(*)> All data on the divided-shield measurements will be normalized to 1 watt using these power calibrations, ' (4)J. L. Meem and E. B. Johnson, Determination of the Pover of the Shield-Testing Reactor -~ JI. Neutron Flux Measurements in the Water-Reflected Reactor, QORNL-1027, Aug. 13, 1951, FOR PERIOD ENDING MARCH 10, 1952 } ’ T - REACTOR GRID PLATE—\ STANDARD FUEL ASSEMBLYW %V R D)) ) s%wm bV oDIDIBD W @fi ST 1% ol s , o bl b ¥ o 3 A ir ¥ * W e ) * ia i I i t w W ' . v ‘s ¥ I v e " x i N .f b ; ¢ i ! e - " T e w ™ R i e Al o kol RL R e s b 7. DUCT TESTS C. E. Clifford M. Hullings F. Muckenthaler .. Abbott A. Simon ~Physics Division Duct work has been directed toward obtaining further experimental cor- roboration of the simplified theory of neutron transmission through cylindrical, air-filled ducts in water. The effects of further vari- ations in duct diameters, lengths, and angles of a single bend (with straight sections of equal length) on neutron transmission through the duct have been measured in the Thermal Column Facility. In addition to the above variations, measurements have also been made to predict the effective source area that contributes to the radiation reaching the exit of a single duct. An increase of a factor of 6 in the calculated source strength resul ted when the source was increased from an area equal to that of the duct mouth to a much larger area. This was measured for only one duct size (3 in. ID) and is reported to indicate the order of magnitude of the effect. This is, of course, very important for extrapolating the single-duct data to a reactor design in which an array is used. ' The agreement with the theory has remained good in that the dose can be predicted for a given duct geometry within a factor of 2 when the geo- metrical attenuation of the duct is as high as 10°%. A report of all the experimental and theoretical work on ducts 1s essentially complete and will be issued during the next quarter. THEORETICAL TREATMENT OF DUCT TRANSMISSION The generalized equation used to predict a measured dose at the end of a duct can be written as follows: b Fflaz { a? Aa? - n . o [ 4] (81,2) 81,% sin 6, Aa? 8l ,, sin &, where D = relative dose {(thermal-neutron flux), n, = relative source strength per unit area, a = diameter of duct, !, = length of mth straight section, A = experimentally determined constant, including neutron albedo, etc., 9m = angle between mth and mth + 1 straight section. For the.sfiécial case of equal- length straight sections with bends of equal angles,the equation becomes: B 2 “ "4 ) [8l2) (812 sin 6] "where m 1s the number of straight sections. This case gives minimum transmission for a given center line length, It should be noted that the equations are not valid for angles of bend that areso small that neutrons can pass directly through the duct Ty o~ o e S g S Y i d 79 &Hi ANP PROJECT QUARTERLY PROGRESS REPORT ‘without a wall collision. The formula becomes increasingly more accurate as the angle of bend approaches 90 degrees. MEASUREMENT OF AIR-FILLED DUCTS IN WATER For correlation with the above theory, measurements were made with various configurations of air-filled, cylindrical ducts 1n the Thermal Column water tank, including straight ducts and ducts with single bends up to 90 degrees. The source areas were also varied. The measured results have been compared with the preceding theory and are within a factor of 2 - for a given duct geometry. Straight Ducts, One end of the duct was placed against a fission source {uranium slugs) in the Thermal Column water tank, and the counter was located in the various slots of ‘the counter holder at the other end of the duct. (These ducts had 1/4-in. lucite walls and were closed at each ‘end with a 1/4-in. lucite disk.) The fission-source box included a movable cadmium shutter to cover the source ‘for background measurements. Measurements were taken on all straight-duct configurations with the full source, which consisted of 24 natural uranium slugs, 1 1/8 by 4 1/8 in., in a rectangular array (9.3 by 12.4 in.). Additional measurements were made on the 6-in. -ID ducts using a 6-in.-~diameter circular source by placing a cadmium shutter with a 6-in.-diameter hole beneath the slugs. Typical curves for the neutron trans- mission, as measured along the center line of the 6-in.-ID ducts, are given -1n Fig. 23, and traverse measurements are shown in Fig. 24. Ducts with Bends., The transmission of neutrons through a flexible rubber 80 6 hose (60 in. in length, 1/4-in. walls) with single bends of 0 to 90 degrees was measured with both the full rectangular source and a 3-in,-diameter circular source, The ducts were sealed with 1 1/2-in. rubber plugs to give an actual air column of 57 inches. The two straight sections shortened as the angle of bend increased, since the hose was always bent with a 12-in, radius of curvature, For each angle the straight sections were of equal length. The smallest angle of bend was equivalent to a displacement of the center of the duct by one diameter to prevent neutrons from reaching the counter without having made at leastone collisionor having penetrated the surrounding water. Center line and traverse measurements for the full source are shown in Figs. 25 and 26. Center line measurements were also made on a 4 1/4-in.-diameter duct with aluminum walls. Comparison with Theory. 1In order to compare the measured results with the predictions of Eq. 1 or Eq. 2, an effective source strength must be determined. To simplify the calcu- lations and also to give some 1dea of the change in duct transmission with energy, the effective source strengths have been arbitrarily defined as the relative thermal flux (the response of a specified counter at 10, 20, and 30 ¢m of water from the end of the duct) multiplied by the geometrical attenuation of the duct, that is, D/n, from Eq. 1 or Eq. 2. It was assumed that the counter was large enough to give a reading pro- portional to the integral of the flux leaving the duct. This can be shown to be nearly correct by a comparison of integrals under the traverses with the center line readings for various ducts. If the formula 1s correct, n, should be a function only of the 0 distance from the duct end, provided, -, L ¢ r ,.m-, [N ¥ i t [N . T -}s— lex et L 'FOR PERIOD ENDING MARCH 10, 1952 SSECHRET— DWG. 14438 DUCTS: 6 in. ID, PLASTIC WALLS 6 L=24in. 4 A L=36in, ® L=48In. 2 B PURE WATER FOR L =24 in. SOURCE: RECTANGULAR,9.3 X 12.4 in. COUNTER AT x = ¢ — P O®S ™ — LOW-PRESSURE, 25 in.,BF3 COUNTER RESPONSE { counts /min) O OO 0 to 20 30 40 50 z,COUNTER DISTANCE FROM END OF DUCT (cm) Fig. 23. Center Line Measurements of Neutron Trans- mission Through Cylindrical Ducts in Water. L=h nor &w% L8e 81 O ANP PROJECT QUARTERLY PROGRESS REPORT ~SEoRET DWG. 14439 3 > 105 8 6 4 \— + DUCTS: 6in. 1D, PLASTIC WALLS _ \ | @ L =24In. A L= 36in. O L=48in V PURE WATER FOR L = 24in. SOURCE : RECTANGULAR, 9.3 X 12.4 in. —\1 COUNTERAT z=6.3 cm — A O OO w N | N LOW-PRESSURE, 25-in.,BF5 COUNTER RESPONSE (counts/min) H o O | n 10 ' ‘ ' -30 -20 -10 0 10 20 30 40 ‘3 ¥, COUNTER DISTANCE FROM DUCT CENTER LINE (cm) Fig. 24. Traverse Measurements of Neutrons in Watef Beyqnd Cylindrical Ducts. o o 82 12Y, - in.,BF3 COUNTER RESPONSE ( counts/min ) FOR PERIOD ENDING MARCH 10, 1952 S ECRET- _ DWG. 14440 105 8 ' , 8 DUCT: 3-in.IDx 60in.; RUBBER WALLS; - 1Y -in. RUBBER PLUG IN EACH END. 4 STRAIGHT N 12,5 deg. BEND 2 G 33.25 deg. BEND 46.5 deg. BEND 104 60.6 deg. BEND 8 ‘ - X 90 deg. BEND 6 SOURCE : RECTANGULAR,9.3 x12.4 in, : COUNTER AT x= ¢ 4 2 103 8 6 0 t0 20 30 40 50 . 80 z, COUNTER DISTANCE FROM END OF DUCT (cm) Fig. 25. Center Line Measurements of Neutron Trans- mission in water Through Cylindrical Ducts with Variable Bends. ' ' ‘ * 83 i B i d g i b e i 50 S s Ml . £ ey oA o i n R, ANP PROJECT QUARTERLY PROGRESS REPORT 84 12Y, -in.,BF5 COUNTER RESPONSE (counts/min.) SECEER- DWG. 14441 n o ©®O [4Y] D Mo » mooa o DUCT: 3in.ID, X 60 in., RUBBER WALLS 1Y, -in. 2 RUBBER PLUGIN EACH END ® STRAIGHT A 13.5 deg.BEND { 33.25 deg. BEND B 46.5 deg. BEND ¥ 60.6 deg. BEND E: RECTANGULAR, 9.3 X 12.4 in. COUNTER AT z=88 cm SOURC -30 -20 -10 0 10 20 - " x,COUNTER DISTANCE FROM DUCT CENTERLINE (cm) Fig. 26. Traverse Measurements of Neutrons in Wafer Cylindrical Ducts with variable Bendsh 30 Beyond of course, that the source i1s either the same size as the duct mouth or es- sentially infinite in extent, That the formula is accurate to within a factor of 2 may be seen in Table 18, for straight ducts, and Table 19, for ducts with bends. These tables summarize the calculations of n, for each duct geometry measured to date. The measurements will continue on a series of 4 1/4-in. cylindrical ducts. The effective source strength, n,, is plotted as a function of the equivalent centimeters of water between the counter and the end of the duct TABLE - FOR PERIOD ENDING MARCH 10, 1952 in Fig., 27. The lower group applies to the 3-in. duct with the 1 1/2-in. rubber plug adjacent to the source, whereas the upper group represents data from ducts that did not have rubber plugs. The difference is not simply ascribable to the attenuation of the plug since this correction had already been applied. No satisfactory explanation has yet been obtained. Further work is being done to clarify the above enigma. The increase 1in n, upon increasing the source size beyond the area of the duct mouth 1is shown in Fig. 28, which also shows angular correlation. 18 Comparison of Calculated Effective Source Strength for Straight, Cylindrical Ducts Low~pressure, 25-in., BF; counter Calculated from counter responseé at 10, 20, and 30 cm of water from end of duct g g iR ol RELELL Ll Skl i DUCT - EFFECTIVE SOURCE STRENGTH DIAMETER LENGTH n, (neutrons/cmz/sec) (in. ) (in.) SOURCE 10 ¢m 20 cm 30 em 6 24 Full(®) 1.05 x 103 1.29 x 102 1.58 x 10 36 Full 2.64 x 102 1.14 x 102 1.51 x 10 48 Full 7.925 x 102 1.08 x 102 1.59 x 10 6 24 6 in.(©) 4.1 x 102 5.62 x 10 7.96 36 6 in. 3.43 x 107 5.28 x 10 7.92 48 6 in. 3.04 x 102 4.91 x 10 7.96 8 924 Full 7.4 x 1072 9.99 x 10 1.33 x 10 36 Full 7.99 x 102 1.04 x 102 1.33 x 10 48 Full 7.1 x 102 9.62 x 10 1.33 x 10 (a) Includes plastic ends. {8) (¢) 6-in,~diameter circular source. o £ o5 ~,< {4t v { P e 24 uranium slugs, 1 1/8 by 4 1/8 in., in rectangular array, 9.3 by 12.4 inches. 85 ANP PROJECT QUARTERLY PROGRESS REPORT TABLE 19 | Comparisohflof Calculafed Effective Source Stréngtfi for Bent, 3 Cylindriqal Ducts Low-pressure, 25-in., BF; counter Caiculated from counter responses at 10, 20, and 30 cm of water from end of duct® DUCT DUCT CENTER LINE ONE-BEND EFFECTIVE SOURCE STRENGTH . DIAMETER LENGTH ANGLE ny (neutrons/em”/sec) (in.) (in.) SOURCE (deg.) 10 ¢cm 20 em 30 cm 3 57 Full(® 0 6.7 x 102 9.18 x 10 | 1.29 x 10 Full 13.5 1.32 x 10° 1.87 x 10%| 3.57 x 10 Full 33.25 1.37 x 103 1.22 x 102| 1.68 x 10 Full 46.5 1.26 x 103 9.1 x 10 9.0 Full 60.6 1.45 x 10° 9.25 x 10 9.2 ' Full 90 6.6 x 10 6.1 x 10 3 57 3 in.(€) 0 1.1 x 102 1.64 x 10 4.22 3 in 13.5 1.71 x 102 2.60 x 10 6.05 3 in. 33.25 1.57 x 102 2.14 x 10 3.29 4.25 40 3 in. 0 2.68 x 102 4.4 x 10 6.95 : 3 in. 90 3.99 x 102 3.92 x 10 4,31 {a) (b) - (0)3 86 -in.-diameter circular source. Includes rubber plugs on all 3-in. ducts. 24 uranium slugs, 1 1/8 by 4 1/8 in., in rectangul ar array, 9.3 by 12.4 in, 'FOR PERIOD ENDING MARCH 10, 1952 v . ~SEERET— . DWG. 14442 104 I I I 1 8 6-in.-1D DUCT, 6-in.-dia. CIRCULAR SOURCE _ 6 O 24 in, — A 36 in. - 4 0 48 in. T 3.in.-1D x 57-in. DUCT, 3-in.~ dia. CIRCULAR SOURCE —— | X STRAIGHT 2 V 13.5 deg. BEND — ¥ 33.25 deg.BEND 03 4Y4-in.-1D x 40-in. DUCT, 3-in.- dia. CIRCULAR N | SOURCE - | 8 , © STRAIGHT - | 6 0o o 90 deg. BEND - RELATIVE SOURCE STRENGTH ( neutrons/cm2.sec) a0 0 10 20 30 EQUIVALENT CENTIMETERS OF WATER FROM END OF DUCT Fig. 27. Comparison of Relative Source Strengths (no) from Various Ducts. Source area equal to duct mouth area. ol 098 " o BoF 3 o s Sk b i R o ol i3 i ki, e, 2, ; ANAP PROJECT QUARTERLY PROGRESS REPORT 88 RELATIVE SOURCE STRENGTH (neutrons/cm2-sec) m SR DWG. 14443 104 8 6 DUCT: 3-in.-IDX 57 in. 4 FULL SOURCE : ® STRAIGHT A 13.5 deg. BEND 2 B 33.25 deg.BEND V 46.5 deg. BEND § 60.6 deg. BEND 103 CIRCULAR SOURCE : 3-in.- dia. 8 ¥ STRAIGHT 0 10 20 ' 30 EQUIVALENT CENTIMETERS OF WATER FROM END OF DUCT *-Fig.‘28. Comparison of Relative Source Strengths (n,) from Various Geometries of the 3-in. Duct. ; Co L e o 0T . T * ot - . ¢ i v o d T s . - PRI Po%e o ] uow b 8. TOWER SHIELDING FACILITY E. P. Blizard C. E. Clifford Physics Division The preliminary proposal(1) for the Tower Shielding Facility has been con- siderably revised, and a firm proposal "will soon be submitted to the AEC. ‘Calculations of spurious scattering from the ground and structure indicate that the shield components (reactor and crew shields) should be at least 200 ft from the ground and free of scattering material in the immediate vicinity of either component. As a consequence, the original configuration has been discarded in favor of the present design, which calculations indicate will satisfy the requirements. A new cost estimate for this Tower Shielding Facility, based on much more complete analysis, indicates that the total capital expenditure will be about two million dollars. TOWER FACILITY DES TGN The tower will consist of two vertical members 300 ft high connected at the top by a 200-ft bridge., At the center of the bridge is a 100-ft cross member from the ends of which will be suspended a shielded reactor on one side and a crew shield on the other (Fig. 29). The distance from the reactor or crew shield to the ground will be variable up to about 250 ft; above 250 ft the scattering from the cross member would interfere. The normal operating altitude will be about 200 feet, The variable height will make 21t possible to identify and (g, p. Blizard, C. E. Clifford, H. L. F. Enlund, J. L. Meem, and A. Simon, “Tower Shielding Facility Proposal,?’ Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 10, 1951, ORNL-1170, p. T73. £y o oo P vy gm;z 3 3 o J. L. Meem A. Simon measure the ground scattering. This is a very welcome potentiality since the calculations of this effect may not be accurate and it can thus be assured that the divided-shield measurements do not incorporate an unknown spurious component. In addition, 1t will be possible to estimate the added crew exposure from ground scattering that will be incurred on take-off and landing. It will be possible to mount large pieces of structural material (e.g., aluminum) at various locations around the reactor or crew shields to simulate airplane structure. Thus the radiation scattering from the airplane itself can be estimated on the basis of direct experimental evidence, EXPERIMENTAL PROGRAM The program for the Tower Shielding Facility as presently conceived will consist of four parts: 1. Calibration of the facility and determination of background from spurious scatterings, 2. Measurement of a shield mockup for the X-6 airplane (Fig. 30}, 3. A general study of divided shields, including optimizationwith respect to weight and parametric studies of reactor-crew separation, weight distribution, diameter of reactor shield, size of crew compartment, etc,, ~ 4, A study of the scattering to be expected from aircraft structure, engines, and ground. 89 L i e . 5. A e ik 90 —RESFRICTED~ DWG. Ct2107 N # A & " o W W - M ‘# 1"/4 W Fig. 29. Proposed 300-ft Tower Shield Facility. sk iAo\ e e S D A o o s 1. i i i, MG, ) i, Ll s i e Ml N €« owo - i . . - R B 2 ' o SECRET— : DWG. C-9394 Rt I A REACTOR =N CONTROLS ™ T WATER” ~ z LEAD —t = SHADOW DISKS — DUCTS REACTOR CREW SHIELD REACTOR SHIELD N TOWER STRUCTURE—”y 16 Fig. 30. X-6 Aircraft Shield Mockup on Tower Shielding Facility. ‘0T HOUVIN ONIONA aorydd H¥od cS61 * 9, NUCLEAR MFEASUREMENTS A. H. Snell, Physics Division The total cross section of iron has been measured on the 5-Mev Van de Graaff from 0.7 to 3.6 Mev with a resolution of approximately 35 kev. Measurements of the inelastic scatter- ing levels in iron are now under way with the accelerator. The time-of- flight neutron spectrometer has been installed in the LITR and used (after optimization of the background-to- count ratio) to scan the transmission of indium at a resolution of 1.25 usec per meter. | MEASUREMENTS WITH THE 5-Mev VAN de GRAAFF ACCELERATOR H. B. Willard C. H. Johnson J. K. Baair J. D. Kingdon Physics Division During the last quarter the 5-Mev Van de Graaff accelerator has been used to obtain detailed information on the total cross section of iron and the inelastic scattering levels excited in iron by fast neutrons. In addition, equipment has been assembled and preliminary measurements have been made on the total cross sections of Li® and Li7 and the fission cross U235 U238u sections of and Total Cross Section of Iron. The total cross- sectlon curve for iron was measured w1th a resolution of 35 kev The dotted curve of this ~ figure indicates the real variation (Fig. 31). (outside the statistical error of 3%) of cross sectlon with energy and shows many unresolved resonances., Inelastic Scattering Levels in - Iron. Measurement of 1nelastic scattering levels in iron by fast neutrons places the first level excited by this process at 0.83 Mev above the ground state. No absolute cross sections have been obtained. TIME-OF-FLIGHT NEUTRON SPECTROMETER G. S. Pawlicki, ORINS E. C. Smith, Physics Division The time-of-flight neutron spec- trometer for operation up to several thousand electron volts has been in- stalled in the LITR and checked with indium. When the hard background level was attenuated by a 3-in.-thick beryllium filter, the transmission of indium was satisfactorily scanned at a resolution of 1.25 psec per meter. Background Measurements. The instrument for background measure- ments, as originally installed in hole HB-1 of the LITR, looked directly at fuel elements, and the transmission of the shutter was not up to design expectations., A spectrum check made with the instrument indicated a surplus of neutrons above a 1/E distribution at energies greater than 1 kev. A large fraction of these high-energy neutrons was subsequently shown to have an energy above 0.1 Mev. As a temporary measure an aluminum filter was first used that did improve the background-to-count ratio at some sacrifice in counting rate and defi- nitely confirmed the suspected reason for the poor background observed. By reloading the lattice to provide 3 in., of beryllium in the beam, the background was reduced to design expectation without loss of counting rate. The tabulation in Table 20 shows the counting rates and backgrounds 93 e T T T T TN T W TR T T e o B B il & i g UNCLASSIFIED _ DWG.14444 5 ‘ 111" 1711 T 1T 17 1T°"7T 1 1 1 1 | 5 . 4 — — T — 5 — 3 l— _ 3 4°— Q b T | | | 2 — b! — ‘.} ~——— MEASURED CURVE | ! AVERAGED CURVE ] ¢ — _ | o o e 0 1.0 2.0 3.0 4.0 NEUTRON ENERGY (Mev) Fig. 31. 'Total Cross Section of Iron. TABLE 20 " Improvement of Count-to-Background Ratio with Aluminum and Beryllium “Filters COUNTS BACKGROUND {No./min) {(No./min} Original loading 209 38 Aluminum filter 149 6 3-in. beryllium filter 256 12 94 observed for 20 wsec channels at a resolution of 6.8 usec per meter, Indium Resonances. With the re- loaded lattice the transmission of indium has been surveyed at a resolution of 1.25 pusec per meter, The three known resonances at 1.43, 3.9, and 9,5 ev were scanned, In addition, resonances were observedat 10.6, 15.4, 23.8, 48, 78, and 90 ev., The region of 3.9 ev and higher will be repeated with separated isotope samples of known purity and enrichment, A L e W x 4 SUMMARY AND The research on high-temperature liquids has been almost entirely con- cerned with the development of a fuel- coolant for the circulating-fuel reactor, and the effort on liquid moderators and coolants has been correspondingly reduced (sec., 10). The proposed reactor loading technique requires that the circulating-fuel solution permit widely varying uranium concentration with near uniform melting point. However, for the added requirement of low (< 10 cp) viscosity, the previously proposed systemn, NaF-BeF, -UF,, would provide a suitable fuel. Of the other available fluorides, however, several compositions in the system NaF-KF-ZrF, -UF, appear promising on the basis of the limited data available., In addition to this system, numerous other systems containing zirconium fluoride are being examined. The effort on corrosion research was divided among static and dynamic tests of fluoride and hydroxide corrosion and static tests of liquid metal corrosion (sec., 11). The severe corrosion experienced in dynamic tests with fluorides at 1500°F -~ despite relative inertness 1in static tests - has redirected attention to fluoride corrosion phenonema. Although there is attack of up to 14 mils, the 0.622- in,~-ID Inconel convection loops will circulate the fuel, NaF-KF-LiF-UF4, for the duration of 500- and 1000-hr tests without plugging. On the other hand, the 300-series stainless steel loops have a life expectancy (before plugging) of 150 hr and the 400-series stainless steel loops plugin less than 50 hours, It is known that the uranium-free fluorides are less cor- rosive than uranium-bearing fluorides and that in tests with the latter uranium dioxide crystals form in the cold zone, The effect of various additives on both fluoride and hy- droxide corrosion is now being evaluated INTRODUCTION and dynamic corrosion of hydroxides 1is being determined in various environ- ments. In one such experiment with potassium hydroxide in Inconel under hydrogen, there was no corrosion or mass transfer after 135 hr at 715°C. The metallurgical processes involved in the construction and assembly of a high-temperature reactor core, includ- ing fabrication of control rods, welding and brazing of core structure, and creep and stress-rupture of metals, are being investigated (sec. 12)., The cone-arc welding method may be adapted for tube-to-header welds in the core, but the complexity of the heat ex- changers suggests the use of a brazing technique. Several boron-free brazing alloys are being tested to determine their gqualities as brazing material and their corrosion resistance. ARE safety rods will probably be con- structed of type-430 stainless steel because of its compatibility with boron carbide, Stress-rupture and creep data have been obtained on fine- and coarse-grained Inconel sheet. Heat transfer research and physical properties measurements have been largely directed to the immediate needs of circulating-fuel reactor studies for these data. Measurements of viscosity, thermal conductivity, heat capacity and vapor pressure of one or more of the various fluoride mixtures are being obtained. The theoretical analysis of heat transfer in a circulating-fuel reactor has established the design parameters for the core fuel circuit (assuming no wall heat transfer). Mathematical natural convections for ligquid fuel elements have been developed for the case of turbulent flow. Some data have been obtained on the heat transfer of boiling mercury, and data should soon be available on the heat transfer of fused salts, hydroxides, and lithium. 97 . 'L ; k r L - o T e e e s i & k k- Kbk piadi, bbb e i i, S e ANP PROJECT QUARTERLY PROGRESS REPORT Radiation damage studies included irradiated fuel capsules and inpile liquid loops and measurements of the effect of irradiation on creep and thermal conductivity. In several of the inpile experiments with fluoride mix- tures in Inconel capsules, the rate of attack on the container material has been considerably higher than that observed i1n the out-of-pile controls, 98 Radioactive decay curves were taken while sodium was being circulated in the inpile loop for 165 hours. Partial confirmation of the previously observed decrease in the thermal conductivity of Inconel at 1500°F under irradiation has been obtained. The creep of nickel under irradiation has a higher rate than 1ts bench counterpart after about 115 hours. e n w FOR PERIOD ENDING MARCH 10, 1952 10. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS Warren Grimes, Materials Chemistry Division The research on high-temperature ligquids has been concerned almost entirely with their development for use as fuels and/or moderators for an aircraft reactor. The properties required of these liquids along with periodic descriptions of progress in development of such materials have been the subject of previous re- ports.(!} During the past guarter, however, effort on reactor coolants and on liguid moderator materials has been markedly reduced, in spite of the potential long-range importance of the latter, to increase the effort on “various aspects of the fluoride program. The interest in a circulating-fuel - ARE has emphasized a need forliquids of very low melting point at low uranium concentration., In addition, it is very desirable to choose the fuel system so that a dilute (sub- critical) fuel solution may be intro- duced to fill the core, and small increments of a concentrated solution of uranium in the same solvent may be added to bring the system to criti- cality. This requires that solutions widely varying in uranium content and melting at temperatures well below the operating range be available and that no high-melting compounds be formed at intermediate concentrations. ' The qualifications seem to be met adequately by the system NaF-Ber-UF4. (I)W. R. Grimes, **Chemistry of Liquid Fuels,” Aircraft Nuclear Propuision Project Quarterly Progress Report for Period Ending March 10, 1951, . ANP-60, p. 127; W. Grimes, *“Chemistry of Liquid Fuels,” Aircraft Nucleaer Propulsion Project - Quarterly Progress Report for Period Ending June 10, 1951, ANP-65, p. 84; W. Grimes, “Chemistry of High-Temperature Liquids,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1951, ORNL-1154, p. 154; W. Grimes, ¢*“Chemistry of High-Temperature Liquids,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending Decenber 10, 1951, ORNL-1170, p. 79. However, 1t appears that the viscosity of this liquid in the lowest melting region of the system is too high., Of the other available materials, several compositions in the system NaF-KF«ZrF4- UF, appear promising on the basis of the limited data available. Numerous other systems containing zirvconium fluoride are being examined. Studies of the 1onic species 1in molten fluorides by determination of transference numbers have not proved capable of unique interpretation. General planning for production of the simulated fuel for the "cold" critical experiment has been completed and experimental preparation of fuel and fuel assemblies is under way. LOW-MELTING-FLUORIDE FUEL SYSTEMS J. P. Blakely R. E. Traber, Jr. L. M. Bratcher C. J. Barton Materials Chemistry Division The previous report{?) presented a preliminary list of five promising fluoride fuel systems of low uranium concentration, of which four contained lithium fluoride as a component, Since lithium fluoride would be attractive as a component only if the heavier isotope of lithium were to become available inhigh concentration, only the NaF-BeF,-UF, system appeared promising for immediate use, However, the toxicity of beryllium fluoride and the preliminary indications of high viscosity of beryllium fluoride-bearing liquids have prompted efforts to find other systems for immediate use. Promising results have been obtained recently with mixtures containing zirconium fluoride, (2)J. P. Blakely, L.. M. Bratcher, and C. J. Barton, “lLow Melting-Fluoride Fuel Systems,” op. c¢tt,, ORNL-1170, p. 79. 99 T T n T ANP PROJECT QUARTERLY PROGRESS REPORT It has been necessary to study the binary and ternary systems of zir- conium fluoride with the alkali fluorides, Work on these uranium-free materials is reported under "Coolant Development™ in the following. Ad- dition of small amounts (1 to 5 mole %) of uranium fluoride to several of the systems has been shown to affect the melting points only slightly, and it appears likely that any of the low- melting coolants obtained from the preliminary studies can be readily converted to fuels., The available data on zirconium fuels are reported under appropriate headings in the following, along with additional data on several systems containing beryllium fluoride. - NaF-BeF,-UF,. Difficulty in es- tablishing reproducible melting points by thermal analysis of mixtures of high beryllium fluoride concentration has been discussed,(%?)> 1In an effort to overcome this difficulty, heating curves have been compared with cooling curves for a number of compositions in this system. Selected mixtures were heated to 850 to 900°C and cooled, with stirring, by techniques previ- ously described. The solidified materials were then reheated and the heating curve, measured with chromel- alumel thermocouples, was recorded in the manner previously described. The data shown in Table 21 afford a comparison between break temperatures TABLE 21 CompariSbn of Break Temperatures from Heating and Cooling Curves of the System NaF-BeF,-UF, ' COMPOSITION (mole %) BREAK TEMPERATURE (°C) NaF BeF, UF, ON HEATING ON COOLING 48 52 0 345, 375, 380 332"" s 51.5 1 330, 365, 385 350" o 51 2 365, 385 340"" 6.5 50. 5 | 3 245, 315, 369, 385 355" " 46 “ 50 4 295, 325, 385 3427" 55 49.5 5 270, 360, 395 352" '40' | 60 0 230, 300, 330, 365 250, 390 ‘39.5“ 59. 4 1 290, 330, 355 , 270, 415 39.2 58. 8 2 325, 355 295, 445 a8 58. 2 3 325, 370, 400, 405 315, %" 430 38.4 57.6 4 325, 365 285, 290, 315 38.0 57.0 5 330, 360 317 e : § Highest break temperature considered most reliable indication of melting point. *-,*. . . - < - ¥ . Supercooling definitely indicated. 100 S e e e W e v on et # taken from heating curves and cooling curves, which were made with the same materials and equipment. The highest temperature at which a break is shown in the heating curve is considered the most reliable indication of melting point available. As the data indicate, melting points for this system are Jittle affected by uranium concen- tration in the range studied. It appears that mixtures of useful uranium concentration may be prepared with melting points below 400°C. KF-BeF,-UF,. A previous document( 3) presented some preliminary data on the system KF-BeF, -UF,. Results of ad- ditional studies with these materials are shown in Table 22. All the ‘mixtures studied show melting points considerably higher than those avail- able in the NaF-BeF,-UF, system, TABLE 22 Break Temperatures from'Cooling Curves of the Systen KF-BeFZ-UF4 FOR PERIOD ENDING MARCH 10, Break Temperatures from Cooling Curves of the System RbF-Ber—UF4 TABLE 23 1952 COMPOSITION (mole %) BREAK TEMPERATURES™ RbF BeF, | UF, (°c) 40 60 0 375, 420 39.4 59.1 1.5 380, 440 38.2 58. 2 3 325, 370, 445 38.0 57.0 5 365, 442 36.0 54.0 10 430, 544 COMPOSITION (mole %) BREAK TEMPERATURES” KF | BeF, UF, (°0) 52 48 0 400, 475, 520 51.2 | 47.3 1.5 397, 514, 533, 563 50.5 | 46.5 3 390, 472, 525 49.5 | 45.5 5 385, 510, 600 44.5 | 40.5 10 385, 525, 607 ‘Table 23. » * Highest break temperature considered most reliable indication of melting point. *3 ‘ . Temperature uncertain, * Highest break temperature considered most reliable indication of melting point. than those for NaF-Ber-UF4 mixtures, the RbF-BeF,-UF, system seems to show promise, especially below 5 mole % uranium tetrafluoride. The physical properties of this system are probably very similar to those of analagous ‘sodium fluoride-bearing systems. LiF-NaF-BeF,-UF,. The above- mentioned heating technique has been applied to various mixtures in the system LiF-NaF-BeF,-UF, in an attempt to verify the very low melting points tentatively reported(?’) for these materials. Although it is possible that mixtures melting below 325°C may be prepared from these materials, the data previously reported seem to be much too low. Additional studies of this system will probably be postponed in favor of more immediate problems. Fuels Containing Zirconium Fluoride, Data froma number of studies of alkala fluoride—zirconium fluoride mixtures ..is reported in a following section on - BbE-BeF,-UF,. Additional data on mixtures of low uranium content in the system RbF-BeF,-UF, are presented in Although it appears that the melting points are somewhat higher (3)J. P. Blakely, L. M. Bratcher, and C. J. Barton, “Low-Melting Fluoride Systems,” op. cit,, ORNL-1154, p. 155, "Coolant Development." Only a limited number of experiments on mixtures con- taining uranium tetrafluoride have been completed and the results are summarized in Table 24. It seems likely from this data and that pre- sented under "Coolant Development™ that the NaF-KF-ZrF,-UF, system may be 101 e o i it I s, e b bt s PR RN Sk 5 ANP PROJECT QUARTERLY PROGRESS REPORT of definite value as a fuel. Measure- ments of viscosity have not yet been made for this material, but microscopic examination of the solidified melt has indicated a complete absence of glasses. It 1s likely that the viscosity of the liquids will resemble those of typical fused salt mixtures rather than glasses. Additional experiments with this and similar systems are under way. TABLE 24 Break Temperatures from Heating and Cooling Curves of Zirconium-Bearing ' Fuel Mixtures COMPOSITION (mole %) | prpay TEMPERATURES® NaF | KF | ZrF, | UF, (°c) 37.5 | 20 |42.5| 0 395, 465, 475 37.1 | 19.8 | 42.1] 1 263, 380, 460 36.7 | 19.6 | 41.7| 2 395, 410, 450 36.3 | 19.4| 41.3| 3 390, 410, 440 35.9 | 19.2| 40.9| 4 375, 451, 505 (?) 35.5 | 19.0 | 40.5| 5 375, 400, 455 5 52 | 43 0 305, 405, 425 4.9 | 51 | 42.1| o0 360, 380, 425 X Highest break temperature considered most re- liable indication of melting point. SIMULATED FUEL MIXTURE FOR COLD | CRITICAL EXPERIMENT D; R. Cuneo L. G. Overholser Materials Chemistry Division The cold critical experiment is ‘designed to yield data at room temper- ature that will be of value in calcu- lating the critical mass of the final high-temperature ARE. The primary requirement of the simulated fuel is that the density of uranium in the tubes of fuel mixture be uniform and equal to the design density of uranium in the ARE fuel at operating temperatures, 102 In addition, it is desirable to have the sodium-to-uranium ratio equal to that in the real ARE fuel and, if possible, to have the over-all density of the simulated and real fuels the same. The hydrogen-to-uranium ratio must be considerably less than 1; this will be possible i1f the water content of the materials is less than 0.01%." It is presently anticipated that the ARE fuels will be simulated for this experiment by packing realistic fuel tubes with a mixture of uranium tetrafluoride and sodium fluoride powders. The Y-12 Production Division will be responsible for preparing the enriched fuel and for filling the fuel tubes. The ANP chemistry group will render assistance in formulation of the mixture and experimental study of tube filling and packing conditions, Preliminary experiments with uranium tetrafluoride and sodium fluoride powders designed to yield information as to effect of particle size, mixture composition, water content, and pack- ing technique on uranium density, over-all density, and ease of segre- gation of the mixture are under way. IONIC SPECIES IN FUSED FLUORIDES M. T. Robinson Materials Chemistry Division Determination of the electrolytic transference numbers in fused fluoride mixtures containing uranium tetra- fluoride was attempted in an effort to ascertain the nature of the ionic species in these systems. The equip- ment and general technique used in the study of electrolysis of fused fluorides have been discussed in a previous report. (%) Uncertainties owing to the use of an electrically conducting (4)M. T. Robinson, ““Jonic Species in Fused Fluorides,” op. c¢it., ORNL-1170, p. 82. i L3 diaphragm in the cell and, in part, to the lack of high accuracy in the analytical methods available limit the usefulness of the data obtained. Consequently, no further studies of electrolytic transference are now anticipated, Before these experi- ments can be improved, an electrical insulator capable of holding the salts without reaction and con- siderably better analytical methods than now exist will be needed. (An excellent review of such experiments has been given by Baimakov and Samusenko.(%)) The following experi- mental results from the recent study are, however, of interest, The number of electrons necessary to reduce one uranium(IV) ion to metal decreases from about four in the more dilute solutions to less than two 1in the more concentrated ones. The changes in the ion fraction of uranium in the two compartments of the cell are not altogether consistent. The anolyte concentration of uranium does not differ significantly from that found before electrolysis, whereas the change in the catholyte concentration is little more than the experimental error. On the other hand, the alkali metal concentration rises sharply in the catholyte and slightly in the anolyte. | : The apparent deviations from Faraday’s law may represent a case of stenolysis, i.e., electrode processes -occurring in the pores of the diaphragm separating the anolyte and catholyte. If all current passing through the cell did so by a stenolytic mechanism, two electrons would apparently suffice to reduce one uranium(IV) ion to metal, whereas complete absence of stenolysis would require four electrons per reduced uranium(IV). Tt is possible that the diaphragm used was (S)Y. V. Baimakov and S. P. Samusenko, “Determi- nation of Transference Numbers of Ions in Fused Salts,” Trans. Leningrad Ind. Inst., 3-26 (1938). ‘FOR PERIOD ENDING MARCH 10, 1952 not completely permeable touranium(IV) and that the deviations from Faraday’s law result froman increase of stenoly- sis with increasing uranium tetra- fluoride concentration. A photo- micrograph of the surface of the diaphragm, however, seemed to indicate a pore size sufficient to pass any of the 1ons probably present, The stenolytic behavior, coupled with the observed concentration changes, indicates that the quadrai- valent uranium is probably not present as Ut4., The small radius of this ion and its large charge indicate a high mobility. That this is not observed is evidence for the presence of some more complicated species. PREPARATION OF STANDARD FUEL SAMPLES G. J. Nessle J. E. Eorgan V. C. Love Jack Truitt C. J. Barton Materials Chemistry Division A total of 68 batches of fluoride fuels was prepared during the past quarter, including two large (3 to 4 kg) batches of pretreated and filtered material for experiments in dynamic corrosion in thermal convection loops. The materials were used to fill equip- ment furnished by other ANP groups as indicated in Table 25. TABLE 25 Disposition of Standard Fuel Samples NUMBER OF TYPE OF EXPERIMENT CONTAINERS FILLED Static corrosion tests 243 Cyc}dtron bombardment 26 Radiation'damage _ 15 Viscosity measurement 8 Density measurement 4 Loop test : ‘ 2 103 e bl o i i el i, o R ki e e, i, B e, R, b b Rl b bl ol PREPARATION OF PURE HYDROXIDES D. R. Cuneo D. E. Nicholson E. E. Ketchen L. G. Overholser Materials Chemistry Division Development of moderator-coolants has been limited during the past quarter to preparation of pure alkal: and alkaline earth hydroxides for corrosion studies. Barium, strontium, and sodium hydroxide have been prepared with acceptable purity, that is, con- taining less than 0.1, 0.06, and 0.1%, respectively, of their carbonates, Purification of the potassium hydroxide has not been so successful. Barium hydroxide has been prepared, in a quantity sufficient for use in dynamic corrosion experiments, by a procedure(%) that yields 1-1b batches of material containing less than 0.1 wt % barium carbonate. Strontium ~ hydroxide prepared by the same process contained 0.06%of strontium carbonate, and spectrographic analysis showed about 0.1% barium hydroxide. Strontium hydroxide may be dehydrated without difficulty but is not yet available in large quantity. Additional batches of sodium hy- droxide have been purified by removing sodium carbonate from a 50% aqueous solution of sodium hydroxide and dehydrating under vacuum at 425°C. - These batches have assayed 100.0% -sodium.hydroxide, and the sodium carbonate content has been less than 0.1% in'all_qases. Potassium carbonate can be removed from potassium hydroxide by dissolution of the latter compound in isopropyl alcohol, but it has not proved possible to remove the alcohol from potassium hydroxide without carbonization and contamination of the potassium hy- ~droxide. Potassium hydroxide prepared (6-)1.. G. Overholser, D. E. Nicholson, E. E. Ketchen, and D. R. Cuneo, *“*Preparation of Pure -Hydroxides,” op. cit., ORNL-1170, p. 84. 104 ANP PROJECT QUARTERLY PROGRESS REPOKRT by dehydration of an aqueous solution containing 5 wt % potassium hydroxide, which had been treated with barium hydroxide, still contained about 0.3% potassium carbonate, The potassium carbonate content of higher concen- tration potassium hydroxide solutiomns decreased, however. There was 0.1% potassium carbonate in the 38 wt % aqueous solution of potassium hy- droxide. Recent experimental determi- nations have shown that barium carbonate is appreciably soluble in potassium hydroxide. Preliminary studies of the feasibility of using calcium hydroxide for this carbonate removal are 1in progress. In addition, equipment is under construction for preparing potassium hydroxide by reaction of pure water with metallic potassium. COOLANT DEVELOPMENT J. P. Blakely R. C, Traber, Jr. L. M. Bratcher C. J. Barton Materials Chemistry Division Investigation of low-melting fluoride mixtures for possible appli- cation as coolants has been continued during the past quarter. Imn addition to possible value as coolants, any low-melting compositions discovered are of potential value as solvents for uranium tetrafluoride in the prepara- tion of fuels of low uranium content, Recent interest in fuel systems containing zirconium fluoride has led to a considerable program of study in phase equilibria in alkali fluoride- zirconium fluoride systems. Although several of these systems show promise, in no case are the data sufficient for construction of the phase diagram, In many of these mixtures well-defined thermal breaks occur well below the ligquidus temperature, It has not been definitely established whether these effects are the result of solad transitions or the presence of some 3 impurity such as ZrOF,, which is known " to be present in the only available zirconium fluoride. LiF-ZrF,. The system'LiF-ZrF4, which is of possible future interest, is under study in order to obtain a complete picture of the alkali fluoride- zirconium fluoride systems. Cooling curves have been run on nine mixtures ranging between 15 and 70 mole % zir- conium fluoride. There appears to be a eutectic at about 39 mole % zir- conium fluoride melting at 560 * 10°C. No congruently melting compounds were observed, but there 1s some evidence for an incongruently melting compound that is probably Li,ZrF . Breaks other than melting points or the eutectic halt were observed at 590, 515, and 470°C in some mixtures, This study will be continued in greater detail at some future date. ‘ NaF-ZrF,. The study of the system NaF-ZrF, is still in progress; mixtures between 10 and 50 mole % zirconium fluoride have been studied. It appears likely that there is a eutectic at approximately 35 mole % zirconium fluoride, melting at 595+ 10°C, Breaks other than melting points or eutectic halts were observed at approximately 735, 530, and 495°C with some mixtures. KF-ZrF,., Although cooling curves have been run on a rather large number of mixtures in the system KF-ZrF,, covering the range from 5 to 95 mole % zirconium fluoride, the melting point of some compositions has not been determined with certainty. The existence of a eutectic at 13 mole % ~zirconium fluoride, melting at 765 * 10°C, was established and a compound K,ZrF,, melting congruently at 910 * 10°C, has been demonstrated. Mixtures in the range 30 to 60 mole % zir- conium fluoride show well-defined breaks or halts at 415 to 435 and 485°C for some compositions. FOR PERIOD ENDING MARCH 10, 1952 RbF-ZrF,, The system RbF-ZrF, has been studied in the range 5 to 60 mole % zirconium fluoride and appears to be quite similar to the KF-ZrF, system., There 1s a eutectic at about 6 mole % zirconium fluoride, melting at 720 + 10°C, and the compound Rb,ZrF, melts congruently at 870 %+ 10°C, It appears likely that there is a eutectic between 30 and 40 mole % zirconium fluoride, melting at 575 + 10°C, but there are also well-marked breaks on the cooling curves for mixtures in this range at about 405°C. NaF-KF-Zr¥,. Cooling curves have been run on more than 50 mixtures in the ternary system NaF-KF-ZrF,, and it should be possible to draw contours for this system when the corresponding binary phase diagrams are completed. The lowest melting composition that has been found in this system contained 43 mole % zirconium fluoride, 5 mole % sodium fluoride, and 52 mole % potassium fluoride. This mixture exhibited no break in the cooling curve above 405°C and stirred down to this temperature. When heated, i1t started stirring at 405°C and showed no further break up to 500°C, at which temperature the experiments were terminated, Other similar systems showed considerably higher melting points., The 405°C melting point therefore seems to require. further verification. Com- positions with about the same amount of zirconium fluoride as the above- mentioned mixture but with more sodium fluoride and less potassium fluoride show well-marked breaks in the cooling curves at approximately 450°C, which may indicate a eutectic melting at this temperature; the composition of ‘this eutectic has not been established. NaF-RbF-ZrF,. The NaF-RbF-ZrF, system has been investigated about as thoroughly as the NaF-KF-ZrF, system. The lowest melting point recorded was 460°C for the mixture containing 52 105 ANP PROJECT QUARTERLY PROGRESS REPORT mole % zirconium fluoride, 11 mole % ‘rubidium fluoride, and 37 mole % sodium fluoride. Other breaks were observed in the cooling curve for this and other compositions at temperatures ranging from 410 to 440°C. These could be caused by a solid transition or by a eutectic of unknown compo- s1t1on. : 7 NaF-KF-LiF- ZrF . Onlyzafew mixtures dwere prepared 1n the four component system NaF KF LiF- ZrF, Increasing amounts of zirconium fluoride were added to the ternary NaF-KF-LiF eutectic (42 mole % KF, 46.5 mole % LiF, 11.5 mole % NaF), Wthh was con- sidered to be one component of a pseudoblnary mixture. The cooling curves that were run after each ad- dition of zirconium fluoride showed that the melting p01nt climbs rather rapldly, reaches a maximum of 810°C at 20 mole % zirconium fluoride, drops sharply to 450°C at 30 mole % zir- ¢onium fluoride, and then rises again. . PREPARATION OF PURE FLUORIDES CG. J. Nessle ~ F. F. Blankenship “J. E. Eorgan W. R. Grimes : ,Matérials Chemistry Division - The lack of successful operation of the thermal loops with fluoride fuel material as presently supplied and the definite plans for an ARE in the relatively near future have emphasized the need for research on equipment for preparation of pure, homogeneous, fluoride solutions and handling of the llqulds so as to keep them free from contamination until they are used. The mass-transfer phenomenon, which tfansfers metal from hot to cold regions of the thermal loops, is still not well understood. However, all the plugs so far observed have contained considerable quantities of oxide sludge admixed with the metal particles. It has not been established that the oxides are primarily responsible for 106 corrosion, but freedom of the system from oxides would certainly help to clarify the picture. In additiomn, it is essential for reactor operation that the fuel be free from sludge formation. The raw materlals for fuel manu - facture are, in general, quite hygro- scopic; 1t 1is difficult to prevent some pickup of water during the un- avoidable handling of the powders. Furthermore, any water picked up by the fluorides will hydrolyze uranium tetrafluoride to insoluble uranium dioxide when the temperature is raised. Uranium tetrafluoride also oxidizes readily at elevated temperatures to yield volatile uranium hexafluoride and soluble and corrosive U0,F, if any oxXygen 1s present. It seems evident therefore that improvement of the corrosion picture would result if satisfactory means existed for pre- paring and handling fluorides so that oxidation or hydrolization could not occur. (onsequently, equipment de- 31gned to prevent these reactions during fuel preparation and transfer are being developed Fuel Preparatlon Eoulpment The fuel preparat1on equipment w111 be a reactor in which all lines in contact with the molten fluor1des are of nickel so that they can readlly be reduced with hydrogen to remove oxide films and scale. The reactor will be charged with dry, pure fluorides, sealed by the gasketed flanges to the gas manifold, and evacuated to remove as much adsorbed oxygen and water as possible at low temperatures. The mixture will then be melted and treated at 500 to 600°C with dry " oxygen-free hydrogen to reduce any UO,F, or UF, formed by oxidation or present in the uranium tetrafluoride used. The water will react with the uranium tetrafluoride at this tempera- ture and yield hydrogen fluoride, which will be swept and pumped out, and o uranium dioxide, which will be sus- pended in the melt, By treatment of the molten liquid at 500 to 600°C with dry hydrogen fluoride, the suspended oxides can be reconverted to fluorides. The excess hydrogen fluoride will be pumped off, the liquid swept with pure helium, and the pure liguid transferred through a nickel filter to a clean "receiver by use of helium pressure. The pure material will be used for capsule and dynamic loop corrosion studies, as well as for all physical property measurements, It is antici- pated that a number of preparations will be made under varying conditions of temperature, exposure time, liquid composition, etc, to define the optimum FOR PERIOD ENDING MARCH 10, 1952 conditions for operation before the material is used for systematic testing, Fuel Handling Equipment. TIf the fuel preparation is to be conducted under such stringent conditions, 1t will obviously be undesirable to have it contaminated before testing. Conse- quently, a capsule-fillingrig is being developed in which corrosion capsules can be treated with hydrogen, hydrogen fluoride, fluorine, or any desired reagent, filled with an inert gas, connected to a source of pure fuel, filled, sealed, and welded without exposure to alr or to an uncontrolled atmosphere. 107 e Rl L e TS Y T Y TR I I Y L A g 3 11. CORROSION RESEARCH W. R. Grimes, Materials Chemistry Division W. D. Manly, Metallurgy Division H. W. Savage, ANP Division The major emphasis in the corrosion testing program returned, during the past quarter, to the procblem of con- tainer materials for the fused fluoride mixtures. This effort was extended considerably followingthe difficulties encountered during circulation of the fluorides by forced convection 1in closed metal systems. Empirical testing of corrosion of fluoride mixtures by the static capsule tech- nique previously described has been continued, and dynamic testing of fluoride corrosion 1s being intensively investigated by several technigques, for example, thermal convection loops, seesaw tests, and standpipes. In addition, a large program of a more fundamental nature dealing with possible reactions of metals with the fluoride melts has been initiated. Static corrosion testing under generally isothermal conditions has been continued with several high- temperature liquids and structural metals by using slight variations of the capsule technique previously described. Other materials have been tested in the molten fluorides to find possible bearing materials and valve seat materials that would not be seriously attacked by the fluoride mixture, Stellite, which 1s a good valve seat material, is apparently unattacked in the fluoride mixture "~ when tested statically.' Additions of zirconium and magnesium as possible .corrosion inhibitors were made to the fluorides prior to running corrosion tests. It was found that magnesium increased the corrosion attack, whereas additions of zirconium de- creased the attack of the fluorides on 300-series stainless steels. Static corrosion tests have also been run on low melting point alloys for secondary coolants, such as sodium-lead alloys and the hydroxides. Low melting point alloys containing mixtures of lead, tin, cadmium, and bismuth have been tested to find suitable container material. It has been found that lead-cadmium and bismuth-cadmium alloys can be stati- cally contained in type-317 stainless steel. Corrosion testing in sodium- lead alloyshas shown that the addition 0of sodium to lead lessens the attack on stainless steels and Inconel. Additions of zirconium, sodium, sodium carbonate, sodium cyanide, and sodium hydride were made to sodium hydroxide prior to running statlc corrosion tests, and the effects of these additions have been tabulated. Al though the static capsule tech- nigque has proved satisfactory for routine screening of materials, 1t lacks sufficient sensitivity to anticipate the results of the more difficult, large-scale, dynamic tests. When routine dynamic tests that can be conducted in large numbers are developed, the static capsule method will be discarded. Thermal convection loops, which have hitherto been used almost exclusively for dynamic cor- rosion tests, have now been supple- mented by a seesaw capsule technique and standpipe tests. Dynamic corrosion of the fluorides and hydroxides has been studied with these devices. Dynamic corrosion testing reveals the corrosion behavior of different alloyswith the fluorides, the effect of inhibitor additions, and provides a means for determining the 109 Sl ¢ By b e R B A i e s s effect on the reaction of various atmospheres, such as argon, hydrogen, oxygen, and vacuum, in contact with the bath., One dynamic experiment has shown that potassium hydroxide can be successfully contained in Inconel without corrosion or mass transfer, For this experiment, the loop was operated under a hydrogen atmosphere for 135 hr with a hot-leg temperature of 715°C and a cold-leg temperature of - 440°C., This effort on container materials for molten hydroxides 1is supplemental to a fundamental approach to the corrosion problem through reactionsof hydroxides with structural metals at high temperatures. STATIC CORROSIONBY FLUORIDES F. Kertesz Materials Chemistry Division D. C. Vreeland Metallurgy Division During the past quarter testing techniques have been improved and corrosion ‘tests of a wide variety of materials in various fluorides have been completed. Even so, weight change and penetration data from the recent tests with dilute fuels do not justify a conclusion as to which metal among Inconel and stainless steels types-316, -321, -347 and -316 ELC is superior. The latest static tests of Inconel and 300-series stainless steels in carefully prepared ~fluoride mixtures support the earlier conclusion that there is little corrosive reaction at 1500°F. A series of screening tests in the untreated fluoride fuel (NaF-KF-LiF-UE,) for 100 hr at 1500°F showed that the refractory metals and Stellite were unaffected and that the stainless steels were attacked to a depth no greater than 1 mil. Analyses of the fluorides after corrosion tests have 110 . ANP PROJECT QUARTERLY PROGRESS REPORT indicated that no major reaction capable of changing the physical properties of the fluoride melt had occurred. The effects of several additives to the fluoride have been determined. Zirconium and uranium dioxide apparently decrease the cor- rosiveness of the fluoride, whereas magnesium and uranium oxide increase the corrosion. - Effect of Pretreatment of Fuel (H. J. Buttram, N. V. Smith, R. E. Meadows, C. R. Croft, Materials Chemistry Division). The original tests with fuels of high uranium content (120 to 140 1b of uranium per ft®) showed extensive corrosion (4 to 8 mils) if the materials, as received, were melted under inert atmospheres and used in the tests. After treatment with stainless steel and Inconel, however, these fuels showed con- siderably reduced corrosion (0.5 to 2 mils) in 100 hr at 800°C. The more dilute fuels (5 to 10 1b of uranium per ft®) containing lithium fluoride and beryllium fluoride, which are under study at present, are not appreciably improved by the treatment procedure; they show 1 to 3 mils of penetration whether treated or not treated. Corrosion of Structural Metals (D. C. Vreeland, E. E. Hoffman, R. B. Day, L. D. Dyer, Metallurgy Division). In addition to a series of screening tests of various materials in an untreated fluoride fuel for 100 hr at 1500°F, some data have been ob- tained on the corrosion of the treated fuel on molybdenum, cold-worked Inconel, and stainless steel. There is no discernable temperature effect between 850 and 1000°C on the cor- rosion of Inconel and several stainless steels, The screening tests of various materials in fluoride fuel (43.5 mole % KF, 44.5 mole % LiF, 10.9 mole % NaF, T W y and 1.1 mole % UF4)-have been com- pleted. These tests were run with dehydrated, untreated, fluoride mixtures for 100 hr at 816°C under vacuum. Molybdenum, columbium, Monel, and Stellite 25 were apparently un- attacked. All the 300-series stain- less steels tested (types-310, -317, -321, -347, and -304 ELC) were attacked to a depth of 1 mil or less, except type-304 ELC, which was attacked to a depth of 2 mils. Stainless steels of TABLE FOR PERIOD ENDING MARCH 10, 1952 the 400 series were not attackedover 1 mil. Z-nickel, which was the most severely attacked material of those tested, was affected to a depth of 5 mils. The attack on Inconel was 3 mils. Details of these tests are presented in Table 26. Molybdenum, Timken Alloy 6 (16% Cr, 26% Ni, 6% Mo, balance Fe), Timken Alloy 3 (16% Cr, 13% Ni, 3% Mo, balance Fe), and a 74% nickel-26% 26 Static Corrosion of Various Materials in the Untreated Fluoride Fuel (NaF-KF-LiF-UF,)* in 100 hr at 1500°F DEPTH OF METAL MATERTAL AFFECTED (mils) METALLOGRAPHIC NOTES Globe iron 2 Surface of specimen very rough Molybdenum ' Surface of specimen roughened Type-310 stainless steel 1 Subsurface voids Type-317 stainless steel 1 Intergranular penetration Type-321 stainless steel 0 to 1/2 Intergranular penetration Type-347 stainless steel 1 Subsurface voids Type-430 stainless steel 1 Slight intergranular penetration and decarburization Type-446 stainless steel 1 Subsurface voids Hastelloy B 0 to 1/4 Subsurface voids Hastelloy C 2 Subsurface voids .Ihébfiéixxfii 7 1 Subsurface voids Stellite 25 (L-605) No visible attack Z-Nickel Voids along grain boundaries Tantalum 1 Surface of specimen roughened Columbium Surface of specimen roughened Monel No apparent attack Nichrome V | 3 Subsurface voids, some following grain : , - R boundaries Inconel 3 Subsurface voids, some following grain o e boundaries Type-~304 ELC staifiieés.steei 2 Subsurface voids, some following grain : boundaries, attack somewhat irregular *Composition in mole %: NaF, 43.5; KF, 44.5; LiF, 10.9; UF,, 1.1 111 T P T e ANP PROJECT QUARTERLY PROGRESS REPORT molybdenum alloy have been tested in a pretreated fluoride fuel {(compo- sition in mole %: NaF, 46.5; KF, 26.0; UF,, 27.5) for 100 hr at 816°C. Molybdenum was not attacked, and the Timken alloys and molybdenum-nickel alloy were attacked only slightly (1 mil or less). Static tests have also been run on specimens of as-received and ap- proximately 20% cold-worked Inconel and 20% cold-worked types-316 and -310 stainless steel in the above-mentioned fluoride fuel at 816°C for 100 hours. This amount of cold working appeared to have no significant effect on the corrosion properties of these materials under the testing conditions employed. Tests have also been run at 850, 900, and 1000°C for 100 hours. The extent of attack varied from 2 to 5 mils, ‘but no definite correlation between temperature of test and amount of attack could be established. Ap- parently these materials are in- sensitive to test temperatures within this range. ' Corrosion by Fluorides with Various Additives. A short series of tests on Inconel and type-309 stainless steel have been run using fluoride fuel (NaF-KF-LiF-UF,) with zirconium and magnesium added to the tests in the form of turnings. The results are summarized in Table 27, The additions of magnesium ap- parently increased the corrosion of both type-309 stainless steel and Inconel. Additions of zirconium appeared to lessen the attack of the fluoride on these metals. In the tests with the zirconium additions 1t was noted that built-up surface layers of 1 1/2 to 2 mils and 1/2 mil in thickness were present on the Inconel TABLE 27 Static Corrosion of Inconel and Type~-309 Stainless Steel by the Fluoride Fuel (NaF-KF-LiF-UF,)* with MagneSium and Zirconium Additivesin 100 hr at 1500°F ' DEPTH OF METAL METAL BATH ADDITIVE AFFECTED (mils) METALLOGRAPHIC NOTES Inconel None 2 ' Subsurface voids Inconel 2% magnesium 5 Large voids Inconel ' 2% zirconium 1 Surface layer 1 1/2 to 2 mils thick, 1/2 to 1 mil of at- t ack beneath the surface Layer Type-309 stainiesé steel None 2 Subsurface voids Type-309 stainless steel 2% magnesium 5 Subsurface voids, irregular attack Type-309 stainleés.steel 2% zirconium No visible attack, specimen gad surface layer 1/2 mil eep *Composition in mole %: NaK, 10.9; KF, 43.5%; LiF, 44.5; UF,, 1.1 112 and type-309 stainless steel, re- spectively. A photomicrograph of the Inconel corrosion specimen is shown in Fig, 32. An attempt is being made to identify these layers by spectro- graphic methods. Preliminary experiments have in- dicated that addition of uranium dioxide to the fuels does not increase the corrosion i1n static experiments. Additionof uranium oxide has, however, greatly increased the corrosion ob- served even in the static tests., It is likely that this is true of any hexavalent uranium compound including UO,F,, which is the expected product of oxidation of uranium tetrafluoride. This is discussed in a separate section below. T T T . e . b S @ aw - FOR PERIOD ENDING MARCH 10, 1952 Melting Point of Fluorides After Corrosion Tests (J. M. Didlake, G. J. Nessle, C. J. Barton, Materials Chemistry Division). Samples from 26 static corrosion capsules and dynamic loop experiments were re- covered to establish whether any changes in the melting point of the material had occurred as a consequence of the corrosion reactions. As in- dicated by the data in Table 28, the melting point of these samples is not affected by the corrosion. In some cases removal of the liquid from the capsule left some very high melting point materials behind. Although analyses of these materials shows them to be wet with the molten fluoride, they seemed to consist of oxides of uranium and the structural metals and re U Fig. 32. Corrosion of Inconel in a Fluoride Fuel [(NaF-KF-UF4i t 2% ZrI for gfl? 100 hr at 816°C. 200X. _ 113 P £ £ b f2 ?-»3 fhge ER e oy E Rt L : b i il S o i iz e i A s ik G . e, A B, v . B R 1 30 M A, . i 4 ! 4 1 I8 ? ANP PROJECT QUARTERLY PROGRESS REPORT of metallic particles from the con- tainer walls. melt had occurred. 'STATIC CORROSION BY SODIUM HYDROXIDE D. C. Vreeland . R. D. Day E. E. Hoffman L. D. Dyer Metallurgy Division It was clear that no major reaction capable of changing the physical properties of the fluoride The static corrosion of sodium hydroxide has been determined with a variety of specimens and the addition of presumed inhibitors. None of a number of special alloys that were tested was severely corroded after 100 hr at 1500°F, although the attack was erratic. The corrosion of struc- tural metals in sodium hydroxide with various additives was in no case reduced to insignificance., Of several refractory materials exposed to the test conditions only vitrified be- ryllium oxide survived without crum-. bling, and it lost considerable ~ weight. Corrosion of Special AlIdys. A few tests of special alloys have been TABLE 28 Melting Points of Samples of Fluoride from Corrosion Tests MELTING POINT (°C) NO. OF SAMPLES FINAL* ORIGINAL " .- CAPSULE MATERIAL Type-316 stainless steel 9 527 530 Inconel ) 1 526 530 Type-316 stainless steel 10 450 452 Inconel 1 450 452 Type-316 ELC stainless steel 2 450 452 Type-321 stainless steel 1 456 452 Type-347 stainless steel 1 450 452 Type-316 stainless steel 2 458 460 Type-316 stainless steel 4 338 345 LOOP SECTION MATERIAL Type-316 stainless steel (hot) 1 455 452 Type-316 stainless steel (cold) 1 455 452 Type-410 stainless steel (bottom) 1 453 452 Type-410 stainless steel (hot) 1 453 452 Type-410 stainless steel (cold) 1 453 452 *Mean value if more than one 114 sample was used. e £ -~ 7 e b Vo by L‘ o Y l.‘ i Gyl 3 C,&_ 7 run in sodium hydroxide for 100 hr at 1500°F. A 26% molybdenum-74% nickel alloy showed light attack to a depth of 2 mils, whereas a 25% nickel-75% iron alloy was attacked to a depth of only 1 mil. Other alloys composed of equal parts of iron and nickel; iron, nickel, and cobalt; and nickel and cobalt have also been tested. Al though one of these alloys was severely corroded, in general, the attack was quite erratic with deep local penetrations. Since the capsules and specimens were machined from as- cast bars without prior working or annealing, it is believed that the cast structure of these materials may have caused this type of attack. It FOR PERIOD ENDING MARCH 10, 1952 is planned to test materials of simil ar compositions that have been hot-worked and annealed so as to have a more uniform structure. Corrosion by Sodium Hydroxide with Various Additives. A series of tests were run with sodium hydroxide and various addition agents, 1including zirconium, sodium, sodium carbonate, sodium cyanide, and sodium hydride. The addition of approximately 50% of sodium cyanide appeared to lessen attack somewhat on type-316 stainless steel and Inconel. None of the agents tested appeared to reduce corrosion to an insignificant level., Details of these tests are presented in Table 29. TABLE 29 Static Corrosion of Structural Metals by Sodium Hydroxide with Various Additives in 100 hr at 15060°F WEIGHT - CHANGE DEPTH OF METAL MATERTAL ADDITIVE (mg/in.z) AFFECTED (mils)| METALLOGRAPHIC NOTES Type-304 stainless steel 2 : 5-mil oxide layer Type-304 stainless steel | 10% Zr 2 1/2 6-mil oxide layer Type-304 stainless steel | 30% Na,COj 3 6-mil oxide layer A-Nickel ' 1% Zr No attack Inconel 12% Zr 3 11-mil oxide layer Type-316 stainless steel | 48% NaCN -286.6 2 Intergranular attack Inconel 49% NaCN -207.3 7 Intergranular attack Inconel T.4% Na Complete Specimen converted penetration completely to corrosion ‘ . product Inconel 50% Na Complete Specimen converted penetration completely to corrosion . product A-Nickel 14% Na +51.1 No corresion product, ‘ but edge of specimen rough A-Nickel 54% Na -2.5 No attack Inconel 31% NaH Complete Specimen converted penetration completely to corrosion product Type-304 stainless steel | 31% NaH -24.9 4 1/2 Oxide lavyer 115 ol ik G, B e v Lead-sodium alloys ANP PROJECT QUARTERLY PROGRESS REPORT Corrosion of Refractory Materials, A number of refractory materials were tested in sodium hydroxide in capsules of nickel at 800°C for 100 hours. Carbides of tungsten, tantalum, columbium, titanium, and silicon and zirconium nitride, all in the hot- pressed condition, disintegrated during the test. Vitrified beryllium oxide lost considerable weight but survived without crumbling. STATIC CORROSIONBY LIQUID METALS D. C. Vreeland E. E. Héffmén R. B. Day L. D. Dyer Metallurgy Division Liquid metal corrosion studies during the past quarter have been limited to the testing of various low melting point alloys of potential use as a secondary coolant and a number of lead-sodium alloys. Inconel ~and types-310 and -317 stainless steel were tested in the low melting. alloys for 100 hr at 1500° F. The type-317 stainless steel was Gnattacked by the 82% lead-18% columbium alloy and only slightly attacked by the 60% bismuth-40% cadmium alloy, whereas Inconel and the type-310 stainless steel were generally subject to somewhat greater attack. containing as little as 5% sodium have been suc- cessfully contained in Inconel and a number of 300- and 400-series stain- less steels with no greater than 1/2-mil attack after 100 hr at 1500°F. Corrosion by Low Melting Point Alloys. A low melting point alloy is being considered as a coolant in a secondary heat exchanger, so several eutectic compositions of low melting point alloys have been employed as corroding mediums in static corrosion tests with types-310 and -317 stainless steel and Inconel. The compositions 116 and melting points of the eutectic alloys used are listed in Table 30. TABLE 30 Melting Points of Various Alloys COMPOSITION MELTING POINT BY WEIGHT (°c) 44% Pb-56% Bi 124 43% Sn-57% Bi 138.5 60% Bi-40% Cd 144 . 68% Sa-32% Cd 176 38% Pb-62% Sn 183 82% Pb-18% Cd 248 In these experiments, which were run for 100 hr at 816°C under vacuum, all the low melting point alloys conta1n1ng tin (except Inconel in’ w68% tin232% cadmium> alloy) proved to. be extremely vigorous in their attack on the metals tested. It is believed that the type of attack in these mediums can be classified either as intergranular, as in Fig. 33 that shows the characteristic penetration along the grain boundaries sometimes accompanied by voids, or as an alloying type of attack in which the attacked surface actually alloys with the molten coolant being tested, as shown in Fig. 34. Inconel did not show much promise in the tests. Type-317 stainless steel was unattacked by the 82% lead-18% cadmium alloy and only slightly attacked by the60% bismuth-40% cadmium alloy. The results of these tests are summarized in Table 31, It is planned to test other structural materials in the low melting point alloys that proved least corrosive in the initial tests. W il » i FOR PERIOD ENDING MARCH 10, 1952 [ UNCLASSIFIED 40 5 ol 4t Lo ‘;jl ,\‘;t N F Sea e BYGS Fig. 33. Intergranular Attack of Type-310 Stainless Steel Tested in *44% Lead-56% Bismuth Alloy for 100 hr at 1500°F. 200X, UNCLASSIFIED _ Y-5328 Fig. 34. Alloying Attack of Inconel Tested in 43% Tin-57% Bismuth Alloy for 100 hr at 1500°F. 100X. i 2 (T @ - 117 & 5 PO £ 74 Sl des i, i o SR s it R D s o © 'ANP‘PROJECT QUARTERLY PROGRESS REPORT TABLE 31 Corr051cn of Types-310 and -317 Stainless Steel and Inconel by Varxouscfi Low Melting Point Alloys in 100 hr at 1500 F A series of static corrosion tests of various materials in three different sodium-lead mixtures (80%. sodium-20% 118 s ,('f“ T R e Ly N Vv b ALLOY . COMPOSITION DEPTH OF METAL BY WEIGHT MATERIAL AFFECTED (mils) METALLOGRAPHIC NOTES .. 43%75n-57% Bi Type-310 stainless steel 10 Many large voids 43% Sn-57% Bi Type-317 stainless steel 11 Irregular attack, many voids . 43% Sn-5T% Bi Inconel 8 5-mils of a uniform layer on surface with an underlying layer of voids, 3-mils in | thickness ) - 60% Bi-40% Cd Type-310 stainless steel 15 Attack very irregular, varying s ' 0 to 15 mils, both grains and grain boundaries attacked in some cases - 60% Bi-40% Cd Type-317 stainless steel 2 Intergranular penetration and ' voids in a few areas ) -60% Bi-40% Cd Inconel Complete pene- Voids throughout entire speci- tration of men, tube attacked to a depth _ specimens of 20 mils ' 38% Pb-62% Sn Type-310 stainless steel | Complete pene- Erratic attack S : ‘ tration of . ; B specimens 38% Pb-62% Sn Type-317 stainless steel Complete pene- Erratic attack, only 2 mils . . _ tration of affected 1n places - 7 \, specimens . 38% Pb-62% Sn Inconel 8 Erratic attack 68% Sn-32% Cd Type-310 stainless steel | Complete pene- Erratic attack, tube failed by : tration of penetration of wall . specimen 68% Sn-32% Cd Type-317 stainless steel Complete pene- Erratic attack : tration of specimen 68% Sn-32% Cd | Inconel 2 Uniform attack . 82% Pb-18% Cd Type-310 stainless steel 2 Intergranular attack 82% Pb-18% Cd Type-317 stainless steel 0 No attack on specimen or tube 82% Pb-18% Cd Inconel 10 Intergranular attack 44% Pb-56% Bi Type-310 stainless steel 6 Intergranular attack varying from 2 to 6 mils 44% Pb~56% Bi Type-317 stainless steel 15 Intergranular attack : 44% Pb-56% Bi Inconel 4 Intergranular attack varying . ' from 1 to 4 mils Corrosion by Sodium-Lead Alloy. lead; 50% sodium-50% lead; and 5% sodium-95% lead) has been completed. The tests were run at 816°C for 100 hr under vacuum. Only in the case of 3 e Inconel in the 5% sodium—95% lead "mixture was the depth of attack greater than 1/2 mil; Inconel was affected to a maximum depth of 3 mils, The various test combinations included Inconel and types-317 and -346 stain- less steel in all the alloys, type-316 stainless steel in the 50-50 alloy, and type-304 stainless steel in the 5% sodium-95% lead alloy. Except for the aforementioned case of Inconel, the 5% sodium-95% lead alloys show no attack whatsoever. All specimens were ductile on bending 180 degrees. It was noted that some of the stainless steel specimens that had been tested in the 80% sodium-20% lead mixture had apparently become carbu- rized. observed before in testing similar materials in sodium.(!’ No carbu- rization of any of the stainless steels was noted with the 50% sodium-50% lead or 5% sodium-95% lead mixtures, which indicates that occurrence of carburization is enhanced by the presence of higher percentages of sodium in the bath metal. Future plans include running static cor- rosion tests with lower percentages of sodium in the bath metal. FACILITIES_FOR 'DYNAMIC CORROSION TESTING Static corrosion tests are satis- factory'for the routine screening of corrosion resistant materials, but they lack the sensitivity to antici- pate the results of dynamic testing. The mass transfer phenomenon and the deep penetration in the hot zone that are characteristic of all the dynamlc thermal convection loops are not observable in static isothermal capsule studies. These latter ex- periments are, however, easy to (I)F. N. Lyon (ed.), Ligquid Metals Handbook, NAVEXOS P-733, p. 90, June 1, 1950. OF% 107 Gt Cf ,.,Ld i This phenomenon has been FOR PERIOD ENDING MARCH 10, 1952 perform in large numbers and under conditions of better reproducibility. Dynamic corrosion testing 1n the thermal convection loops in which the bulk of the testing has been per- formed has now been supplemented by a number of techniques that promise to duplicate the phenomena shown in the large loops with much smaller and more economical equipment. One of these technigues is the standpipe test 1n which controlled thermal gradients may be introduced. The most promising, however, 1s the seesaw technique with which fluid motion as well as thermal gradients are readily attained. Thermal Convection Loops. The large thermal convection loops, which have been in operation for almost two vears, have been described in previous reports.(2'3'%) The loops are usually fabricated of 1/2-in. pipe and are shaped roughly in the form of a rectangle 1 ft wide and 3 ft high. When the bottom and an adjacent side of the rectangle are heated, convection forces in the contained fluid establish a flow that attains a velocity of up to 8 ft/min de- pendlngupon the temperature difference across the hot and ¢old sides of the rectangle. & Recently, smaller convection loops have been used in these ex- periments. These loops afford the simplest and most direct means of studying mass transfer, since the (2)E. M. Lees, J. L. Gregg, and R. B. Day, “PDynamic Corrosion Tests in Thermal-Convection Loops,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1951, ORNL-1154, p. 124. (B)E. M. Lees, *“Thermal-Convection Loops,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending June 10, 1951, ANP-65, p. 150. E. M. Lees, “Thermal Convection Loops,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending Marchk [0, 1951, ANP-60, p. 203. 119 B B R i, B W & ki L ANP PROJECT QUARTERLY PROGRESS REPORT characteristics are such as to ac- centuate the effects leading to mass transfer. The loops can be adapted to allow the exclusion of oxygen, water vapor - and to some extent metallic oxides — from the system. The loops are formed of 1/2-in.-ID tubing with a nominal wall thickness of 0.035 in. into the form of a rectangle, 7 by 17 in., with a loading tube extending above the hot leg of the loop. A dehydration pot is connected with the loop by standard 1/4-in. tubing. Glass tubing is attached at the tops of the loading legs of both the loop and the dehydration pot by means of Kovar seals for connect1ng ‘vacuum and hydrogen lines. The furnaces for heating the loops _are constructed from heating coils “‘mounted in split, cylindrical ceramic forms fitted around the sections of ;the_loop._ Seesaw Corrosion Tests. The seesaw tests require a sealed metal tube approximately one-third full of molten “material. “'a furnace capableof heating to 1700°F ;and up to 18 sealed corr081on capsules The capsules are placed in To create a temperature differentr1al, the capsules may be placed so that one end emerges from the furnace. The tubes are gently rocked on a central fulcrum similar to a seesaw so that the molten medium flows by grav1ty to both ends of each tube The apparatus is shown in Fig. 35. After several thousand cYcles,'the tubes are sectioned for examination. It is anticipated that in this manner the mass transfer and enhanced corrosion phenomena can be obtained and a variety of condltlons for 1mprovement studled D1fferent1a1 Temperature Tests. A | type of test frequently referred to as the standpipe test involves capsules similar to those used in the seesaw tests, except that they are almost completely filled with the fluid and 120 with desired pretreatment, are held stationary in a vertical position. This is not a true dynamic test because the fluid is essentially immobile. With this type of test, however, the temperature gradient may be controlled so that it is hottest at the top, bottom, or center. Normal convection provides the only means of circulation. After a suitable time, the tubes are sectioned and examined. Another design with which a large thermal gradient may be obtained employs a U-shaped tube. In this apparatus there is a large temperature differential between the two legs, but there is the smallest possible AT from top to bottom of each leg (to minimize local convection currents). The object of such tests with this apparatus is to determine the contri- bution of temperature difference alone to corrosion and/or metal transport. A third type of apparatus will permit one portion of a metal specimen to be water-cooled internally. One end of the metal will be immersed in a crucible containing stirred or un- stirred molten test fuel at aspecified temperature. This equipment will be contained in a 5-1 pot furnace. Rotating Dynamic Corrosion Tests. Another tester incorporates means for preparing the li1iquid to be used and for inserting a rotating specimen that acts as a small pump for agitating the liquid. Basic components of the apparatus are a lined pot for the preparation of the eutectic and a lined receiver with a screen between the pot and receiver. The sequence of operations includes preparing the eutectic from components in the pot, | and then up-ending the device to allow the molten fluoride to run into the receiver. The pots are then separated in a dry box and the receiver fastened onto the stuffing-box assembly that includes a weighed andmeasured rotating s By T b L ¢t Fig. 35. Seesaw Apparatus for Dynamic Corrosion Tests. ‘0T HOUVIN ONIANI d0oldidd 404 cS6T B2l Lo caig D L g Do Rgsio i cBb g - e e R phenomena, ‘materials. ANP PBOJECT QUARTERLY PROGRESS REPORT test specimen. After a test at any desired temperature the specimen can be completely examined and the eutectic analyzed. ' Forced Convection Loeps. Forced circulation loops are not routinely used for corrosion research. Several pump loops, heat exchanger loops, etc. are operating with both sodium and the fluorides, but these are primarily for the development of the mechanical equipment (sec. 3). The flow rates attainable in thermal convection loops have been adequate to expose corrosion When corrosion tests with higher flow rates are necessary, they will be performed in forced convection loops. ‘-DYNAMIC'CORROSION BY FLUORIDES A program for‘dynamic testing of various fluoride mixturesin containing materials is being conducted in the large (1/2 in. ID) thermal convection loops. Duringthis period the emphasis ‘has been on screening tests to deter- mine the more feasible container At present none of the stainless steels or Inconel will satisfactorily contain the fluoride mixtures under dynamic conditions at 1500°F, In general, the uranium ‘bearlng fluoride (NaF-KF-LiF- UF,) ls more corrosive ‘than the eutectlc mixture NaF KF-LiF. Furthermore, with the above fluoride fuel mixture, no _ 400-series stalnless steel loop has ~operated more than 15 hr before plugging, whereas no 300-series stain- ‘less steel loop has operated for more ~than 150 hr before plugging. ‘loops, on the other. hand, generally ‘run to scheduled termination at 500 hr ‘Inconel without plugging, but the loop material is usually attackedto 10 mils or more. Preliminary results of fluoride corrosion in seesaw tests have sub- stantiated previous results, which 122 indicated that uranium-free fluorides were less corrosive than similar systems containing uranium tetra- fluoride. In these seesaw tests, as ~well as similar standpipe tests, the formation of uranium dioxide crystals in the cold zone was observed. Corrosion by Fluorides in Thermal Convection Loops (G. M. Adamson, K. W. Reber, Metallurgy Division). It is of particular interest that no system or mechanical device other than convection loops has ceased operatlon because of plugging, parent metal pipe failure, or liquid leakage (except at poorly designed welds). It is equally true, however, that some of the other systems would not have endured had the walls and parts been thin enough to permit full penetration of 1ntergranular corrosion., All loops, with one exception, have been of the same design and have been fabricated from 1/2-in,-ips, sched- ule-40 pipe. Table 32 is a summary of the cor- rosion data from all loops, including those still in operation, in which tests have been made with fluorides. It is evident that considerable cor- rosion occurred in all loops examined so far and that the presence of uranium in the coolant accelerates plugging. The two stainless steel loops that contain fluoride mixtures without uranium show no signs of plugging after 350 hr of operation, whereas all stainless steel loops containing the uranium-bearing fluorides have plugged. The 300-series stainless steels with uranium-bearing fluorides usually plugin from 100 to 200 hr in convective systems, and the life appears to be "on the order of 10 hr/mil (0.001 in.) of thickness. The 400-series stainless steels usually plug in 50 hr or less. Inconel loops, on the other hand, have not plugged in any of the 500- and w 4 B w ¥ ak -'.“‘ - . N w» * - .y . . oy 1 4o o “w o - *® . " . -y (] 1 @ Corrosion Data from Inconel and Stainless Steel Thermal Convection Loops Operated with Various Fluorides HOT MENIMUM MINIMUM . TIME OF REASON LEG TEMPERATURE TEMPERATURE METALLOGRAPHIC. NOTES . CIRCULATION FOR TEMPERATURE (starr) (terminatien) BAP! LOOP NO. MATEREAL COOLANT (hr) TERMINATION (°F) (°F) X-RAY STUDIES HOT LEG COLD LEG 78 Inconel Flinsk 12¢% 1000 Scheduled 1500 1220 1220 Intermittent layer of pits, § mils, {No attack, ne mass transfer ) : intergranular attack, 10 mils 111 Type-316 stainless steel Flinak 12 173 Leak 1525 1210 1150 Very cvarse-grained metal, inter- Very coarse-grained metal, granular attack 10 mils, some continuous layer, 1 mil, ) ‘ reduction of wall no attack Pt . 8.7 A 112 Type-316 stainless steel] Fulinak 14¢9) 82 Plug 1500 1210 960 Second phase in hot leg, possibly intergranular attack up to 8 mils Intermittent mass transfer Q‘T"i ~ in cold leg i layer $ 113 Type-316 stainless steelf Fulinak 14 ' 123 Plug 1600¢ D 1250 ? Intergranular attack 8 mils, some |Thin mass transfer layer . . pitting 210 Inconel Fulinak 14 - 500 Scheduled 15G0 1245 1245 Layer of pits 10 mils, with maxi- Slight surface roughening v mum of 13 mils; layer on surface Pl _ : 2 mils €_§,35 211 Enconel ' Ffilinak 14 506¢ Scheduled 1500 1250 1245 $Second phase in hot horizontal Layer of pits, 4 to 8 mils, vough |Thin continucus layer on : - end cold legs surface surface, surface rough 212 Inconel Fulinak 14 : 38 Leak 1500 1230 1230 Intermittent layer of pits, 5 mils 40 Type-410 stainless steel|Fulinak 14 9 Plug 1564 1246 1115 Second phase in hot horizontal and cold legs, possibly in hot leg 43 Type-410 stainless steel| Fulinak 14 12 Plug 1500 1230 1166 Second phase in both parts of hot leg 48 Type-43¢ stainless steelf Fulinak 14 8 Plug 1500 1250 ? Second phase found in all sectioms 49 Type-430 stainless steel] Fulinak 14 9 Leak 1500 1250 1150 275 Type-347 stainless steel| Fulinak 14 39 Leak 1500 1250 1100 Second phase in hot horizontal leg {Intergranular attack, 8 te 13 mils,|Layer of 1 mil with nonme- grains 50 times larger than in cold} tallic particles on and in leg it 251 Type-310 stainless steel| Fulinak 14 . -73 Plug 1500 1260 1140 Second phase in hot horizontal leg, possibly in cold leg 104 Nickel Fulinak 14 500 Scheduled 1500 1260 1185 Second phase in all parts of hot leg, possibly in cold leg 118 |Type-316 stainless Fulinsk 14 147 Plugtf? steell ) 214 Inconel Flinak 12 + NaK 500 Scheduledt)? L 116 Type-316 stainless steel{ Flinak 12 350 (s 119 Type-316 stainless steelf Flinak 12 + NaK 350 5 213 Inconelif ®} Fulinak 14 200 o 216 Inconel Fubena 17¢%? 225 e 211 Inconel Fubena 17 375 8 365 Nimonieft? Fulinak 14 225 (g 276 Type-347 stainless steelf Fulinak 14 19 (g {@)aat recorded temperature, usually 1 to 2 hr before termination. €8 Flinak: 11§ mole % NaF, 42.0 mole % KF, 46.5 mole % LiF. (Pylinak: 10,9 mole % NaF, 43.5 mole % KF, 44.5 mole % LiF, 1.1 mole % UF,. (d)This loop held at 1500°F for 70 hr and then turned up to 1600°F for the remsining time. ()ifydrogen-fired with H, atmosphere. YA (’)Lcmps recently terminated, other data not available. (S)Laop still operating. (A)Fybena: COMhis loop two-thirds usual size. 51 mole % BeF,, 47 mole X NaF, 2 mole % UF,. ‘0T HOMVW ONIONA QOI¥dd H0d cS61 i R, i e L .. o L Rl W, o Ao -G e e, i i PG el L i ANP PROJECT QUARTERLY PROGRESSJREPORT 100-hr testé, and corroéion &ata indicate a life on the order of 100 hr/mll of thlckness Chemical analyses of hot- and cold-leg materials after operation indicate an increase in concentration of heavier components in the cold leg and of lighter components in the hot leg. The tube walls of the hot and coldlegs from an Inconel loop (No. 210) that ran for 500 hr with the fluoride fuel (NaF-KF-LiF- UF,) are shown in Fig. 36. The attack shown 1in the hot leg is more than can be tolerated if a thin-walled tube is to be used. Very little, if any, attack is found in the cold-leg section. The hot and cold legs of a type-316 stainless " steel loop (No. 113) that plugged after 123 hr of circulating the fluoride fuel are shown in Fig. 37. The mechanism and nature of the plugging in the stainless steel loops have notyet been determined. Metallic dendrites can be found in these loops, “but as yet it is not certain that Kthey are present in sufficient quanti- ~ties to cause the plugging. Corrosion by Fluorides in a Seesaw Furnace (A. D. Brasunas and L. S. Rlchardson, Metallurgy Division). Two seesaw tests were made with the fluoride m1xture_(10.9 mole % NaF, 3.5 mole % KF, 44.5 mole % LiF, 1.1 mole % UF,) contained in Inconel tubes in which the tubes were "rocked” for 162,000 cycles (450 hr). The hot-zone temperature was approximately 800°C and temperature drops of 120 and 180°C were maintained. Manytiny, metal dendrites were observed em- ‘bedded in the fluoride at the hot and cold ends of the tubes. They were presumably attached to the wall prior ‘to the solidification of the fluoride and have been identified as face- centered-cubic metgl with a lattice parameter of 3.541 A. An analysis of these crystals indicated the following 124 ~and 81.5% nickel. ‘uranium tetrafluoride. 13.6% iron, 4.9% chromium, The fluoride mixture was also analyzed for metal con- stituents, and chromium was found to be the major metallic impurity. This explains the change in composition of the crystals and tube surface from that of the original Inconel (14% Cr, 6.5% Fe, 79.5% Ni). composition: The cold zone also contained appreciable quantities of black crystalsof high melting point embedded in the fluoride. These crystals were ‘identified by x-ray diffraction as uranium dioxide. The oxygen may have come from impurities added to the fluoride during preparationor loading, The roughened surfaces of the hot zones of the tubes indicated that some ‘crystal deposition may have occurred throughout the tube. Preliminary experiments employing the seesaw capsules indicated that the uranium-free fluoride systems do not show the mass transfer phenomenon and that they are generally less corrosive than similar systems containing The mass transfer phenomenon 1s not yet under- stood, but it is probably caused by the reaction of higher valence com- pounds of uranium,. o Standpipe Tests of Fluoride Cor- rosion (A. D. Brasunas and L. S. Richardson, Metallurgy Division). Results have been obtained from tests with Inconel capsules containing the fluoride fuel (10.9 mole % NaF, 43.5 mole % KF, 44.5 mole % LiF, 1.1 mole % UF,). In duplicate tests, with a maximum temperature of 815°C and a temperature difference of 165°C, one capsule showed very slight mass transfer and the other none. In both these tests, however, separation of a second, reddish-black phase occurred, which was assumed to be uranlum dioxide. 1952 FOR PERIOD ENDING MARCH 10, . 0“ ' f Hot and Cold Legs of an Inconel Convection Loop After F1008 o0 Sect lating the Fluoride Fuel (NAF . 36 Fig. ircu &an i QO o™ — —t = © [ o~ a3 Nt w0 bt = % C = = o =] 0' 3 [y, o - O c o = pnd — g - B o =] [ e i ) - O vow E oo ' Q maw oyt 3 0 o] o T H . ~ - 123 ~ Ly 250X. 1250°F. - section, C e . e b § R ORI TP ST 4 k. i i e, ¢ e & %é‘gi?;;m L o e Ehaie %gss AR SR \\w@&@%\ e e = SR Lt SR et T«617 s \§?~:§&\%&§ % : SRR e . : SRk 5 e e quf‘ BT Em e 2T i & ; R W%’%‘S; 2 5 e 5 SRS AR e 2 G 2 R R e N S R S o o PN e Fig. 37. Sections of Hot and Cold Legs of a Type-316 Stainless Steel Convec- tion Loop After Circulating the Fluoride Fuel (NaF-KF-LiF-UF,) for 123 Hours. (a) Cold leg section, <1250°F. 250X. (b) Hot leg section, 70 hr at 1500°F, 63 hr at 1600°F. 250X. ' 126 | £ - . L E- < - - . = - : M - = I - ¥ " o . _ DYNAMIC CORROSIONBY HYDROXIDES The dynamic corrosion of sodium hydroxide in nickel convection loops and of potassium hydroxide in Inconel loops has been investigated., Both these systems have been operated in a carefully maintained hydrogen atmos- phere, and comparable runs have been "obtained with the sodium hydrox- ide-nickel system in air and in vacuum. All these loops were carefully cleaned and hydrogen-fired. The hydrogen itself was first dehydrated and then loaded and maintained under the desired pressure. Appreciable mass transfer occurred in the three sodium hydroxide-nickel thermal convection loops that have been operated. The presence of oxygen definitely increased the oxidative corrosion in the loop; this type of corrosion can possibly be completely eliminated if the loop 1s operated under an atmosphere of hydrogen. One Inconel-potassium hydroxide loop operated for 135 hr showed neither mass transfer nor appreciable cor- rosion. The maximum hot-leg tempera- ture was only 715°C. Preliminary tests in the seesaw furnace and standpipe have been conducted with sodium hydroxide. Standpipe tests under hydrogen atmos- phere have shown neither mass transfer nor oxidation with hot-zone tempera- tures of 740°C., Both seesaw and standpipe tests under vacuum have shown the now-familiar crystal for- mation (predominately nickel oxide) and mass transfer. Corrosidh“By'HydrOXides in Thermal Convection Loeps (G. P. Smith, J. V. Cathcart, W. H. Bridges, Metallurgy Division). Five Inconel loops con- taining potassium hydroxide and four nickel loops containing sodium hy- droxide have been operated. These ST R (T g e G i M " FOR PERIOD ENDING MARCH 10, 1952 loops areof the second type previously described under "Thermal Convection Loops.”™ The care taken in these experiments is exemplified by the following procedure: 1. The loop and hydroxide (in the attached loading pot) were hydrogen-fired at temperatures up to 150°C, 2. The hydroxide was dehydrated by maintaining the system under vacuum for 48 hr at 500°C. 3. The system was refired with hydrogen. 4. The hydroxide was loaded into the loop under hydrogen pressure, 5. The hydroxide in the loop was maintained under hydrogen, air, or vacuum, as the case may have been. Operation of one of the sodium hydroxide-nickel systems ended prema- turely. The remaining three loops were operated in air, vacuum, or hydrogen, and in each loop a con- siderable amount of mass transfer occurred between the hot and cold legs. In the loop operated under vacuum for 317 hr, there were moder- ately heavy deposits of nickel in the cold leg both in the form of dendritic needles and as a dense, compact layer on the loop walls. The loop run under an oxygen atmos- phere for 117 hr (Fig. 38) showed no polished regions even in the hot leg. However, the wall surfaces of the _entire inside of the loop were covered with a heavy, black powder that was found to contain nickelous oxide, metallic nickel, and an unidentified constituent. Crystals of nickel could be distinguished in the cold leg, and there was an extremely heavy deposit of nickel crystals in the 127 U 1 . ANP PROJECT QUARTERLY PROGRESS REPORT X > i * : , iy - s n"’:_{;:‘z(: 3 - A : -~ 5 - ., . 4 5 € "\ . i 3 - & 4 o . A . 3 % 1 4 . 3 { g . * . . ] : w v . = s -4 . - PR Fig. 38. Nickel Thermal Convection Loop Operated for S 117 hr with Sodium Hydroxide Under an Air Atmosphere. o : Y . L T o » top crosspiece of the loop. There was no evidence of intergranular - attack. The sodium hydroxide-nickel system, which operated under hydrogen for 296 hr (Fig. 39), showed considerable mass transfer between the hot and cold leg. The hot region showed a high surface polish that extended over the entire high-temperature region. It was significant that no oxidation occurred inside the loop, as evidenced by a complete absence of black powder. Metallographic examination of this loop has not been completed, and hence no information is available concerning the extent of removal of metal from the hot-leg walls, Operation of four of the five Inconel loops with potassium hydroxide ended prematurely, but the fifth loop was operated for 135 hr with hot- and cold-leg temperatures of 715 and 440°C, respectively (Fig. 40). The run had tobe stopped at 135 hr because a failureof the temperature controller allowed the cold leg to freeze. No mass transfer occurred in this loop. Virtually no change could be noted in the wall thickness of either the hot or the cold legs. A slight smoothing out of the irregularities in the loop wall was noted in both the hot’ and - cold legs, but otherwise there was, little ev1dence of corr031on Corros1on byVSodium Hydrox1de 1n1 Seesaw Tests (A. D. Brasunas and L. S Richardson, Metallurgy D1v131on) Several tests were made in the seesaw dynamic corrosion apparatus by using ASTM-grade nickel in conjunctlon with molten sodium hydrox1de in a vacuum. An abundance of crystals was formed' in the cold zones of the tube afterm' 117 hr, and the usual surface polishing was noted in the hot zones (Fig. 41). Similar tests with nickel and sodium hydroxide produced, in addition to the usual metal crystals, appreciable FOR PERIOD ENDING MARCH 10, 1952 quantities of nonmetallic crystals. Both the black, comb-like crystals and the hexagonal, green platelets ~were positively identified as nickel oxide. Standpipe Tests of Sodium Hydroxide Corrosien (A. D. Brasunas and L. S. Richardson, Metallurgy Division). Two standpipe tests with molten caustic have been completed. One of the tests with nickel and sodium hydroxide was operated under vacuum and the other under a hydrogen atmosphere. For the test in vacuum, the maximum hot-zone temperature was 820°C and a thermal gradient of 140°C was maintained over the length of the pipe. A moderate amount of metal crystal formation occurred, and some oxidation was noted. The test under hydrogen, with a maximum hot-zone temperature of 740°C and a temperature gradient of 140°C, showed neither mass transfer nor oxidation, FUNDAMENTAL CORROSION RESEARCH Several fundamental approaches to an understandingof corrosion phenomena are being pursued. It i1s still too early to say how successful these studles'will be; however, simple, free-energy and equilibrium-constant ‘ ~data have confirmed some of the ex- ‘J{perlmental results, for example, " chromium, iron, andnlckel are attacked by uranium tetrafluoride. Chromium was the most severely attacked and nickel the least. Measurements of the potential differences between cells of various fluorides are being undertaken. Electrode potentials i1n sodium hy- “droxide will be measured as a means of ascertalnlng the purity or thermal history of nickel, nickel oxide, and sodium hydroxide. Polarographic studies of the sodium hydroxide-nickel oxide system will be undertaken if preliminary tests are promising. 129 130 B IR O PP UNCLASSIFIED " LI wi %, €. e i Fig. 39. Nickel Thermal Convection Loop Operated for 296 hrwith Sodium Hydroxide Under a Hydrogen Atmosphere. ~FOR PERIOD ENBING MARCH 10, 1952 UNCL ASSIFIED Y-5417 kS i - P . . [ i B [N = L B e D . [ — 3 P r . B B & e -> I Fig. 40. Inconel Thermal Convection Loap Operated for 135 hr with Potassium Hydroxide Under a Hydrogen Atmosphere. o b s 131 ANP PROJECT QUARTERLY PROGRESS REPORT Fig. 41. with Molten Sodium Hydroxide in Seesaw Apparatus | UNCLASSIFIED | Y- 5936 : rSéctional View of L-Nickel Specimen’After 117 hf.(14,000 cycles) Note polishing at hot zone and abundant crystal formatlon at cold zone. Possible Efifiilibria'Among Fluorides and Metals (L. E. Topol and L. G. Overholser, Materials Chemistry Division). The values of the free energy and equilibrium constant of the following reactlon were computed at 1000 and 1500°K from known thermo- dynamic data (5) M+ xAF = MF, + x4, where M is iron, chromium, or nickel, and A 1s potassium or sodium. The activities were calculated assumingM and AF tolnaof unit activity and - = 1/(x+i) (5) L. Qu1ll (ed.), The Chemistry and Metallnrgy of Hlsc‘ellaneous Materials: Thermo-= dynamics, McGraw-Hill, New York, 1930. 132 Table 33 lists the values of AF and K and also approximate estimates of the activity of the metallic fluoride. These results indicate: sodium fluoride is less corrosive than potassium fluoride and the reaction increases with temperature; chromium is the most soluble of the metals considered, iron is second, and nickel last; and the trivalent cations are slightly less reactive than the divalent (however, from free energy considerations the trivalent 1ons are the more stable at the temperatures considered). Converting the values of a to parts per million, a concentration of 10-1 ppm is found for Fe*? at 1000°K. This figure is much less than the ex- perimentally determined value (similar results are found for the others), ‘and if the activity of AF 1s corrected (since mixtures of alkali fluorides are used), the metallic fluoride (MF) concentrations are further de- creased. Thus;it'seems impossible to explain the cofroSibn of métals by alkali fluorldes at hlgh temperatures by postulatlng the above-mentioned 'equlllbrlum‘reactlon to be the chief effect. FOR PERIOD ENDING MARCH 10, 1952 where M is iron, chromium, or nickel, and the corresponding equilibrium constant is ' B (UF3)* (MF,) (M) (UF,)* in which the activities of the molecular species are denoted by parentheses. TABLE 33 Free Energles and Equxllbrlum Constants for Reactlons of_ : Metals with Alka11 Fluorldes ' ‘ M AF T (°K) AF (cal) K | e Fe*? KF - 1000 +84, 800 10-18.6 10- 6 1500 80, 000 10- 12 10-4 Fet? KF 1000 143,700 10-31.4 10-7-5 1500 138,200 10202 10°° Crt? KF 1000 69, 800 10°15.3 10°% 1500 62,000 10°9-1 10-3 Cr*s KF 1000 111,700 10244 10°% 1500 101, 000 10-14-7 10°3%-7 Nit?2 KF 1000 93,800 1500 90,500 Fe*? NaF 1000 91,000 1500 86,500 - o _ The available free-energy data(®) and several assumptions were used to evaluate the reaction of iron, nickel, and chromium with uranium tetrafluoride at 1000°K and the act1v1ty of the metalllc fluorlde The reaction involved may be written - M+ xUF, = £UF; + MF,, The activity of ¥ = 1, and assuming Cypg = yp, = G then x x+1 from which K 1/(xz+1) a = [___] . (UF )x/(x+1). x* 4 133 S, Sl R L If uranium tetrafluoride is the sole reactant present its activity 1is also 1, and yields 1/(x+1y K ) a = - 1 xx If the uranium tetrafluoride is one of several components of a system, its activity will be approximated by its mole fraction, and therefore the metallic fluoride activity will be decreased. Table 34 gives the values found for free energy, AF, equilibrium constant, K, and metallic fluoride act}V}tlgs, a, (if Qyp, = 1), a, glf equilibrium concentration of UF, = in- - itial concentration = 1 mole %), and a, (1f egquilibrium concentration of UF, = 0.1 mole %). From these results it is seen that al though the trivalent forms of iron ‘and chromium are more stable, the divalent i1ons will be preferentially ANP PROJECT QUARTERLY PROGRESS REPORT will readily be reduced by the pure metal. Of the metals considered chromium i1s the most readily attacked, iron is next, and nickel last. 1In addition, a comparison of these results with those found for the alkali fluorides indicates that uranium tetrafluoride is the much better oxidizing agent of the two. (It should be stated that the accuracy of the data used 1s not certain and even qualitative conclusions may be in error. ) There 1s some experimental evidence that beryllium oxide will react with uranium tetrafluoride at high tempera- tures to form uranium dioxide and beryllium fluoride, according to the following formulas: 2Be0 + UF, = U0, + 2BeF, , (U0,) (BeF,)? K = ~——r———— (Be0)? (UF,) forméd. In fact, it can easily be shown thermodynamically that any (the activity of a molecular species trivalent iron or chromium formed 1is denoted by parentheses). From TABLE 34 Free Energies and Equilibrium Constants for Reduction | of Uranium Tetrafluoride with Metals M AF (kcal) X a, a, ay Fé+2 +20.0 10-4-4 10-1-6 10-2-°9 10-3-6 Fe'? ‘ 45.0 1093 1027 10-4 2 10" S Cr+2 4.0 10-0-88 10-9.5 10-1-8 10-2-5 Cr*3 13.0 10-2-85 10-1 10-2-5 10-3-2 Ni 928.0 10-6- 1 10-2-2 10-3:5 10°4- 2 el f 134 Gomid it available data,(3'%'7) the free energy and equilibrium constant of this reaction were calculated to be t+4.,2 kcal and 10°%+?, respectively, at 1000°K. If the beryllium oxide and uranium tetrafluoride are assumed to be of unit acF1Y1ty, and %50, = %Ber, a, the equilibrium constant is given by 4a° . This results in a value of 10°°+* = 0.4 for the approximate activity of uranium dioxide and 0.8 for beryllium fluoride at equilibrium. , Although the accuracy of the data is unknown, there seems to be sufficient evidence for the assumption that the above reaction occurs to a reasonable extent. ' EMF Measurements in Fused Fluorides (L. E. Topol, L. G. Overholser, Materials Chemistry Division). A series of experiments are planned for studying the possible mechanisms of fluoride corrosion by measuring the potential differences of various cells containing the fluorides of nickel, chromium,or iron dissolved in fused alkali fluorides. Concentration cellsof ;hé following type will be investigated: M/MF+(C) in fused AF/MF-(C) in fused AF/M where M is iron, chromium, or nickel, and A 1s sodium, potassium, or lithium. (G)L. Brewer, L. A. Bromley, P. W. Gilles, and N. L. Lofgren, The Thermodynamic Properties and Equilibria at High Temperatures of Uraenium Helides, Oxides, Nitrides, and Carbides, MDDC-1543, Sept. 20, 1945. : (7)0. Kubaschewski and E. L1, Evans, MHetal- lurgical Thermocheristry, Academic Press, New " York, 1951. - FOR PERIOD ENDING MARCH 190, 1952 Binary and tertiary mixtures of the alkali fluorides will be studied, and the effect of small additions of uranium fluoride will be evaluated, if possible. Electrode Potentials in Fused Sodium Hydroxide (Ambrose R. Nichols, Jr., Materials Chemistry Division). The apparatus previously described(®’ has undergone continued modification over the past three months, It consists of a nickel vessel within which 1s placed a porous cup, which divides the contents into two electrode compartments. A nickel electrode is suspended in each compartment., Chromel - Alumel thermocouples in nickel walls are located against the outer wall of the nickel cup and in each of the compartments. The whole assembly si1ts at the bottom of a closed stain- less steel container that rests in a 5-in. pot furnace. Helium purified by passing over copper turnings at 450°C and through magnesium perchlo - rate tubes 1s passed through the container at a pressure slightly in excess of atmospheric. When a measurement 1s to be made, the desired amounts of purified sodium hydroxide and nickel oxide are weighed in a dry box and placed in the two electrode compartments of a nickel container. The closed container is then removed to the furnace and the necessary connections made. Helium circulation is started before heating begins. After the chosen temperature has been reached, the potential between the two electrodes is determined by using a type-K potentiometer, although in some casesa recording potentiometer has also been used. (S)A. R. Nichols, ‘*“EMF Measurements in Hy- droxides,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 10, 19571, ORNL-1170, p. 110. e g ’ ks Pil 2 ¥, Y 135 ANP DIVISION QUARTERLY PROGRESS REPORT The results have thus far not been reproducible. The measured potentials have changed with time, even to the extent of reversal of polarity. It was observed in each case that the electrode in the more concentrated solution developed a deposit of fine nickel crystals, whereas that in the more dilute solution appeared to have undergone polishing. This mass transfer has so far appeared to be independent of temperature over the narrow temperature range used. Three principal experimental difficulties exist and each could account for the poor results: (1) fail- ure to find a suitable diaphragm material; (2) lack of knowledge of the true nickel oxide concentration in the solutions in whichthe electrodes are placed; and (3) difficulty of maintaining a uniform temperature ‘throughout the cell. Until these problems are overcome, it 1is not possible to ascertain the effect of such factors as the purity or the thermal history of the nickel, nickel ‘oxide, or sodium hydroxide. Pélarbgraphy of Sodium Hydroxide in Silver and Platinum (R. A. Bolomey, Materials Chemistry Division). In ‘previous reports¢{®) it was stated that anhydrous sodium hydroxide heated to a temperature of 350 to 600°C in either a silver or platinum crucible in vacuum resulted in the production of ionic species that gave character- istic polarographic waves. It was also mentioned that the voltage at which the peaks occurred was tempera- ture-dependent. To date, 85 polaro- grams on sodium hydroxide in silver crucibles and 88 polarograms on sodium hydroxide in platinum crucibles have been obtained. Even though all the data were not obtained at the same temperature, enough information has been obtained to analyze the reproducibility of the results. Such an analysis revealed considerable 136 randomness in the waves and that the curves of a given series show only qualitative similarity. The source of randomness in the position of the polarographic waves is not understood. It is possible that it is to be found in the instru- ment itself as a result of the method employed to collect the data. At- tempts to obtain the data by allowing the readings to come to equilibrium after each setting proved to be impractical because of the slow rate of equilibrium attainment. The present me thod of collecting the data requires that the potential on the cell be varied at a uniform rate, since any fluctuation in the rate of potential changes can produce anomalous effects on a current-voltage curve of the type obtained with a polarograph and stationary microelectrodes. It may be that these anomalous effects are more apparent when operating the cell at the high temperatures required in this work than when operating at room temperatures in aqueous solutions. Equipment for differential thermal analysis has been gathered so that preliminary tests may be made to study the feasibility and applicability of this method to the study of corrosion mechanisms. It 1s intended first to apply the method to the system sodium hydroxide-nickel oxide under an inert atmosphere. Magnetic Susceptibility of Stain- less Steel Exposed to Fluorides (W. C, Tunnell, ANP Division). It has been observed repeatedly that the surfaces of all normally nonmagnetic steels become magnetic after exposure to high-temperature fluorides. Surface layers have been removed and analyzed, and it appears that the iron-to- chromium ratio increased. A micro- spectrographic analysis, under electro- magnetization using a colloidal dispersion of iron particles, revealed ¥ ! Y o ! Ty BV, ?‘ 0w P eidgacs [ the magnetic material to be at the grain boundaries where intergranular penetration had occurred. In one test using a gasket ring of type-316 stainless steel, the magnetic surface was submerged in concentrated nitric acid, and a 1- to 2-mil layer of stronglymagnetic material separated that was chemically analyzed as >90% iron, 2.5% chromium, and 1.5% nickel. The material under the layer was not magnetic. In another test the hot section of an Inconel loop that had been exposed to Fulinak was examined, since 1t seemed more magnetic than the cold section. Immersion in con- centrated nitric acid left pits and made the material less magnetic, and the remaining surface layer appeared to be pure nickel. Another specimen FOR PERIOD ENDING MARCH 10, 1952 of type-316 stainless steel that had been in contact with Flinak became porous after treatment with concentrated nitric acid and lost all its magnetic properties. It may be concluded that, as 1in most intergranular corrosion of stain- less steels, the chromium 1s prefer- entially leached at the grain bound- aries, possibly by the usual device of carbide precipitation, which exposes excess 1iron and nickel for attack and gives the magnetic susceptibility, Photomicrographs appear to bear out a phenomenon of this type. The attack on iron always appears greater than on nickel, as has been shown, for example, by tests where the rate of penetration appears to be proportional to the amount of iron in the metal, 137 T T—— TR T ON TSR WY IV TR0 1 et i, o S e 1 £ A Ak e g A e o i i, S SR R Nk BN G e i ) . S o M R B i Bt s S, Bl B e e SRR i iom o i . 2 ik R g SR i 12. METALLURGY AND CERAMICS W. D. Manly, Metallurgy Division T. N. McVay, Consultant The drawing of tubing from the previously prepared solid fuel plates has been accomplished, and the results of the drawing on the bond continuity and oxide distribution of the solid fuel elements are discussed. From this work it was found that the best core material to use in the uranium oxide mixture was iron powder. Com- patibility experiments have shown boron carbide to react rather ex- tensively with the 300-series stain- less steels and Inconel but to be inert to type-430 stainless steel; therefore the ARE safety rods will probably be constructed of type-430 stainless steel. The solid-phase bonding of metals is always difficult with materials that are in intimate contact at elevated temperatures. Consequently, self-welding experiments are being conducted on the possible ¢combinations of materials to be used in the safety rods. ' The use of the cone-arc welding process for the production of tube- to-header joints is being investigated to determine the applications and limitations of the process, since there is a need for a reliable auto- matic or semiautomatic welding method for the production of the many tube- to-header joints in ANP core and heat exchanger designs. A description of the equipment and operating procedure and a discussion of the variables being studied are presented. Because of the complexity of the ANP type of heat exchangers, it i1s probable that extensive brazing will be necessary in fabrication. Preliminary work for heat exchanger assemblies indicates that brazing can produce sound joints if proper control of the brazing variables 1s exercised. Nicrobrazing alloys, Pd-Ni, Mn-Ni, Ag-Pd, Ni-Cr-Sr, I h i‘;e‘ '.Mé' §~£:’ i and Ni-Cr-Si-Mn, are being studied and their corrosion in hydroxides and fluorides is being investigated. Since Inconel has been designated as the structural material for the ARE extension, rupture and creep data are being obtained at the operating tem- perature of the reactor. Data are presented to show (1) the time re- quired to produce deformations of various percentages and for rupture to occur as a function of the applied stress for fine- and coarse-grain Inconel, and (2) the minimum creep rate and per cent deformation per hour for the two types of Inconel plotted as a function of applied stress. The oxide ceramics or combinations of ceramics and oxides appear most promising for reactor application be- cause they offer the best combination of structural integrity and thermal properties. Ceramic coatings for Inconel and stainless steel are being tested. Of those tested, the NBS Ceramic-Coating A-418, with its high temperature and oxidation resistance characteristics, shows promise as a coating for the ARE radiator and pos- sibly other structural parts. FABRICATION OF REACTOR ELEMENTS E. S. Bomar J. S. Coobs Metallurgy Division Tubular, solid, fuel elements in which the fuel compact is between two concentric tubes are being de- ve loped but have not yet proved en- tirely satisfactory. The bonding and distribution of the uranium dioxide powder between the tubes after cold drawing has varied from very poor to’ P LR vy e Ty 139 ANP DIVISION QUARTERLY PROGRESS REPORT fair. It is hoped that better initial bonding and a different cold-drawing technique may yet yield good, small, tubular, solid, fuel elements. ' Tests for solid-phase bonding (self-welding) of the movable parts of the ARE control rods at reactor temperatures (1472°F) indicate a fair amount of reaction between the boron carbide type-316 stainless steel and boron carbide Inconel systems in 100 hours. Of the few materials tested the reaction between boron carbide and type-430 stainless was the least marked ‘Cold Drawxng of Tubular SOlld Fuel Elements. Tubular, solid, fuel ele- ments fabricated by joining two seml- circular laminated plates either with seam welds or by "rubberstatic"(1) pressing methods have been subjected to cold-drawing operations at the Superior Tube Company. All samples, except the one prepared by rubberstatic pressing, were given an initial hot reduction of 62 to 75% to bond-pressed core, picture frame, and cladding. This operation was carried out on the flat stock before forming into tubes. The cores of the samples were 30 vol % uranium dioxide and 70 vol % metal (type-302 stainless steel, iron, or nickel). A reductlon schedule based on com- _ mercial practice was set up at values ranging from 18 to 28% after a test reduction of 38% caused a tube failure. "The weld seams were satlsfactorlly - smoothed in this pass, and the tubes were subsequently reduced 87.5% in seven passes, which yielded tubes of 1/4-in. OD by 0.015-in. wall thickness. Three of these tubes were then further reduced to yleld tubes 1/8 in. Bemar, “Fuel-Element Fabrication,” Adircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1951, ~ ORNL-1154, p. 149. (I)G. M. Adamson, E. S. Coobs, 140 W 4 e T e OD by*"‘ : and J. H. 0.015-in, wall thickness. Metallo- graphic examination of specimens taken from the tubes showed that bonding and uranium dioxide distribution varied from very poor to fair for tubes pre- pared from rolled plate and was also fair for one specimen prepared by rubberstatic pressing. Sections trans- verse to the direction of drawing are shown in Fig. 42 at reductions of 87.5%, that is, for 1/4-in.-0D tubes. In general, further reductions to 1/8-in.-0D tubing only exaggerated the defects found in the 1/4-in.-0D stock. Although plug drawing was not tried on any of the composite tubes, 1t was the opinion of some of the technical staff at Superior Tube Company that plug drawing would not disturb the core of a laminated structure so much as drawing on a mandrel, which was the technique used in all drawings to date. Rod drawing produces slip cracks in the core because of slightly uneven reduction of the two cladding layers, but it i1s believed that plug drawing should give uniform reduction. Variation in the reduction schedules used for these tubes will be tried by using the Metallurgy Division draw- bench and emphasizing less severe reductions. Mandrels are now on order to supplement dies already on hand. . Equipment for plug drawing 1is not avallable at present ARE Control Rod. At the operating temperature of the ARE, there exists the possibility of solid-phase bonding of metallic components that are in close contact. This phenomenon could possibly lead to malfunctioning of the control rods if it occurred between the control rod cladding and the walls of the tubes in which they are housed. Several tests have been made in which samples of two different metals were pressed together at 1472°F. One- eighth-inch-diameter specimens of the oo s f, e > i, { A E:\ k ; fw.r f‘ oo F ka; FOR PERIOD ENDING MARCH 10, 1952 Y5299 1 SO T R & TYPE-3(6 STANLESS : o o | STEEL CLADDING . XY, . S CORE BY VOLUME 30% UO2 (-100 MESH AS RECEIVED 70% Ni(-325 MESH) PR | " | o | ' ;-*TYPE~316 STAINLESS e | S / STEEL CLADDING CORE BY VOLUME 30% U0, (~200, +325 MESH) 70% TYPE-302 STAINLESS STEEL (-325 MESH) Fig. 42. Transverse Sections Through %-in.-OD,Cold-Drawn, Tubular, Solid Fuel Elements. (a) Tube prepared from rolled plate. 200X. (b) Tube prepared by rubberstatic pressing, 200X, (Prints reproduced 83% of original size. ) 141 i, ™ T e g ke A e | ¥ systems. dissimilar materials were subject to 250 psi for 100-hr intervals. This pressure is probably higher than would be encountered radially at the control rod and control-rod-tube contact points. Type-316 stainless steel and Inconel, in various atmospheres, were used in the initial tests. However, results of compatibilify'tests made available after a portion of the welding tests had been run indicated a fair amount of reaction between the boron carbide-type-316 stainless steél and boron carbide-Inconel systems. A less marked interaction was found when boron carbide was "contained in type-430 stainless steel. The solid-phase welding schedule was revised to include the type-430 vs. type-316 stainless steel and Inconel ~ However, the only bonding evidenced that could not be removed ‘with finger pressure was between similar metals. Tests of the type-430 vs. type-316 stainless steel system with tank helium atmosphere are now being conducted. Exposure to the ‘other atmospheres will be carried out if results indicate that this 1is desirable. | CONE-ARC WELDING P. Patriarcar G. M. Slaughter Metallurgy Division The feasibility of the use of the cone-arc welding process for production of tube-to-header joints has been demonstrated by previous 1nvestiga- tions.(%?’ The welding group of the ‘Metallurgy Division is investigating the applications and limitations of the process, since the need for a reliable automatic or semlautomatic process for production of the many tube-to-header joints required by (2)E. R. Mann, Means for Making Unifornm Circular Heliarc Welds by Deflecting the Ion Beanm Continuously, ANP-63, Apr. 9, 1951. 142 ANP DIVISION QUARTERLY PROGRESS REPORT ANP core and heat exchanger designs is evident. Equipment. The apparatus consists of a G-E, 200-amp, direct-current welding generator with the addition of a resistor circuit that permits the use of currents as low as 3 amp, 1if desired. A Miller Electric Company, high-frequency, spark-gap oscillator provides the high frequency required for starting the inert welding arc without having to touch the 1% thoriated tungsten electrode to the work. The apparatus has been provided with a timer and contactor control circuit to render the arc timing an automatic operation. Principle of Operation. The cone- arc welding technique is particularly suited to welding a tube to a header plate. The nozzle of the welder 1is a permanent Alnico magnet that sets up lines of flux to the tube and plate over which the nozzle 1s held. Since the welding arc will strike from the electrode tip to a point on the periphery of the tube, the arc stream is cut by the lines of flux at a small angle. This angle sets up resultant forces tangential to the tube periphery in the plane of the header sheet, and the forces rotate the arc around the tube at very high speeds. To the observer the arc appears as a cone; hence, the name "cone arc." The arc raises the temperature of the tube edge and the periphery of complete header hole in a uniform manner until the melting point its reached and a circumferential weld results. Experimental Procedure. Some of the variables that may affect the operation are: 1. Arc current and time, 2. Arc distance, which affects the arc voltage and the angle between the arc stream and the magnetic lines of flux, 1y £, P i‘b ‘3 Ll \“"3 T L A g’:}; ot E ~arc was struck. 3. Magnetic nozzle to work distance, 4. Choice of inert gas and rate of flow, which affect the arc voltage (higher for helium than argon) and the turbulence, 5. Work geometry, that is, tubing ' size, header thickness, and heat transfer uniformity of header hole- ‘to-hole distance, hole-to-header- edge distance, and centering of work and electrode. In the initial experiments header holes without tubing were used, since it was assumed that the behavior of the arc in the formation of a circum- ferential molten pool would yield representative information with or without a tube. The header material ‘used was type-304 stainless steel sheet 1/8 in. thick. Holes were 3/16 in. in diameter and spaced 3/8 in. center to centéer. Welding conditions were as " follows: 1. 1/16-in. diameter electrode, 2. Electrode to plane of work dis- tance, 0.070 in., 3.. Magnetic nozzle (tip of soft iron) to plane of work distance, 0.41 il’l,, ’ a4."Argon flow, 30 ft /hr, %5, Open 01rcu1t voltage, 70 - 6}H'Arc'voltage,_11, 7. Curfént,'#ariéblé;' 8. Tlme, varlable . 4‘3 w" ' Dur1ng weldlng,pools of molten 'metal formed around the periphery’ of the hole about 20 sec after the The pools grew until they encompassed the entire periphery of the hole and effected the desired ~decreasing arc txme,} FOR PERIOD ENDING MARCH 10, 1952 weld. © As would be expected, 'the size of the heat-affected zone, as de- termined by the size of the heat-tinted zone surrounding the header hole, diminished with increasing current and The limiting value of permissible 'hole-to-hole distance will probably be a function of the arc current and time for a given header thickness and hole size. Experiments will be conducted to de- termine these limiting values, since ANP designs may require as small a hole-to-hole distance as possible in order to obtain the maximum number of heat exchanger tubes per unit area of header sheet. The results from the few experiments made were inconclusive, but it 1s ex- pected that further work during the coming quarter may yleld sufficient information for a comprehensive study of the quality of cone-arc welded, tube-to-header joints as a function of the welding variables. BRAZING P. Partriarca G. M. Slaughter Metallurgy Division Experiments were made to determine the feasibility of brazing since 1t is quite probable that this method may be used extensively in the fabrication of the ANP type of heat exchangers. .‘Tube'bompacts with tubes, 0.100 in. OD with 0.010-in. wall thickness, have “‘been successfully brazed to baffle - plates that were 0.020 in, in thick- ness. Preliminary work on such as- ... semblies indicates that brazing can " "be used advantageously in the produc- --.~tion of sound joints if proper control is maintained of such brazing variables ~as joint fit, joint geometry, degree of base-metal cleanliness, and qualaty of the furnace atmosphere during the brazing process. Corrosion tests of Nicrobraz and a 60% palladium-40% ) | 143 Sk, A 2 RGR Ka dinl v 2 gl s i, ‘@yend of a 6-1n. —long butt joint. ;legs + Inconel strips, S oo e, L0 'de31rable ANP DIVISION QUARTERLY PROGRESS REPORT nickel alloy in sodium hydroxide and in UF,-NaF-KF-LiF mixture showed the palladlum -nickel alloy to be far more corrosion resistant in both fluids at 1472°F, but the attack by the fluorlde on elther braze was severe. Flow Iests. A relat1vely long-term program for evaluating various brazing alloys for the high-temperature ANP type of application is anticipated because of the promising results ob- tained in the experiments on the feasi- blllty of bra21ng A series of tests were conducted to determine the mini- mum temperature at which the various bra21ng alloys flow readily. The test spe01men consisted of a moderate amount of brazlng alloy placed at one Both of the joint were 0.062-in. one of which was milled flat on the mating surface to facilitate the production of a tight- fitting 301nt The specimens were then heated untll the temperature was ‘i‘found atwhlch the braz1ng alloy flowed “freely up the entire joint length of .6 inches. “fi}these experlments that any oxide scale It became apparent during ofi the metal specimens is highly un- - The brazing alloy does not ea51ly wet scaled surfaces and flow around a 301nt is 1mpeded A very dry hydrogen atmosphere, ‘with a dew point in’ the range of -60°F or below, prevented scallng in most cases; therefore moisture was ‘apparently the ‘prominent factor in scale formation. It was suspected that if the hydrogen continuously reduced any oxide and there were subsequently moisture forma- "tion in the hydrogen atmosphere, the exit dew point of the hydrogen would decrease with increasing flow rate. A series of experiments proved this hypothesis. Tests with an inlet-gas dew point of -78°F and high hydrogen " flow rates showed that exit dew points ‘approximating those at the inlet were attained. The experiments were sig- nificant in that they made it obvious o with a higher magnification, that a relatively large hydrogen flow rate during furnace brazing was de- sirable. ' “ Corrosion of Brazing Alloys. The resistance to corrosion of the various brazing alloys, alone and in combina- tion with various base metals such as found in joints, is important in the selection of suitable joining media; therefore emphasis has been placed on the results of corrosion tests of the brazing alloys in fluorides and hy- droxides. Extensive tests wereé con- ducted on Nicrobraz, a boron-containing alloy, which is an excellent brazing alloy for use in conventional high- temperature applications. Samples of the pure alloy were subjected to corrosion experiments in fluorides and hydroxides and the baths were later analyzed chemically for boron. A sample of Nicrobraz treated in sodium hydroxide for 100 hr at 1500°F, showed the heavy corrosive attack characteristic of this medium. There was generally heavy surface attack to a depth of 60 mils and severe local attack. The sample tested in the NaF-KF-LiF-UF, mixture for 100 hr at 1500°F was moderately pitted, par- ticularly near the intermetallic components, to a depth of 4 mils. Other surface areas were relatively untouched. Results of the chemical analyses of the contents of both the fluoride and hydroxide baths 1indicate that boron is preferentially leached from the alloy. An attempt will be made to determine the actual percentage loss of the boron from the alloy. .The corrosion resistance of Nicro- brazed tube-to-header joints with Inconel as the base-metal has been studied. A section of a tube-to- header joint exposed to sodium hy- droxide exhibited excessive attack both at the brazed joint and on the base-metal, as shown in Fig. 43e¢ and, in Fig. 43b. The relatively mild attack of o TR D : % v + 144 DU A 1952 'FOR PERIOD ENDING MARCH 10, Corroéion Tést of Nicrobrazed Inconel Tube-to-Header Speéimens 43. Fig. Exposed to NaOH for 100 hr at 1500°F. (b) 250X. (a) 25X, 145 e [ bi 3 gy e e i, . R itk e st e B i s ol Bt et e i il i R N o, i i B B i, ALl . B M B 300 50 b e ot B oo IR s BC i i Bl B B it w TARY R TR TN e S e * ANP DIVISION QUARTEBLY PROGRESS REPORT the mixture NaF-KF-LiF-UF, on these joints is shown in F1g 44, which 1s a photomlcrograph of a joint tested in’ the fluoride mlxture. Chemical analy- ses of the baths will be given in later reports. A few experlments have also been conducted on the 60% palladium-40% nickel and 60% manganese-40% nickel alloy systems. Small ingots of the palladium-nickel alloy were tested for 100 hr at 1500°F in both the sodium hydroxide and the NaF-KF-LiF-UF, mix- ture and gave promising results. The corrosion resistance of this alloy was excellent the attack was less "than 2 mils in both cases. An Inconel 'tube to- header joint brazed with the 60% manganese 40% n1ckel alloy was Fig. 44. Corrosion Test of Nicrobrazed Inconel Tube-to-Header Joint Ex- tested for 100 hr at 1500°F in sodium hyd}oxf&é; and there was moderate attack ‘at the joint (Fig. 45). More extens1ve'corr051on tests on,these alloys will be performed, and the baths will be analyzed for the presence of the various elements in the brazlng alloy being 1nVest1gated ~Similar tests will be conductéed on other high-temperature brazing alloys of current interest, including 75% silver-20% palladium-5% manganese, 64% silver-33% palladium-3% manganese, 60% palladium-37% nickel-3% silicon, 16.5% chromium-73.5% nickel-10% silicon, 16.5% chromium-71.5% nickel- 10% silicon-2.5% manganese. No data are yet available on the resistance to corrosion of»these ‘alloys. Methods posed to NaF-KF-LiF-UF, for 100 hr ‘at 1500°F. 250X. 146 FOR PERIOD ENDING MARCH 10, 1952 LR R Fig. 45. Corrosion of Inconel Tube-to-Header Joint Brazed with a 60% Manganese~-40% Nickel Alloy after 100 hr at 1500°F in NaOH. 100X, have been devised for coating the inside of the standard nickel test capsules with brazing alloy so that corrosion resulting from dissimilar metal combinations can be eliminated when testing samples of the pure brazing alloy. When brazed joints are tested for corrosion, a capsule of the same metal as that of the joint base-metal is used to eliminate dis- similar metal combinations. MECHANICAL TESTING OF MATERIALS R. B. Oliver J. W. Woods : - C. W, Weaver Metallurgy Division The structural integrity of Inconel at the operating temperature of the ol reactor is being accurately determined. Elongation, rupture, and creep rates vs. stress curves have been obtained for fine- and coarse-grained Inconel specimens tested in argon at 1500°F, Three tube-burst tests with stresses up to 4000 psi have been running con- tinuously for 1300 hr without develop- ing detectable leaks. Inconel Creep and Stress Data. Elongation, rupture, and creep rate Vvs. stress curves were obtained with fine- and coarse-grained Inconel sheet specimens 0.065 in. thick, 0.500 in. wide, and 3 in. long. "The extension was measured during the test by ob- serving the relative displacement of two platinum strips fastened at opposite ends of the test section. A Gaertner " tcovas 1 147 o1 £ ANP DIVISION QUARTERLY PROGRESS REPORT micrometer microscope having a least division of 50 pin. was used to evalu- ate this relative motion. The time to produce deformations of 0.1, 0.5, 1, 2, 5, and 10% and the time for rupture to occur, as a func- tion of the applied stress for fine- grained Inconel, are shown in Fig. 46. Also presented is the minimum creep rate in percentage of deformation per hour as a function of the applied stress. The percentages appearing above the rupture curve are the total elongations at the several stresses. The fine-grained material was cold rolled and bright annealed at 1650°F to produce a grain size of approxi- mately 0.105 mm (90 grains/mm?). Similar data for the coarse-grained Inconel-sheet specimens are presented in Fig. 47, but the curve representing the times to produce deformations of 10% is omitted since it is nearly coincident with the rupture curve. This Inconel sheet was annealed for 2 hr at 2050°F in a hydrogen atmos- phere to produce a grain size of ap- proximately 0.250 mm (15 grains/mm?). Two separate creep rates were ob- tained for the fine-grained specimens tested at 3000 and 1850 psi; these rates were observed during three different tests. Linear sections were observed on the strain-time plots at 50 to 150 hr and again at ‘about 300 and 600 hours. The earlier of the two linear periods exhibited creep rates of one-third to one-half of the rate found during the later linear period. Tube-Burst Tests. Three tube-burst tests with tangential stresses of about 1000, 3000, and 4000 psi, re- spectively, have been running for ap- proximately 1300 hr without developing detectable leaks in the type-316 stain- less steel tubes. These tubes, with 480 mils OD and 10-mil wall thickness, 148 were loaded internally'with‘argon under pressure and exposed to an ‘environment of stagnant air. Since no method exists to gage tubular specimens during the test, the re- sults will be based on before and after measurements, ‘ - T A second phase of this work is to obtain data on physical properties of metals in the fluorides. Apparatus used for a similar study of sodium has been modified and calibrated so that fluorides may be used. Two tube-burst tests were run in NaF-KF-LiF-UF, (10.9 mole % NaF, 43.5 mole % KF, 44.5 mole % LiF, 1.1 mole % UF,) by using the apparatus described in the previous ANP quarterly report. Type-316 stainless steel has withstood a hoop stress of 2320 ps1i and the Inconel 2170 psi. Both these tests have been 1n operation over 400 hours. Stress-to-rupture tests in the same environment and temperature but of 1000-hr duration were also run on Inconel and type-316 stainless steel tubes. These materials withstood hoop stresses of 1200 psi and 1700 psi, respectively, and the ~diameters increased 1.7 and 0.7%, respectively. Operation of Creep and Stress- Rupture Equipment. The center of gravity of the lever-arm counter weight on each testing machine was raised approximately 2 in. above the plane of the knife edges prior to making a series of creep and stress-rupture tests. With the counter weights thus raised, the tare weight of the lever arm was 1ncreased by the same amount that the spring load of the compressed sealing bellows was decreased, so that a constant load could be maintained during test. The 1initial, or tare, load of each testing machine was measured with wire strain gages mounted on a duplicate specimen and the constancy of the load was also verified. ' i 6¥1 STRESS (psi) 0.00014 30000 20000 10000 9000 8000 7000 §000 5000 4000 3000 2000 1000 o4 I3 Fig. CREEP RATE (% per hr) 0.004 Q.01 0.1 10 46. INCONEL SHEET, GOLD-ROLLED AND ‘ANNEALED AT 1650°F GRAIN SIZE, APPROXIMATELY 90grains /mm2, 0105 mm DA TESTED IN ARGON AT 4500°F TIME 1S SHOWN FOR 04,05, 1,2,5, AND 10% EXTENSION AND RUPTURE VS. STRESS; EXTENSION MEASURED OPTIGALLY 10 TIME (hr) 100 1000 Creep and Stress-Rupture Data for Fine-Grained Inconel Sheet. wor UNCLASSIFIED DWG. 14445 10 10000 ‘0T HOMVIW ONIONT @0I¥dd ¥od ¢cS6l1 TEERTI e ey 0ST . o UNCLASSIFIED _ C ' _ DWG. 14446 : : : ‘ CREEP RATE (% per hr) 0.0001 0.004 oN 01 : 1 10 © 20,000 INCONEL SHEET, ANNEALED FOR 1hr AT 2050°F GRAIN SIZE APPROXIMATELY 15grains/mm? 0.250mm DIA TESTED IN ARGON AT 1500°F (0000 _ 9,000 239 8,000 ~_ 22% 7,000 —t > ~ 6, £ 5000 I T~ = 4 4000 S o> 5 1 Se— \ ~35 3000 TIME iS SHOWN FOR 04,05, 1, 2, AND 5% 2,000 EXTENSION AND RUPTURE VS. STRESS; EXTENSION MEASURED OPTICALLY ol { 10 100 1,000 - 10,000 TIME (hr) Fig. 47. Creep and Stress-Rupture Data for Coarse-Grained Inconel Sheet. J40ddH SSHHSOH_d ATHALYVNO NOISIAIQ ANV . CERAMICS LABORATORY T. N. McVay, Consultant Consideration of the possible appli- cation of ceramics-coated materials to reactors has indicated several po- tentialities that require further development. A ceramic coating has been applied to a high-temperature radiator and appears to offer excellent oxidation resistance. Petrographic examination of fluoride fuels is being undertaken and additional equipment -~ mainly high-temperature furnaces - have been added to the laboratory facilities. Ceramic Applications to Reactors. Some general conclusions can be given concerning the usefulness of ceramics in reactors on the basis of an analy- sis of their physical properties and other information. ¢ 1. Glass and glass-bonded ceramics, which include nearlyall "conventional” ceramics, are not useful in high- temperature reactors as structural elements because of a tendency to soften at relatively low temperatures, brittleness, poor mechanical and thermal shock properties, and poor thermal conductivities. However, 1f they can "be used as liquids, grains, powder, or coatings they may be considered. 2. Oxide ceramics generally have most of the above-mentioned disad- vantages, often to a lesser degree, but they do not exhibit the tendency to soften at low temperatures. All the oxides with the exception of beryl- lium oxide have thermal conductivities considerably less than those of most metals. If the thermal and Strfictural_ requirements are not too great and the coolant corrosion problems can be (3)J. R. Johnsen,. Ceramic Materials as Related to the Reactor Program, ORNL CF-52-1-144, Jan, 18, 1952. FOR PERIOD ENDING MARCH 10, 1952 met, oxides may be used in reactors, perhaps even as structural material. Oxides in combination with other ma- terials offer the greatest promise. 3. Carbides, nitrides, borides, hydrides, and sulfides have been investigated, but most of them need further study. Beryllium carbide was intensively investigated and developed in the NEPA project and will be studied further under G.E. It has many de- sirable properties and if successfully coated to prevent corrosion (particu- larly oxidation) and loss of fissicn products, should prove useful in high- temperature reactors. Although many of these unusual materials are very refractory, they are also very re- active at high temperatures, particu- larly in the presence of oxygen and water vapor. 4. Ceramic-metal combinations (Cermets) have been studied intensively by groups sponsored by the Air Force. Both structural materials and coatings have been developed. A few that have shown promise are Al,0,-Cr, Al,0,-Fe, Al,0,-Ni, MgO-Ni, and TiC-Co. The Cr-Al1,0, Cermets have some interesting properties such as reasonably high strength at temperatures up to 2372°F and very good oxidation resistance at these temperatures. They are somewhat brittle, however, and are not too resistant to thermal shock. This property depends on the oxide content. Fabrication techniques have been well developed for some shapes. Consider- able further study will be necessary to determine their usefulness in high- temperature reactors. Coatings for the Radiator. A con- siderable amount of work has been done this quarter in applying ceramic coatings tostainless steel and Inconel, since these may be required in a liquid-to-air radiator. Much promise for high-temperature and oxidation resistance for reactor materials has 151 been shown by NBS Ceramic-Coating 'A-418. Samples coated to about 2 mils thickness are currently undergoing testing. The computed oxide composi- tion of the frit (glass phase of the coating) is: R Si0, '37.5 B,0, | 6.5 -BaOQ 44.0 CaO , , 3.5 " ZnO 5.0 Al,0, 1.0 ZrO2 2.5 100.0 Similar coatings developed for the Air Force have been shown to protect stainless steel at 2000°F for over 200 hr, or 1800°F for an almost in- definite period. Testing procedures for these coatings are described in . the literature. ’ (4) (_4)W. N. Harrison, D. - G. Moore, and J. C. Richmond, *“Ceramic Coatings for High-Temperature Protection of Steel,’”” RP1773, J. Research Nat, Bur. Standards 38, 293-307 (March 1947). 152 -ANP DIVISION QUARTERLY PROGRESS REPORT Ceramic Laboratory Egquipment. A high-temperature molybdenum-wound furnace designed by Thomas Shevlin, Consultant from Ohio State University, will be used for Cermet work. The shell and accessory components are being fabricated in the Y-12 shops. The refractories and windings are either on hand or on order. =~ The high-temperature x-ray furnace and the small, high-temperature vacuum furnace have been built. The thermal- diffusivity and thermal-expansion equipment have been designed and are in the shops. The high-temperature dilatometer 1is 1n use. Microscopic Examination 6f Fluo- rides. Petrographic examination of fluoride fuels is under way, and the solubility of uranium tetrafluoride in beryllium fluoride glass is being studied. Five mole per cent of the uranium tetrafluoride appears to be soluble in the beryllium fluoride. No glass was present in the one fused mixture of NaF-KF-ZrF,-UF, studied. L My &t FOR PERIOD ENDING MARCH 10, 1952 13. HEAT TRANSFER AND‘PHYSICAL PROPERTIES RESEARCH H. F. Poppendiek, Reactor Experimental Engineering Division Viscosities of NaF-KF-LiF eutectic have been determined over a wide temperature range by utilizing three types of viscometers that range from about 8 centipoises at 550°C to 3 centipoises at 800°C. The addition of ~up to 30 wt % of uranium tetrafluoride increased the viscosity of this fluoride from 5 to 7 centipoises at 700°C but only from 4.0 to 4.3 centi- poises at 800°C. Some preliminary thermal conductivity information on this coolant has also been obtained. Heat capacity determinations have been made for two different compositions of NaF-BeF,-UF, mixtures over wide temper- ature ranges. A table summarizing physical properties of materials of interest to ANP 1s included. A series of corrosion failures has made it impossible, for the present, to obtain fundamental heat transfer data for fused salts and hydroxides. ‘The heated-tube, lithium system has been completed and 1s to be used to determine heat transfer characteristics for this coolant in the near future. Some experimental boilingheat transfer data for a mercury system have been obtained. The mathematical analysis of circu- lating- fuel heat transfer for turbulent flow has been evaluated for a series of Reynold’'s and Prandtl’s moduli for the case of no wall heat transfer. Mathematical solutions for the natural convection of liquid fuel elements have been developed for turbulent flows. VISCOSITY OF FLUORIDE MIXTURES M. Tobias S. I. Kaplan Reactor Experimental Engineering Division J. M. Cisar ANP Division F. A, Knox F. Kertesz Materials Chemistry Division The viscosity of the various fluoride fuels proposed for the circulating- fuel reactor is of great importance, since a low value (less than 10 centi- poises at 800°C) 1is required. The mixture NaF-KF-LiF-UF, appears to meet this requirement, whereas NaF-KF-UF, does not. Viscosity of NaF-KF-LiF. The viscosity of the NaF-KF-LiF eutectic (11.5 mole % NaF, 42.0 mole % KF, 46.5 mole % LiF) was measured by using three different viscometers, a modified Brookfield viscometer, an efflux unit, and the previously described falling- ball instrument.(!? The data ob- tained(2+3) range from about 3 * 1 centipoises at 550°C to 3 + 1 centi- poises at 800°C. (I)F. A. Knox and F. Kertesz, *“Brookfield Viscometer,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1957, ORNL-1154, p. 136. (2)y, Tobias, Measurements of the Viscosity of Flinak, Y-F30-6, Feb. 26, 1952. (B)S. I. Kaplan, Viscosity Measurements of Flinak by the Falling-Ball Viscometer (to be issued), 153 Lt idiaat TERT O ) e i / ANP DIVISION QUARTERLY PROGRESS REPORT Viscosity'of NaF—KF-LiF-UF4. The previously described(*) apparatus con- sisting of a modified Brookfield viscometer and a controlled atmosphere furnace with certain improvements to guard against hydrolysis and oxidi- ‘zation was used during the past quarter to determine the viscosity of NaF-KF- LiF-UF,. Increasing the amount of uranium fluoride, up to 30wt %, failed to cause a substantial change in the viscosity, as shown in Takle 35. TABLE 35 Viscosity of NaF-KF-LiF-UF, MIXTURES as a Function of yYranium Tetrafluoride Concentration UF, CONCENTRATION*| VISCOSITY + 0.5 (centipoises) T (wt %) 700°C 800°C 0 5.2 4.0 2 5.2 4.1 15 | 6.0 4.3 30 7.2 4.3 *Solvent (mole %): 11.5, NaF; 4.2, KF; 46.5, LiF. Modifications ofViScosity*Apparatfis. . The efflux and Brookfield viscometers are being installed in a dry box filled with inert gas to study the viscosity of beryllium-containing fluoride mixtures after hydrofluorination to remove insoluble oxides formed by oxidation and hydrolysis. A new, nickel viscometer tube has been 1in- stalled in the falling-ball apparatus for further high-temperature salt work. (4)F. A. Knox and F. Kertesz, ‘*Viscosity of Fluoride Mixtures,” Aircraft Nuclear Propulsion . Project Quarterly Progress Report for Period Ending December 10, 1951, OBRNL-1170, p. 126. 154 The length of the measured fall path has been increased by about one-fourth and the amount of material needed decreased by one-third, THERMAL CONDUCTIVITY OF LIQUIDS AND SOLIDS L. Cooper M. Tobias W. D. Powers S. J. Claiborne Reactor Experimental Engineering Division The thermal conductivity apparatus was further checked by calibration with a molten metal. The value of the thermal conductivity of lead obtained with this device agreed within 13% of the value given in the literature. Experiments to study Flinak (LiF-NaF- KF eutectic) were initiated; however, at the end of one run a corrosion failure (attack on a weld and a stain- less steel bellows) halted the experi- mental work. The preliminary results of this single run indicated that the thermal conductivity of Flinak was severalfold greater than that of the fluoride heavily laden with uranium (46.5 mole % NaF, 26.0 mole % KF, and 27.5 mole % UF,) that had been studied previously. Modification of the system will be attempted so that corrosion will be minimized, Upon the completion of this modification, further thermal conductivity measure- ments of the fluorides will be made. A new, longitudinal flow apparatus, originally developed to investigate solids, has been erected. This device will now be used for studing ligquids with high thermal conductivities. Tt is believed that the minor free- convection currents that might occur in this system will not significantly influence the heat transfer (conduction being the important heat transfer mode in this case). # HEAT CAPACITIES ‘W. D. Powers R. M. Burnett G. C. Blalock Reactor Experimental Engineering Division The enthalpies and heat capacities of the following salt mixtures have been determined by the use of Bunsen ice calorimeters. For the 76 mole % NaF-12 mole % BeF,-12 mole % UF, mixture at 250 to 465 C, Hy (s) - Hyo( (s) = 0.22T - § (cal/g) ¢p = 0.22 £ 0.02 (cal/g°°C) , ‘and at 520 to 1000°C, Hy (1) ~Hyop (s) = 0.32T - 35 (cal/g) , c, = 0.32 £ 0.03 (cal/g °C) . For the 25 mole % NaF- 60 mole % BeF - 15 mole %UF4 mixture at 280 to 1000° C HT - HO,OC = 0.32T - 45 (cal/g) ; 4~ ¢, = 0.32 + 0.02 (cal/g-°C) . The enthalpy temperature curve of the second salt mixture exhibited no dis- contlnultles and thus 1ndlcated the character1stlcs of a glass. At present the heat capac1t1es of Inconel potass1um hydroxide, barium fhydroxlde, and a lithium-sodium potassium- fluorlde mixture are belng determined. 'VAPOR PRESSURE_OF LIQUID FUELS R. E. Moore Materials Chemistry Division Previous determinations of the vapor pressure of uranium tetrafluoride FOR PERIOD ENDING MARCH 10, 1952 are in disagreement; therefore new measurements were made between 1037 (slightly above the melting point) and 1185°C., The apparatus and pro- cedure employed were described previ- ously.¢%*6) The data may be repre- sented by the equation log,, P(mm Hg) = (-9130.5/T) + 7.774 , which gives a maximum variation from the measured values of 4.6%. The heat of vaporization, as calculated from the equation, is 42.1 kg-cal/mole. In earlier work Johnson(7) used a static method for determining the vapor pressure of solid uranium tetrafluoride, and Ryon and Twitchell(8) measured a series of boiling points at reduced pressures., The highest vapor pressure obtained with the solaid was approximately 6.5 mm at 1000°C, but in the present work this pressure "was observed at about 1040°C. The boiling-point method gave a much higher heat of vaporization than that calculated from the data in this report; that is, the pressures ob- tained at higher temperatures were considerably higher, and those obtained at lower temperatures considerably - lower, than the values reported herein, The difficulty of obtaining reliable boiling points in the low pressure range and the fact that the points were scattered m1ght account for some of the differences. Vapor preéstres of sodium fluoride and potassium fluoride, the other two components of the NaF-KF-UF4'eutectic (S)R. E. Moore and C. J. Barton, “Vapor Pressure,” op. cit., ORNL-1154, p. 1317. (S)R. E. Moore, “Vapor Pressure of Ligquid Fuels,” op. cit., ORNL-1170, p. 126. (7)K. 0. Johnsson, The Vapor Pressure of Uranium Tetrafluoride, Y-42, Oct. 20, 1947. (B)A. D. Ryon and L. P. Twichell, Vapor Pressure and Related Physical Constants of lIranium Tetrafluoride, H-5.385.2, July 25, 1947. 155 ANP DIVISION QUARTERLY PROGRESS REPORT mixture, have been determined by Wartenberg and Schulze.(®) The data for sodium fluoride and potassium fluoride, as well as the data reported here for uranium tetrafluoride, were extrapolated to 1267°C, the highest temperature at which the vapor pressure of the eutectic mixture was measured. The ideal partial pressures at 1267°C of each of the components of the mixture, as calculated by Raoult’'s law, show 6.4 mm for sodium fluoride, 31 mm for potassium fluoride, and 19 mm for uranium tetrafluoride, The ideal total pressure is therefore approxi- mately 56 mm, whereas the experimental vapor'pressure'at 1267°C is 11.6.¢5) Assoc1at10n in the mixture probably accounts for the large deviation from ideal behavior. A 11ke1y poss1b111ty is the presence of complex ions formed by association of fluoride ions and uranium tetrafluoride molecules. PHYSICAL PROPERTY DATA M. Tobias Beactor Experlmental Engineering D1v1s1on ' Summaries of the available data on the physical properties of fluoride salts and other reactor materials are "presented in Tables 36 and 37. Most ~of the data were obtained by the ‘physical properties group at ORNL; ‘however, pertinent physical property measurements obtained by other organi- zations are also included. The ORNL physical property measurements are being made for the purpose of quickly supplying the ANP project with data of ‘a reasonable accuracy. The general (9) Pressure of Some Salts. 568 (1921) v, Wartenberg and H. Schulze, “Vapor 11" Z. Elektrochem. 27, 156 . 5. Density: assigned to the data tables are noted as accuracies presented in the follows: 1. Melting point: within about +10°C. 2. Heat Capacity: within about *10%. 3. Thermal conductivity: preliminary checks on the thermal conductivity device for liquids, made by using molten lead 1indicated that the data fell within about +30% of the known values. An error analysis, assum- ing pessimistic Chromel-Alumel thermocouple deviations, suggested that the NaF-KF-UF, data were accurate to within about +30%. Further checks are belng made 4. Viscosity: the results so far are preliminary. A definite accuracy cannot yet be assigned. within about +5%. Only the existence of preliminary physical property data has been noted on Tables 36 and 37; when the results have been confirmed, the data will be listed. NATURAL CONVECTION IN CONFINED SPACES WITH INTERNAL HEAT GENERATION D. C. Hamilton L. Palmer F. E. Lynch R. F. Redmond Reactor Experimental Engineering Division Analytical solutions have been developed for natural convection systems in which the heat source and the wall flux are uniform and the aspect ratio is very high (i.e., L/d very large). Three memorandums have been published regarding the laminar LST wy TABLE 36 Physical Properties of Fluoride Salts APPROXIMATE (¢} MELTING POINT HEAT CAPACITY THERMAL CONDUCTIVITY VISCOSITY DENSITY VOLUME EXPANSION MOLE % (°C) {cal/°C-g} {Btu/hr- ft2-°F/ft) (centipoise) (g/cc) COEFFICIENT NaF-BeF,-UF, 76, 12, 12 480 Solid, 0.22 at 250°C < T < 465°C; liquid, 0.32 at 520°C < T < 1000°Ct®) NaF-PeF,-UF, 25, 60, 15 635 0.32 at 280°C < T < 1000°C ‘*! {starts solidifying) NaF-KF-UF, 46.5, 26, 2.5 530 Solid, 0.15 at 240°C < T < 535°C, 0.53 at 550°C < T < 750°C'® 470 - 1.15x 1073 T 2.82 x 1074 at 535°C (eutectic) tiquid, 0.23 at 535°C < T < at $35°C < T < 1000°C )| 3.24 x 10"% ac 1600°C 1000°Cte? NaF-KF-UF, 48.2, 26.8, 25 558 4,54 - L1 x10°3 T at 2.84 x 104 at 600°C 600°C < T < 900°C N 3.10 x 10°% at 900°C NaF-KF-LiF 11.5, 42, 46.5 455 Preliminary results Preliminary 2.39 - 5,9 %x 1004 T at 2.84 x 10"% at 530°C (Flinak) results: 8 at | 530°C < T < 850°C'#8} 3.12 x 10°% at 850°C 550°C to 3 at 800°C UF,-LiF-NaF-KF | 1.1, 44.5, 10.9, 43.5 455 2,65 - 9.0 x 104 T at 4.15 x 1074 at 530°C (Fulinak) : 530°C < T < gso°c'™ 4.79 x 10-4 at 850°C UF, 103541 0.073 + 6.1 x 1075 T at 280°C < T < 950°C'*! (G)C. J, Barton, ORNL Materials Chemistry Division, personal communication. (6)y, D, Powers, ORNL CF-51-11-195, Nov. 30, 1951. (C)w . D. Povers, ORNL CF-51-9-64, Sept. 13, 1951. (d)1, Basel and M. Tobias, ORNL CF-51-7-169, July 31, 1951. (e)s, 1. Kaplan, ORNL CF-51-8-97, Aug. 13, 1951. (f)3. Cisar, ORNL CF-51-11-78, Nov. 14, 1951 (8)5. Cisar, ORNL CF-51-12-91, Dec. 14, 1951, (h)5, Cisar, ORNL CF-51-11-198, Nov. 30, 1951. (i), o D. Ryon and L. P. Twitchell, H-5,385.2, July 25, 1947, ‘0T HOUYW ONIGNA aoIydd dod cS6tl 8S1 ol gl TABLE 37 * Physical Propertiés of Miscellanéous Materials Type-316 stainless steel 1370 to 1400 (k) <1< 980°C Solid, 0,1093 + 5.66 x 10°5 T at 150°C < T < 1000°C \* 32 at 300°C 9.02 at 100°C (A) 12. 4 at 500°C APPROXIMATE - MELTING POINT HEAT CAPACITY THERMAL CONDUCTIVITY VISCOSITY (°c) (cal/°C*g) (Btu/hr*ft2+°F/f¢) (centipoise) " Sodium hydroxide 323(a) Liquid, 0.49 at 340°C < T < 990°C " | 0.81 at 520°C(¢) 4.0 at 350°c(4) : 2.2 at 450°C 1.5 at 550°C _ | e 1.0 at 650°C Lead-bismuth alloy 125(¢) Liquid, 0.035 at 175°C < T < 1000°C | 5.32 at 160°C ¢ (55.3 mole % Bi) 6.53 at 320°C Nickel 1452 F) Solid, 0.12 + 3.1 x 10" T at 240°C | 36 at ¢°c(8) Li thium 186 {f) Liquid, 1.001 + 2.76 x 10°° T at 20.6 to 26.6 at m.p.'"’ 0.60 at 205°C (<) 250°C < T < 1000°C 0.41 at 1000°C Zirconium 1700 (£ Solid, 0.0697 + 3.62 x 10°5 T at 11.3 at room temp. /) 150°C < T < 1050°¢ (i) 11.2 at 200°C 10.7 at 500°C Mol ybdenum 2620 ) Solid, 0.0675 at 150°C < T < 1050°C‘*?| 70.3 at 540°C (%) 62.9 at 870°C 56.5 at 1145°C (a) personal communication. C. J. Barton, ORNL Materials Chemistry Division, (f)Chemical Engineer’s Handbook, 2d, ed., p. 313-367. €)rbid., p. 949. (b) . W. D. Powers, ORNL CF-51-11-195, Nov. 30, 1951. (h)United States Steel Corporation, Fabricetion of U.8.8. . ST 5 (C)H. R. Deem, Battelle Memorial Institite, personal Stainless and Hgat Resisting Steels, 1950 ed. communication. . (i)w. D. Powers, ORNL-1154, p. 134. (PH. R. Stephen, NEPA IC-50-4-20, Apr. 10, 1950. “7)G. Bing, F. W. Fink and H. B. Thompson, BMI-65, Apr. 16, g ‘¢)iiquid Metals Handbook, p. 31, June 1, 19%0. 1951 (E)E. Mikol, ORNL-1131, Feb. 14, 1952. LH0d3d SSIYI0Ud ATHALHUVAD NOISIAIG dNV * i &) flow case,(10+11.123 Ope analysis in- volves a parallel plane system and consists of solving the heat con- duction equation by using a simplified velocity distribution; a similar analysis has been made for a circular pipe system; and a third analysis con- sists of solving the hydrodynamic and heat flow equations simultaneously, The temperature distributions and critical Reynold’s moduli for the third analysis, which was more exact than the first, were almost 1dentical. The following comments apply only to the ideal systems treated in these memorandums. JIn small tubes (0.2 1in. in diameter) a temperature reduction as a result of natural convection does not seem to occur for practical values of the variables. The ratio of the difference in center line wall temper- ature and the temperature when con- duction is the only mechanism present 1s represented by ¢ . For laminar flow in very large tubes, @, may be as small as 0.3. In the case of turbulent flow, ¢, is smaller by an order of magnitude, To attain turbulent flow and the resulting low values of ¢0 in small tubes, large temperature dif- ferences are necessary. It should be noted, however, that systems using large-diameter tubes would yield turbulent flow with much smaller temperature differences. An apparatus is being constructed to obtain an over-all temperature difference for a cylindrical system in (10)yy . Hamilton, H. F. Poppendiek, and L. D. Palmer, Theoretical and Experimental Analyses of Natural Convection Within Fluids in which Heat is Being Generated, Part I, ORNL CF-51-12-70, Dec., 18, 1951. (Il)From A. Simon to H. E. Sterm, Letter Regarding Agreement on Tower Calculations, ORNL CF-52-1-1, Jan. 2, 1952. (12)D. C. Hamilton, R. F. Redmond, and L. D. Palmer, Theoretical and Experimental Analyses of Natural Convection Within Fluids in which Heat 1is Being Generated, Part III, ORNL CF-52-1-2, Jan, 11, 1952, FOR PERIOD ENDING MARCH 10, 1952 the turbulent flow region., The new design will facilitate the measure- ment of wall temperature. Mercury will be used as the heat generation medium, ' A flat plate apparatus 1s being designed to permit visual study of the velocity distribution in the laminar region, (13> A dilute sulfuric acid solution will be used as the heat generation medium, ANALYSIS OF HEAT TRANSFER IN A CIRCULATING-FUEL SYSTEM H. F. Poppendiek L. Palmer Reactor Experimental Engineering Division Mathematical solutions for laminar and turbulent heat transfer in the circulating-fuel systems were described in a previous report,(!*) The wall- mixed mean fluid temperature difference in the established flow region for laminar flow 1is Gr02 [llF ‘- 8] t -t = —— w mm k 48 where G = volume heat source, ro - pipe radius, k = fluid thermal conductivity, 2 d Gr, (d4], dq : — = pipe wall heat transfer rate. dA}, {13)R. F. Redmond, Theoretical and Experimental Analyses of Natural Conwvection Within Fluids in wvhich Heat ts Being Generated, Part V, ORNL CF-52-1-5, Feb, 12, 1952. (14)H. F. Poppendiek and L. Palmer, Forced Convection Heat Trensfer in a Pipe System with Volume Heat Sources Within the Fluids, Y-F30-3, Nov., 20, 1951. 159 T Lo ~ ANP DIVISION QUARTERLY PROGRESS REPORT The wall-mixed mean fluid temperature ~difference for turbulent flow is found to be a function of Reynold’s modulus (Re), Prandtl’s modulus (Pr), the volume heat source, the fluid thermal 2x1073 10 10-2 10° 104 conductivity, the pipe radius, and the function F., The turbulent solution for the case of no wall heat transfer (F = 1) has been evaluated for a range of Re and Pr and graphed in Fig. 48. UNCLASSIFIED DWG. 14447 105 10° Re AFig. 481 Dimensionless, Wall-Mixed Mean Temperatureflnifferéncé assiFunction of Reynold’s and Prandtl’s Moduli for a Heat Transfer System with Insulated Pipe Walls, 160 L s, O ‘the Metalloy Corporation. These solutions for laminar and tur- bulent flows can be used in designing circulating-fuel systems, HEAT TRANSFER COEFFICIENTS Heat Transfer in Molten Lithium (H. C. Claiborne and G. M. Winn, Reactor Experimental Engineering Division). The apparatus for measuring heat transfer coefficients with an electrically heated tube was completed. Internal cleaning-and degreasing were accomplished by thorough flushing with ethanol and trichlorcethylene. The system was outgassed by heating and repeated flushing with argon scrubbed with lithium and eutectic sodium- "potassium alloy to remove the oxygen The flush- ing procedure consisted of raising the pressure to about 35 psig with scrubbed argon and bleeding the pressure down to about 2 psig. After filling with lithium, the system, when not in operation, was kept under 35 psig with scrubbed argon to prevent air con- tamination, and nitrogen contaminants, Sixteen pounds of lithium was added to the system through the filter in 4-1b batches. The lithium used was the low-sodium grade purchased from To keep contamination down to a minimum, the film of nitrides and oxides adhering to the 1-1b cylinders of solid lithium was shaved off and the solid metal soaked 1in trichloroethylene. This operation was done in a dry box filled with argon. The lithium was success- fully pumped around the system (via the test section by-pass) by the two electromagnetic pumps in series, The electromagnetic flow meter was cali- brated by using the drain tank that had been previously calibrated with water, ~ After the system was prepared for obtaining heat transfer data, it was FOR PERIOD ENDING MARCH 10, 1952 found that an electrical current could not be passed through the test section, Examination revealed that the test section had melted at one point,. Apparently too much current was passed through the section while testing the transformer circuit. A new test section is being fabricated and other minor changes are being made. Another attempt to get heat transfer data will be made as soon as this work 1is complete. Heat Transfer to Fused Salts and Hydroxides (H. W. Hoffman andJ. Lones, Reactor Experimental Engineering Division). Flinak was removed from the heat transfer system and replaced with sodium hydroxide during the last quarter., Corrosion by the Flinak had caused leaks in several welds and in one of the mixing pots, and'it was necessary to replace these parts of the system., A redesigned mixing pot consisting of a section of 1-in. nickel pipe 2% 1in, long and capped at both ends was installed. The fluid enters at one end of the pot tangentially to the inside surface of the pipe, passes through a perforated nickel disk, and leaves at the bottom. A thermocouple is located in front of the exit. Entrance Region Heat Transfer in a Sodium System (W. B, Harrison, Reactor Experimental FEngineering Division), The modifications and additions to the experimental system proposed in the last quarterly report{(13) have been made, and preparations are being made for loading the system with sodium, Extremely high values of the heat transfer coefficient should be achieved with the use of sodium in an entrance region (e.g., 200,000 Btu/hr ft2:°F); therefore good wetting of the copper (IS)W. B. Harrison, ““Entrance-Region Heat Transfer in a Sodium System,” op. cit., OBRNL-1170, p. 118. . 161 - ANP DIVISION QUARTERLY PROGRESS REPORT test section by the sodium is of con- siderable importance. There are no sodium wetting data available for copper, but the data for similar metals indicate that copper may be well wetted by sodium even at the low operating temperatures (up to 300°F). A test will be conducted to determine the degree of wetting and possible ways of improving it 1f it should be poor. The study will consist of measuring relative electrical re- sistances across a few copper-sodium interfaces as functions of temperature, The interfaces proposed at present are sodium in contact with copper that has been silver plated, mercury plated, and oxidized. The equipment has been assembled for this experiment, and data should be available in the near future. Work on the heat transfer system 1is being deferred in order to ‘capitalize on any positive results of the wetting studies, HEAT TRANSFER OF BOILING *_ LIQUID METALS W. S. Farmer Reactor Experimental Engineering Division ~ The previous boiling-mercury data for the horizontal plate system(16) have been analyzed and compared with existing information in the literature on nucleate boiling and free con- vection, A plot of this data is shown in Fig. 49. The Nusselt numbers (hL/kR) obtained experimentally fall below the values predicted by cor- relations of Insinger and Bliss(17) for nucleate boiling and Jakob¢(!¥®) for free convection, Over most of the (16)“’. S. Farmer, “Heat Transfer in Boiling- Liquid-Metal Systems,' op. c¢it,, ORNL-1154, p. 138. (17)T. H. Insinger, Jr. and H. Bliss, “Trans- mission of Heat to Boiling Liquids,”” Trans. A4m. Inst. Chem, Engrs. 38, 491 (19490). (IB)M. Jakeb, Heat Transfer, I, 640, Wiley, New York (1949). 162 temperature difference range the experimental values are approximately one-fourth the predicted values, Existing correlations may not correctly take into account Prandtl’s number and wetting, which may be of importance in liquid metal beoiling. UNCLASSIFIED DWG. 14448 o o S MERCURY TEMPERATURE, 340° F 500 300 @ ® A4Jé WHERE 7¢=METAL SURFACE TEMPERATURE 7, =BULK LIQUID TEMPERATURE HEAT TRANSFER COEFFICIENT, 4 (Bfu/hr'ffz-oF) [e]e; 20 50 100 200 o T,-7, {°F) Fig. 49. Heat Transfer Coefficients for Boiling Mercury. Minor changes are to be made in the horizontal flat plate system in order to evaluate wetting characteristics and heat transfer surface resistances. In addition, a small, inexpensive, mercury circulation unit is to be con- structed. Mercury will then be used as the condenser coolant for experi- ments in which the heat transfer coefficients of boiling sodium will be measured, The experimental apparatus for in- vestigating heat transfer coefficients of boiling liquid metals by using a horizontal tube geometry was completed this quarter. Preliminary tests using water as a boiling fluid are now in progress. ' FOR PERIOD ENDING MARCH 10, 1952 14. RADIATION DAMAGE D. S. Billington, Solid State Divisien A, J. Miller, ANP Division Studies have continued on the the effect of radiation on the sta- bility of airplane reactor constitu- ents. Considerable data have been obtained from experiments in the X-10 graphite pile, the higher flux LITR, and the Y-12 éyclotron. In all phases of the work, preparations are being "made to carry out additional studies in the LITR and to utilize the MTR facility when it becomes available. This work has been concerned largely with the stability of fused fluoride "salt mixtures of the type proposed as fuel for circulating fuel reactors. In several of the inpile experiments with the salts in Inconel capsules, the rate of attack on the container material has been considerably higher than that observed in the out-of-pile controls, However, the evidence is not conclusive that the increased corrosion rate is caused by the radia- tion field. Other radiation damage studies have included inpile liquid metal loop experiments and creep of metal. So- dium was circulated in the 1inpile loop for 165 hr, during which time radioactive decay curves were taken. The thermal condictivity of metals under irradiation has also been studied. The earlier creepand thermal conductivity data have been partially substantiated by recent tests. One such creep test with nickel indicates that the inpile creep rate becomes higher than its bench counterpart after about 115 hours. Additional details on radiation damage studies are con- tained in a quarterly report of the Physics of Solids Institute. (!’ (I)Physics of Solia’s'Institute Quarterly Progress Report for Period Enrnding January 31, 1952, ORNL-1261 {in press}. IRRADIATION OF FUSED MATERIALS G. W. Keilholtz Materials Chemistry Division The effects of radiation on the stability of fused fluorides in con- tact with Inconel have been investi- gated by use of the X-10 graphite pile, the LITR, and the Y-12 cyclotron. Reasonably detailed inspection of materials from experiments carried out in the X-10 graphite pile and cyclo- tron at 1500°F showed no positive evidence of fuel decomposition or increased corrosion of the Inconel when compared to out-of-pile control tests. However, two fused fluoride fuels irradiated in the LITR at power densities of 800 and 84 watts/cc in approximately 1/8-in.-ID tubes have exhibited increases in the corrosion rate as indicated by analyses of the fuel for corrosion products. One of these has the composition NaF-KF-UF,, 46.5, 26, and 27.5 mole %, respec- tively, in which the power dissipated when irradiated in the LITR is about 800 watts/cc of material. The second fuel has a composition NaF-BeF,-UF,, 47, 51, and 2 mole %, respectively, and a power dissipation of about 84 watts/cc. This is in the power range of that occurring in the large-diameter tubes of the proposed ARE. 1In both cases there 1s some question as to whether the inpile tests are being carried out in a manner sufficiently isothermal to prevent increased cor- rosion rates by thermal gradients in the capsules. In the case of the low-uranium-concentration fluoride mixture, irradiated at 84 watts/ce, analytical results are available from only one experiment. These results are 1n sharp contrast to those from 163 B ialh s ok roriat Lo WP T ANP DIVISION QUARTERLY PROGRESS REPORT an experiment on a similar beryllium type of fuel containing a higher per- centage of uranium, which had a power dissipation of 554 watts/cc. The latter experiment showed no evidence " of increased corrosion, on the basis " of an impurity analysis of the fuel, as compared with the control. Preparations are being made to carry out studies on the escape of xenon from the melt in the X-10 graph- ite pile and on fuel stability in the MTR. Pile Irradiation of Fuel (J. G. Morgan, P. R. Klein, C. C. Webster, - B. W. Kinyon, M. J. Feldman, H. E. Robertson, Solid State Division). Experiments during the past quarter have been conducted entirely on fused fluoride fuels in the LITR. ‘Results of these experiments as com- pared with the control runs are shown in Table 38. The corrosion of the Inconel in some LITR runs is not only deeper and more general than that in the controls, but there are also two kinds present - the intergranular-type found in the control runs and a type of corrosion that is confined within the grain and shows evidence of being preferential in the grains attacked. The capsules are considerably cor- roded above the point where the liquid level is assumed to be. There 1is a small amount of such corrosion in the capsules irradiated in the X-10 graph- ite pile but little evidence of cor- rosion of Inconel in contact with the vapor phase in control runs. Chemical analyses in Table 38 indicate that less chromium is dissolved in the irradiated fuel than in the control TABLE 38 " LITR Tests on Fused Fluoride Fuels in Inconel at“1500°F ‘ c : , INCONEL. COMPONENTS IN 'FUEL COMPOSITION | TIME | IRRADTATION FUEL AFTER TEST ~ {mole %) (hr) (watts/cc) Ni Cr Fe CAPSULE CONDITION 46.5 NaF- 115 800 26,878 | 1878 | 16,329 | Generally corroded to 26 KF-27.5 UF, ' depth of 2 to 3 mils o 115 Control 1,034 | 754 | 1,270 | Occasional 1 to 2 mil , intergranular attack 161 800 45,169 950 6,414 Generally corroded to depth of 2 to 3 mils 161 Control 1,100 645 2,655 | Occasional 1 to 2 mil intergranular attack 136 800 1,380 160 1,220 No corrosion* 47 NaF-~ 143 84 80,000 | 2450 3,100 Inspection incomplete 51 BeF,-2 UF, T 136 Contrel 1,123 | 3300 811 Inspection incomplete 95 NaF- 139 554 1,490 | 1060 | 1,200 | Inspection incomplete 60 BeF,-15 UF, 131 Control 1,540 | 7380 | 2,300 Inspection incomplete *At 824°F (solid) 164 .y samples but that there 1s a large increase in the nickel and iron con- tent of the irradiated fuel compared with that in the control. ' Cyclotron Irradiation of Fuel (W. J. Sturm and M. J. Feldman, Solid State Division, R. J. Jones and R. L. Knight, Electromagnetic Research Division). Fuel irradiations in the cyclotron with 20-Mev protons described in the previous quarterly report¢?’ were carried.out at 30 to 415 watts/cc of fuel for a short period of time, usually an hour. During the past quarter smaller Inconel capsules have been developed toavoid terminations of runs because of uneven temperatures. A helium cooling system has been de- signed and is being constructed to allow higher power dissipations. With the Inconel microcapsules i1t was possible to bombard a lithium-bearing fuel - KF-NaF-LiF-UF,, 43.5, 10.9, 44.5, and 1.1 mole %, respectively - up to 8 hr with 100 to 400 watts/cc. Fuel analyses and metallographic ex- aminations are being made. Additional fuel irradiations are being carried out by North American Aviation with the use of the 60-in. Berkeley cyclotron. INPILE CIRCULATING LOOPS 0. Sisman , C. Ellis W. W. Parkinson W. E. Brundage A. S. Olson R. M. Carroll C. D. Baumann Solid State Division Sodium was circulated at a velocity of 1 ft/sec in the X-10 graphite pile through a loop of Inconel for 50 hr at 1000°F and 115 hr at 1500°F. An (2}W. J. Sturm, M. J. Feldman, R. J. Jones, J. 8. Luce, and C. L. Viar, “Cyclotron JIrradiation of Fuel and XOH Capsules,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 10, 1951, ORBNL-1170, p. 143. FOR PERIOD ENDING MARCH 10, 1952 electromagnetic pump was used, and the temperature of the material in the pump cell was 1000°F. The flow rate of sodium through the loop gradually decreased, and eventually the loop could not be operated. When the ac- tivity of the loop has decayed suf- ficiently, an examination will be made to determine the condition of the loop and the cause of the flow stoppage. At various times while the flow was stopped, radiocactive decay curves were obtained for the part of the loop outside the pile. The curves are being examined for long-lived corrosion products from the Inconel tube walls. A second sodium loop for the X-10 graphite pile and a sodium loop for operation in the LITR are being con- structed. A fused fluoride fuel loop for operation in the MIR is being de- signed. CREEP UNDER IRRADIATION J. C. Wilson J. C. Zukas W. W. Davis Solid State Division It was previously reported¢?®’ that a cantilever creep test at 1500°F and 1500 psi showed that X-10 graphite pile irradiation at a flux of 4 X 10'° fast neutrons/cm? caused an increase in total creep strain of about 20% in type-347 stainless steel after about 250 hr of exposure, which was the duration of the tests. Extrapolation of the bench and inpile curves to longer times indicated that the dif- ference between them increased with time. In order to obtain further information on this point, a 500-hr test was run in the pile with no (3)J. C. Wilson, J. C. Zukas, and W. W. Davis, “Creep Under Irradiation,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1951, ORNL-1154, p. 170. 165 | | strain measuring microformer (since in 500 hr the creep strain would ex- ceed the maximum microformer travel). The beam deflection was measured with cathetometer after withdrawal from the reactor. The strain in the bench test exceeded that of the inpile test by 5%. Since this is contrary to what would be expected from extrapolation of the earlier data, the experiment will be repeated. ‘One 1inpile test was run on an electrolytic nickel sample at 1300°F at a maximum fiber stress of 2000 ps1. The material was annealed at 1500°F for 5 hr following a cold reduction of 40%. The temperature of 1300°F was chosen as the maximum at which the grain structure was known to remain stable during the test. Figure 50 shows a plot of creep rate vs. time for both inpile and bench tests. In common with the earlier work on type- 347 stainless steel, the curves show that the inpile strain-time curve SESRE DWG. 14122R1 N\ \ . BENCH TEST ~ \ o—— IN/PILE TEST > - w CREEP RATE (pin./in./ hr) © ) ELECTROLYTIC-NICKEL,CANTILEVER CREEP SPECIMENS 2000-psi MAXIMUM FIBER STRESS 1300°F (ANNEALED 5 hr AT 1500°F AFTER 40% COLD REDUCTION . W 100 150 TIME (hr) Fig. 50. Comparison of Bench and Inpile Creep Rates of Nickel. 166 becomes more linear and shows a lesser strain but a higher rate than its bench counterpart after about 115 hours. At about 120 hr the inpile rig had reached an approximately con- stant strailn rate, whereas the rate in the bench test was still decreasing. RADIATION EFFECTS ON THERMAL CONDUCTIVITY A. F. Cohen L. C. Templeton Solid State Division In a preliminary relative thermal conductivity experiment on Inconel, previously reported,(4) a large de- crease in thermal conductivity was observed after three days of irradia- tion at approximately 1517°F in the X-10 graphite pile. To check this result, a carefully annealed Inconel specimen was irradiated at 482 and 1067°F and showed apparently no effects from their radiation. When raised to 1517°F there was again some apparent lowering of thermal conductivity, the cause of which has not yet been de- termined. In relative thermal con- ductivity tests of high-purity cobalt- free nickel in the X-10 graphite pile, no lowering has been observed during long periods at low tempera- tures and several weeks of testing 1in the region of 1500°F. An absolute thermal conductivity test is being run in the LITR at 1500°F on Inconel that has been heat-treated to pre- cipitate all the carbides that can be precipitated by thermal treatment. In an absolute thermal conductivity test on type-316 stainless steel in the fast flux of the LITR at 212 to 392°F the thermal conductivity was shown to be unaffected by the irradia- tion. (4)A. F. Cohen, ‘“Radiation Effects on Thermal Conductivity,” op., ctit., ORNL-1154, p. 171. AP - 1 A - e Wi e - - T A 0 g T m——— e e Ty v e e R ow "y Ty Sl ¢ 5 T 4. e 5 X i Tl e @’ SUMMARY AND The survey of the supercritical- water reactor by the Oak Ridge National Laboratory’s subcontractor, Nuclear Development Associates, Inc.,, is com- plete. NDA has concluded (sec. 15) that a reactor of this type may be developed for supersonic propulsion but that the reacteor i1s an 1intricate machine and will present a difficult design job. Although OBNL will not pursue this reactor cycle because of the prevailing belief in the greater potentialities of low-pressure liquid- coolant reactors, NDA will continue its work, together with Pratt and Whitney, under direct AEC contract, The large, analytical chemistry program required in support of the materials research program included the routine analysis of 574 samples and the development of new analytical INTRODUCTION procedures where necessary. This development (sec. 16) 1is largely con- cerned with the analysis of fluoride mixtures for corrosion products and changes in composition, The "List of Reports Issued" (sec. 17) includes formal reports and informal documents on all phases of the ANP Project. A directory of the research projects of the Aircraft Nuclear Propulsion Project of the Oak Ridge National l.aboratory 1is given in sec., 18. The research projects of the Laboratory’s subcontractors on the ANP Project are listed, as well as the research now 1in progress at the Laboratory. The research projects being performed by ORNL for the ANP programs of other organizations are included and marked as such, 169 I e e o B i ot it i et b i o b b i R B M ik il bt el B b o e ook R M et i Tt TR AT T e i oo B . bk AR 15. THE SUPERCRITICAL-WATER REACTOR Nuclear Development Associates, Inc. The supercritical-water cycle, first proposed in Report Wash-24,(1) has been under continuous 1investigation by the Nuclear Development Associates, Inc., for over a year. completed their analysis of the super- critical-water reactor for ORNL and a final report has been written. (%) This report concludes the subcontract work by NDA for ORNL on the super- critical-water reactor. NDA will, ‘however, continue its work in this field in conjunction with Pratt and Whitney Aircraft Division in a study of the engine and hardware aspects of this system. The conclusions and recommendations in the final NDA report are quoted in detail below. This study was concerned only with the reactor and shield; the power plant and engine aspects of the proposal are outside the scope of the present summary. : ‘DESCRIPTION OF REACTOR The reactor consists of a structure of stainless steel plates immersed in a vessel of water that is above the critical pressure, so that it can be heated to high temperatures without - phase change. The plates contain the fuel and provide surfaces for trans- ferring the heat to the water. The water serves several functions: it is the coolant that carries away the heat from the plates; it is the moder- ator that slows down the neutrons in the chain reaction; through variations in its density 1t controls the re- activity of the machine; it is the neutron reflector that surrounds the "reactor core; and it is the innermost portion of the reactor shield. (I)Application of a Water Cooled and Moderated Reactor to Aircraft Propulsion, AEC Reactor Development Division, Wash-24, Auwg. 18, 1950. f2)The Supercritical Water Reactor, Nuclear Development Associates, Inc., ORNL-1177, Feb. 1, 1952. NDA has now CONCLUSIONS OF THE NDA STUDY It is the conclusion of NDA that it is possible to develop a reactor of this type but that it is an intri- cate machine and will present a dif- ficult design job. NDA does not be- lieve that this reactor represents an easy short cut on the difficult road to supersonic flight but that it may offer one not-impossible path to that goal. It is further stated that this study represents only an early step toward evolving such a reactecr and definitely does not provide a pre- liminary design ready for detailing. An attempt is made to display the potentialities of the machine and to set forth at least some of the prob- lems - the study has not solved the problems. The impressions that have been gained 1in regard to particular items are: 1. The heat output (400,000 kw) re- ported¢!’ appears attainable with a reactor having a modest size core (2.5-ft-square cylinder) and a reasonable fuel inventory (20 kg). 2. The present estimate (85,000 I1b) for the weight of a divided shield with a crew compartment of 1in- termediate size compares favorably with the allowance (90,000 1b) made in the report, or with the somewhat higher weights allowed for in the preliminary Boeing studies. 3. This shield, like other divided shields, does not permit normal approach to the airplane after it has landed. Additional awkward shielding provisions are needed for ground handling. 4. In view of industrial experience with high pressures, the pressure 171 T W T P Tt T T ANP DIVISION QUARTERLY PROGRESS REPORT (5000 psi) called for in this re-. actor does not 1in 1itself appear to constitute a major difficulty. Departure from standard pressure vessel practice will be neces- sitated by the low-weight require- ment in the present application; “although this will pose important design problems, it 1is thought that they can be worked out satis- factorily. There is considerable experience with water and stainless steel, ahd, at least in the absence of radiation effects, the use of these materials appears to be quite promising. A type of fuel element filate under development at ORNL appears promis- ing for use in this reactor. Incorporation of the basic fuel plates into satisfactorily cooled assemblies is a complex problem that will require major effort. The reactor calls for a very fine- scale fuel-bearing structure worked at tremendous heat load. A particular difficulty is that of maintaining equal temperatures in parallel cooling streams be- cause of their sensitivity (aris- ing from the large expansion of the fluid) to differences in heat load or other quantities. This tends to increase the amount by which the maximum wall temperature " exceeds the exit mixed temperature 172 of the coolant. Thus, the im- portant temperatures are sensi- tive to complex details of the reactor power pattern and to 1im- perfections in the design and fabrication of the machine. Variations in water density pro- vide a substantial amount of self-regulation, as well as a convenient mechanism for slow 10. 11. external control of the reactor. The machine appears quite amenable to control under steady condi- tions. A detailed study of startup pro- cedure, which would also involve the power plant, has not been made. However, it appears pos- sible, with only density controls, to start up the reactor when it is attached to a simple external system, - ' A useful amount of shim control can, in principle, be obtained by variation of water density 1in the reactor. However, the effects of density change upon the power pattern (cf. 8 above) have not vet been investigated, and 1t may prove desirable to incorporate mechanical or other slowly acting controls to assist 1in startup or in shim control. RECOMMENDATIONS OF THE NDA STUDY If the supercritical-water reactor is deemed sufficiently promising to be of further interest for aircraft or other applications, the following recommendations for further work are in order: : 1. Refinement of reactor and shield design generally, ‘ Development of fuel assembly de- sign and fabrication, Out—of~pilé work on heat transfer, corrosion, and fluid’f}ow, ' Inpile tests of corrosion, water decomposition, structural materi- als, and fuel element life in the relevant range of neutron fluxes, temperature, and pressure, Experimental and theoretical work on the géneral "flat flux" concept of reactor design, ' [ 3] Critical-assembly experiments for the supercritical-water reactor in particular, including such items as fuel content, lumping effects, power pattern, and the effects of water-density variation and nonuniformities, Semi mockup measurements of over- all shield effectiveness and of the radiation heat load in the neighborhood of the pressure shell, Re finement of stability, control and startup considerations, with FOR PERIOD ENDING MARCH 10,1952 attention to the coupling between reactor and power plant, Consideration of a reduced-power reactor experiment for studying the dynamics and control of startup, and also consideration of any other conceivable experiments that might shed light on these problems, e.g., the coupling (via a water- density sensing device) of an electrically heated supercritical- water system to a neutron simulator. 173 T 3 16 . ANALYTICAL CHEMISTRY C. D. Susano Analytical Chemistry Division The major portion of the effort in the field of Analytical Chemistry 1is concerned with the analysis of ternary and quaternary eutectics composed of alkali-metal, beryllium, and uranium fluorides that have been subjected to corrosion tests 1n stalinless steel and nickel alloy containers. Studies made during the past quarter led to revisions in the methods for the de- termination of uranium, iron, nickel, chromium, manganese, and molybdenum. The revised methods are yielding satis- factory results. The determinations of beryllium, total alkali metal, fluoride, and silicon present more serious problems, but progress 1is being made in the modification of existing methods or the development of new methods for determining these constituents. The feasibility of separating beryllium from uranium by the use of anion exchange resins is being studied, with very promising results. A method has been worked out for the determination of fluoride, which, it is believed, will show a marked improvement in accuracy and expenditure of time over the presently used pyrohydrolysis method. Efforts are being made to adapt the colori- metric silico-molybdate method for ‘the determination of silicon in fluo- ride eutectlcs Several further attempts to remove ‘boron from diatomaceous earth by washing with hydrochloric acid have shown that the boron content can be reduced to 50 ppm or less. : STUDIES OF DIATGMACEOUS EARTH J. C. Whlte W, J Boss Analytical Chemistry Division Several attempts to remove boron from one type of diatomaceous earth (with the trade name "Sil-0-Cel") by washing with hydrochloric acid in concentrations from 33 to 67% resulted in lowering the boron content from 300 ppm to 50 ppm or less. This in- vestigation has been hampered by the lack of a more sensitive method for the determination of boron in quanti- ties less than 50 ppm; therefore the ultimately possible reduction of the boron content is indeterminate at this time. A method for the determination of boron that involves the color complex between boron and 1,1'-dianthrimide in sulfuric acid(!’ is more sensitive than the spectrographic method being used, but its application to diato- maceous earth samples has not been successful, ANALYTICAL STUDIES OF FLUORIDE EUTECTICS J. C. White C. K. Talbott W. J. Ross C. M. Boyd Analytical Chemistry Division The major portion of the effort for the ANP program in the field of Ana- lytical Chemistry 1s concerned with the analysis of ternary and quaternary eutectics composed of alkali-metal, beryllium, and uranium fluorides that have been subjected to corrosion tests in stainless steel and nickel alloy containers. Determinations of uranium, beryllium, total alkali metals, and fluorides are made for the purpose of following possible changes in compo- sition of the eutectics, and detérmina- tions of impurities such as iron, (I)J. C. White and W. J. Ross, “Studies of Diatomaceous Earth,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 10, 1951, OBRNL-1170, p. 153. 175 T T T ANP DIVISION QUARTERLY PROGRESS REPORT nickel, chromium, manganese, and molyb- denum are required in order to de- termine the extent of corrosion of the metal container. Since the pres- ence of silicon in these mixtures has been detected by examination with . the mass spectrometer, methods of determining silicon are also being deve loped. The adaptation of existing methods of analysis to the fluoride eutectic type of sample has been the principal problem of the group. Uranium. The determination of uranium in fluoride eutectics is being satisfactorily accomplished by the potentiometric method of titrating uranium(IV) with ferric sulfate. (2’ Béryllium."The separationtxfberyl- lium from uranium is generally based on the selective precipitation of one component, such as the precipitation of beryllium hydroxide in carbonate solution. The determination of beryl- liam is ordinarily completed gravi- metrically by precipitation of the hydroxide, but a volumetric procedure ‘would be preferable. The feasibility of separating beryllium from uranium by the use of anion exchange resins is being studied. Both column and batch tests, with the use of Dowex-1 1in the chloride and sulfate forms, have in- dicated that uranium is adsorbed by such resins and that beryllium is not adsorbed. Such a separation will permit the direct titration of beryl- lium. Another method for the separation of uranium and beryllium, recommended by Fischer (reported by Rodden(?’), depends upon the extraction of the uranium salt of l-nitroso-2-naphthol with amyl alcohol. This method will be 1nvestigated. '(2)C. J. Rodden (Editor-in-chief}, Analytical Chemistry of the Manhattan Project, p. 70-71, . McGraw-Hill, New York (1950). 176 ‘Total Alkali Metals. In the pro- cedure now in use, all the cations except the alkali metal ions are pre- cipitated by NH,OH from a sulfate solution of the eutectic and evaporated, and the alkali metals are weighed as sulfates. Total Fluoride. The determination of total fluoride in the eutectics is accomplished by hydrolysis of the powdered eutectic in the presence of U;04 and steam at 1000°C, followed by the titration of the hydrogen fluoride contained in the distillate. The préblem of removing interfering ~ions 1s well on the way to being solved by a method that involves fusion of the material with sodium carbonate. Leaching the carbonate fusion mixture with water dissolves the fluorides, which can then be separated from the insoluble residue by filtration. Beryllium, uranium dioxide, and all the metal impurities except molybdenum, generally a very minor component, are retained in the residue. Fluoride 1is precipitated in the filtrate as lead chlorofluoride, which may be dried and weighed. Alternatively, a volumetric determination of chloride, after dis- solution of the precipitate in nitric’ acid, also yields the fluoride content. Recovery of fluoride was from 99.3 to 100.6% complete when the chloride concentration ranged from two to five times the theoretical amount. The volumetric method is less satisfactory, and further work is being done to correct its deficiencies. = The carbonate fusion method offers two main advantages over the hydrolysis procedure: (1) it is more adaptable to use on a large scale, 1in that 1t requires no special equipment and 1is more rapid, and (2) it permits the determination of a number of constitu- ents on the same sample; in fact, when necessary, all the desired constituents ~ e R i i i - can be determined on a single sample of the fused mixture. Nickel. The dimethylglyoxime method being used for the colorimetric de- termination of nickel employs special treatment for application to fluoride eutectics.¢® Chromium. The use of diphenyl- carbazide to form a violet complex with chromium(IV) for the colorimetric determination of chromium 1is well established; choice of a suitable oxidant, unfortunately, is not. Per- manganate has been used satisfactorily in this capacity, but sodium bismuthate has also been extensively studied. Manganese. The colorimetric method for the determination of manganese (*’ prescribes the use of sodium bismuthate as the oxidant to produce permanganate ions. This method, however, has proved unsatisfactory for the corro- sion-test samples currently being received, because they have been found to contain high chromium-to-manganese ratios. As an alternative, the periodate method has been studied and found to be more applicable, owing chiefly to the possibility of re- tarding the oxidation of chromium without affecting the oxidation rate of manganese through the control of the acidity of the system. Silicon. In the colorimetric de- termination of silicon, the complex molybdosilicic acid is extracted with either butyl or amyl alcohol, and the molybdenum blue complex is measured spectrophotometrically. Although this ‘method yields results that are some- times difficult to duplicate, it is (3)From J. C. White to L. J. Brady, Effect of pH and Concentration of Ammonium Persulfate on Colorimetric Determination of Nickel by the Dimethylglyoxime Method, Y-B31-324, Jan. 17, 1952. _ '(4)1. C. White, *Study and Analysis of Fluoride Eutectics,” Analytical Chemistry Division Quarterly Progress Report for Period Ending December 26, 1951, ORNL-1233, p. 62. FOR PERIOD ENDING MARCH 10,1952 widely used for the determination of soluble silicon in low concentration ranges. Dissolution of the fluoride eutectic to yield all the silicon in the form of soluble silica is a critical step. Hoffman and Lundell(®’ made a thorough study of the determination of silicon in the presence of fluoride and recom- mended sodium carbonate fusion of the solid. Small concentrations of sili- con can be leached out of the fusion mixture with water. The use of this method in preliminary tests 1indicated the presence of substantial quantities of silicon 1n certain fluoride eutec- tics. Aneffort 1s being made to adapt the method for application to the fluoride eutectics and to the indi- vidual components of the eutectics. SOLUBILITY OF BORIC ACID IN WATER A. P. Fraas ANP Division G. F. Petersen The solubility in water of the boric acid, H;BO;, is so low that the increase in the water density 1is slight. However, by adding 11.85 g of lithium hydroxide to 100 ml of water, 1t was possible to effectively dissolve 66 g of boric acid per 100 ml of water, with a resulting density of 1.2 g/cc. This method has been considered as a device for reducing shield weight. CLARITY OF BORATED WATER H. P. House Analytical Chemistry Division Little work was carried out on this project during the quarter, and no further work is planned because the use of borated water is no longer contemplated in the ARE. (5) ;. 1. Hoffmen and G. E. F. Lundell, “Determi- nation of Fluorine and of Silica in Glasses and Enamels Containing Fluorine,”” RP110, J. Research Nat. Bur. Standards 3, 583 {(June 1929),. 177 T AT EC T TR R T T TR BT IR T e ANP DIVISION QUARTERLY PROGRESS REPORT STUDIES OF ALKALI AND ALKALINE EARTH HYDROXIDE COOLANTS J. C. White Analytical Chemistry Division No work has been done on analytical methods applicable to alkali and alka- line earth hydroxide coolants this quarter because of the increased em- phasis on the study of fluoride eutectics. ' ANALYTICAL SERVICES H, P. House L. J. Brady : J. W. Robinson AnélYfidal Chemistry Division Analytical service work is now centralized in the ANP analytical Chemistry Laboratory in Building 178 9201-2. Over 70% of the samples. analyzed this quarter were from reactor chemistry studies of reactor fuels and coolants, Emphasis shifted from alkali and alkaline earth hydroxides to fluorides, uranium, alkali metals, and corrosion products in various fluoride salt mixtures. Other determi- nations, such as the purity of salt mixtures, oxygen in helium and argon, and 1ron in beryllium oxide, were also ‘made. A summary of service analyses performed is shown in Table 39. TABLE 39 Summary of Service Analyses Samples on hand 11/2/51 149 Number of samples recéived' 638 Total number of samples - 787 Number of samples reported 574 Backlog as of 2/2/52 213 AT A et e ey REPORT NO. Y-F15-10 ORNL-1177 OBNL- 1255 Y-F26-29 ANP-66 Y-F10-75 Y-F10-76 dRNL~1176 Y-F10-77 Y-F10-81 Y-F10-85 Y-F10-63 Y-F10-78 Y-F10-82 Y-F10-89 FOR PERIOD ENDING MARCH 10, 1952 17. LIST OF REPORTS ISSUED TITLE OF REPORT Design Design Study of Closely-Coupled Reactor- Heat-Exchanger-Shield Combinations for Use with a Fused Fluoride Circulating Fuel The Supercritical Water Reactor Basic Performance Characteristics of the Steam Turbine-Compressor-Jet Aircraft Propulsion Cycle Regarding Homogeneous Aircraft Reactors Reactor Physics Some Results of Criticality Calculations on BeQO and Be Moderated Reactors Some Results of Kinetic Studies on the ARE Design of 10 June 1951 The Circulating Fuel ARE Core Series of December 13, 1951 Criticality Calculations for Hydrogen Moderated Reactors from Microscopic Data Physics of the Aircraft Reactor Experiment Statics of the ANP Reactor - A Preliminary Report Effect of Potassium ih fihe Fue]-Coolafit Solution in Two ARE Reactors A Flux Transient Due To a Positive Reactivity Coefficient Water Moderated Reactors The ARE with Circulating Fuel-Coolant Effect of Structure on Criticality of the ARE of January 22, 1952 AUTHOR(S) DATE ISSUED {(to be issued) A. P. Fraas G. F. Wislicenus 2-1-52 Nuclear Development Associates, Inc. {to be issued) © H. Cohen A. P. Fraas W. B. Cottrell 1-29-52 C. B. Mills J. W. Webster 10-15-51 0. A. Schulze - J. W. Webster 10-19-51 M. J. Nielsen C. B. Mills 12-27-51 M. Shapior 11-29-51 S. Preiser C. B. Mills 1-5-52 C. B. Mills 1-8-52 C. B. Mills 1-21-52 C. B. Mills 1-14-52 C. B. Mills 1-7-52 C. B. Mills 1-11-52 C. B. Mills 1-29-52 179 ANP DIVISION QUARTERLY PROGRESS REPORT REPORT NO. Y-F10-79 CF-52-1-2 CF-52-1-3 CF-52-1-5 CF-52-1-4 CF-52-1-76 CF-51-11-195 CF-51-11-198 CF-51-12-91 Y-F30-5 Y-F30-6 180 TITLE OF REPORT Effect of the Delayed Neutrons on the Kinetic Response of the S;atidnary Liquid Fuel ARE l Theoretical and Experimental Analyses of Natural Convection Within Fluids in which Heat is being Generated. Part III. Heat Transfer from a Fluid in Laminar Flow to the Walls to a Cylindrical Tube: A Simplified Velocity Distribution was Postulated Theoretical and Experimental Analyses of Natural Convection Within Fluids in which Heat is being Generated. Part IV. Heat Transfer from a Fluid in Turbulent Flow to Two Parallel Plane Bounding Walls Theoretical and Experimental Analyses of Natural Convection Within Fluids in which Heat is being Generated. Part V. Natural Convection Velocity Measurement in a Parallel Plane System " Theoretical and Experimental Ana]&ses of Natural Convection Within Fluids in which Heat is being Generated. Part VI. Heat Transfer from a Fluid in Turbulent Flow to the Walls of a Cylindrical Tube Heat Transfer in Nuclear Reactors Physical Properties Heat Capacities Density of Fulinak Density of Flinak Physical Property Charts of Some Reactor Coolants, Fuels, and Miscellaneous Materials Measurements of the Viscosity of Flinak Viscosity Measurements of Flinak by the Falling-Ball Viscometer AUTHOR( S} C. B. Mills . C. Hamilton R. F. Redmond L. D. Palmer o o H. F. Poppendiek Hamilton’ L. D. Palmer o 0 R. F. BRedmond D. C. Hamilton . Poppendiek L. D. Palmer i 3 = = . Lyon ¥W. B. Harrison W. D. Powers G. Blalock J. M. Cisar J. M. Cisar M. Tobiasr H. F. Poppendiek ,M' Tobias S. I. Kaplan DATE ISSUED 1-7-52 1-11-52 {to be issued) 2-12-52 (to be 1issued) No &ate 11-30-51 11-30-51 12-14-51 1-28-52 2-26-52 (to be issued) o {1 b REPORT NO. " ORNL-1131 Y-F10-86 Y-F10-87 Y-F10-91 Y-F5-73 ORNL- 1046 CF-51-8-290 CF-51-12-185 Y-F5-75 ORNL-1142 CF-52-3-1 ORNL-1217 CF-51-12-70 " FOR PERIOD ENDING MARCH TITLE OF REPORT The Thermal Conductivity of Molybdenum over théKTemperapure Range 1000-2000°F Critical Masses of Some Alkali-Hydroxide Moderated Reactors The Power Géneration in a B4C Curtain ARE " on One End of the ARE No. 1 Reactor ARE with Ffie]-Cdolant in the Reflector Shielding Research Notes on the Ideal Unft Shie}d A Nuclear Plate Camera for Fast Neutron Spectroscopy at the Bulk Shielding ' Facility An Estimete of the Photoneutron Flux in the Water Surrounding the Bulk Shielding Facility Reactor - ' Estimate of Background at Tower Shielding Facility Rough Notes From Selected Shielding Literature . Multiple-Crystal Gamma -Ray Spectrometer Gamma-Ray Spectra from the Divided Shield Part I Neutron Attenuation Through Air-Filled Ducts in Water Heat Transfer Research Theoretical and Experimental Analyses of Natural Convection Within Fluids in which Part I. Heat Transfer from a Fluid in Laminar Flow to Two Parallel Plane Bounding Walls: A Simplified Velocity Distribution was Postulated ‘Heat is being Generated. AUTHOR(S) E. P. Mikol C. B. Mills C. B. Mills C. B. Mills C. B. Ellis L. A. Wills J. L. Meem E. B. Johnson J. M. LaRue G. P. Letz T. J. Morley -J. N. Renaker A. Simon C. B. Ellis F. C. Maienschein F. C. Maienschein C. E. Clifford, et al. D. C. Hamilton . Poppendiek L. D. Palmer = g 10, 1952 DATE ISSUED 1-21-52 1-21-52 2-27-52 7-10-51 1-11-52 8-24-51 12-17-51 12-5-51 (to be issued) 3-3-52 (to be issued) 12-18-51 181 { ANP DIVISION QUARTERLY PROGRESS REPORT REPORT NO. TITLE OF REPORT AUTHOR(S) DATE ISSUED Addendum to Classified Addendum to Unclassified.Memb ~ D. C. Hamilton - ! 12-19-51 CF-51-12-70 CF-51-12-70 on Natural Convection in H. F. Poppendiek » Confined Spaces | L. D. Palmer ' CF-52-1-1 Theoretical and Experimental Analyses of R. F. Redmond 1-22-51 Natural Convection Within Fluids in which D. C. Hamilton ' ~ Heat is being Generated. Part II. Heat L. D. Palmer Transfer from a Fluid in Laminar Flow to Two Parallel Bounding Walls Chemistry Y-F35-2 The Vapor Pressure of Some Salts II. Z. fur H. v. Wartenberg 1-21-52 Electrochemie 2%, 568-573 (1921) (translated H. Schulze by Mary E. Lee) ‘ Y-B31-324 Effect of pH and Concentration of Ammonium J. C. White 1-17-52 f ‘ Persul fate on Colorimetric Determination . of Nickel by Dimethylglyoxime Method- Y-F35-1- Concerning the Thermal Behavior of Sodium E. G. Bunzel 1-3-52 sy : ~ Compounds, Specifically of Sodium Oxide F. J. Kohlmeyer : and Sodium Sulfide and Their Reactions s with Metals, Z. fur Anorg. Chem. 254, ' 1-30 (1947) (translated by Mary E. Lee) ;: Y-F35-3 The Phase Diagrams for the Systems KF-MgF2 H. Remy . 1-23-52V 1 and RbF2. Rec. Trav. Chim. 59, 516-525 W, Seeman ] {1940) (translated by Mary E. Lee) % | . ' Metallurgy and Ceramics ; CF-51-11-186 Operation of a Ni-NaOH Thermal Convection Metallurgy Division 711;29-51 r ; Loop Y-B32-88 Changes in Melting Point of Fluorides G. J. Nessle | - 2-12-52 During Corrosion Tests C. J. Barton CF-52-1-144 Ceramic Materials as Related to the Reactor J. R. Johnson - 1-18-52 Program CF-52-1-192 Safety Rods for the ARE : W. D. Manly 1-29-52 ' E. S. Bomar ¢ Miscellaneous “ b Y-F26-31 ANP Information Meeting of February 20, 1952 W. B. Cottrell | 2-28-52 Y-¥26-23 ANP Ihformation Meeting of November 14, 1951 W, B. Cottrell 11-21-51 182 FOR PERIOD ENDING MARCH 10, 1952 ‘ 1 18. DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS AT ORNL() March 1, 1952 I. REACTOR AND COMPONENT DESIGN A. Aircraft Reactor Design 1. 2. B. ARE 1. 2. Survey of Cirdu]ating-Fue] Cycle Survey of Air-Water Cycle (G.E.) Reactor Design Core and Pressure Shell Fluid Circuit Design Pressure and Flow Instrumentation Structural] Analysis Thermodynamic and Hydrodynamic Analysis Remote Handling Equipment Shielding Studies for the ARE Flectrical Power Circuits Control Studies High-Temperature Fission Chamber Control System Design Control Rod Design ~ Buildifié Cafiséruétion. Internal Design E. Reaétbr Stéticsr N 1. :Sia@iés of ghg Circulatihg-FhéfVRéactors .‘2: 'Péféfietfié&8£fidiéé SfHH;O:Mbdéféted.Circfi]atifig-r | - fFue] Bgacpor | , . Statics of the-CfItiéalcEiperiméhfs' . Statics of the NaOH Moderated Reactor Préparation'of‘Reactor Calculations and Cylindrical Coordinates for the IBM 9704-1 9204-1 9201-3 9201-3 6201-3 9201-3 9201-3 9204-1 9201-3 9201-3 1000 2005 2005 9201-3 7503 1000 9704-1 o704-1 9704-1 9704-1 9704-1 Fraas Lane, Noderer Hemphill, Wesson Cristy, Lawrence, Jackson, Eckerd, Scott Hluchan Maxwell, Walker Lubarsky, Greenstreet - Longyear, Palmer Hutto, Alexander Enlund Walker Hanauer Epler, Kitchen, Ruble Estabrook Nicholson Company Browning " Mills, Edmonson, - - Uffelman, Johnson Mi]ié, Edmonson, Uffelman, Johnson Mills, Holmes Mills, Edmonson, Uffelman, Johnson Edmonson -(1)This directory was previously issued as Directory of Active ANP Research Projects ot ORNL, by W. B. Cottrell, Y-F26-30, March 1, 1952, 183 ANP DIVISION QUARTERLY PROGRESS REPORT | 6. Problem of Minimum Critical Mass ' - 9704-1 = Coveyou ; 7. Efiiggioggstrxbut1on of Therma] and Eplthermal 9704i1 ‘ Coveyou ¢ 8. IBM Calculatlons for the ORNL ARE Proposals ' 9704-1 Mills, Edmonson . : 9. 1IBM Calcu]atlons for(}EZBeactors (G E.) o 9704-1 Leeth (G.E.), Johnson 7 10. Age Calculatlons of Hydrogen Moderated Beactor (G E.) 9704;1 7 Macauley, Leeth (G.E.) 11. Temperature Coeff1c1ents of Reactivity 9704-1 ~ Osborne 12. Effect of Voids o 9704-1 Tamor F. Reactor Dynamics 1. Kinetics of the Circulating-Fue]-Béactof N 9704-1 Ergen, Prohammer, ' : ' ' , , ' Thompson, Coveyou G. Critical Experiments _ 1. ARE Critical Assembly o 9213 | Callihan, Zimmerman, Keen, Williams, Haake, Scott, Kennedy (P&W) bk il s 2. Air-Water Reactor Critical Assémb]y (G.E.). T 9213 ‘Same as ébove_ H. Pump Development‘ 1. Gas-Sea]‘Cehtriffigal Pump | - | 9201-3 McDonald, Cogb 2. Frozen-Seal Pump for Sodium and Fluorides 9201-3 Cobb, Huntley, Smith, i : : ‘ ' ; - ‘ . Taylor 3. ARE Pump Design and Development | o 19201-3 Cobb, Grindell, Taylor, . ‘ ) : : S Southern _ 4. E]ecnromagnetic Pump ' | 9201-3 McDonald, Southern, Wyld 5. Canned-Rotor Pump - 9204-1 Richardson 6. Rocking-Channel Sealless Pump ' S _ BMI Dayton 7. Seals for NaOH Systems | | BMI Simons, Allen - 8. .Seals for High-Tbmperafiure Systems | ' 9201-3 CoBb,'Johnson,'Sfiith, : . . : oo : A ‘ 7 : Tay]or, Grindell 1 I. Valve Development | j | 1. Self-Welding Tests ' ' 1 9201-3 Adamson, Petersen, Reber 2; Bellows Tests at High Temperatures | - | 9201-3 Johnson ' 3. Valves for High-Temperature Systems o 9201-3 Cobb, Ward, Johnson J. Heat Exchanger and Radlator Development 1. Nak- to-NaK Heat Exchanger o | . 9201-3 Fraas, Wyld, LaVerne, : - Petersen 2. Sodium-to-Air Radiator : A 9201-3 Fraas, LaVerne, | : . , Whitman, Petersen 3. Boeing Turbojet with Na Radiator ) " 9201-3 Fraas, LaVerne 4. Fuel-to-Liquid Heat Exchanger , ' . 9201-3 Fraas, Whitman, ' ' ‘ . ' i ) McDonald, Salmon, Bailey 184 ks FOR PERIOD ENDING MARCH 10, 1952 5. Fuel-to-Gas Radiator 6. Radiator and Heat Exchanger Design Studies Instrumentation 1. Heater Development and Application 2. Development of High-Temperature Measurement Techniques \ 3. Devejopment of Flow and Pressure Measuring Devices for Fluoride Systems 4. Leak Detection for Fluoride Systems 5. Monitoring Equipment for Sodium Leaks 6. Development of High-Temperature Strain Gage 9201-3 9201-3 9201-3 9201-3 9201-3 5201-3 K-1005 Cornell Aero. 1X. SHIELDING RESEARCH Cross-Section Measurements 1. B W N 5 6. 7 8 Neutron Velécity Selector Analysis for He in Irradiated Be Total Cross Sections of N'* and o' (G.E.) Elastic Scattering Differential Cross Section of N** and 0% (G.E.) . . 6 . Total Cross Sections Of.Ll , L17, Beg, BiO, B12, cl2 Fission Cross Sections Cross Sections of Be and C Inelastic Scattering Energy Levels shielding Measurements 1. 2. ~NLON U e . Na Bremsstrahlung Measurement Divided-Shield Mockup Tests (G.E.) -Bulk.Shiéfding Reactor Power Calibration Bulk Shiejdiné Reactor Operation ‘Heat Release per Fission VMetaf_Ducfi.Tésfis Air Duct Tests (G.E.) ‘Shielding Theory and Calculations 1. Survey | Survey Report on Shielding Shielding Section for Reactor Technol ogy Calculations of Removal Cross Sections Theory of Neutron Traasmission in Water 2005 3026 9201-2 9201-2 9201-2 9201-2 3001 9201-2 3010 3010 3010 3010 3001 3025 3001 3022, 9204-1 3022 3022 3022 McDonald, Bailey, Salmon, Whitman, Tunnell Fraas, Bailey, Salmon, Whitman McDonald, Affel McDonald, Affel Cobb, Southern, Taylor Cobb, Southern, Taylor Cameron, McKown Puffer, Grey, Donovan Pawlicki, Smith Parker Willard, Bair, Johnson Willard, Johnson, Bair Johnson, Willard, Bair Lamphere, Willard Clifford, Flynn, Blosser Willard, Bair, Kington Meem and group Johnson, McCammon Holland, Leslie, Roseberry Meem Hullings, Hull Sisman Clifford, Flynn, Blosser Blizard, Welton Blizard Blizard Blizard, Enlund 185 i e 'i.‘i’i;.;i S Sk ik i i st ANP DIVISION QUARTERLY PROGRESS REPORT 3. Interfiretation of Pb-HZO Lid Tank Data 6. Divided-Shield Theory and Design 7. Air Duct Theory (G.E.) 8. Shielding Section for Reactor Handbook 9. Cbnsu]tation on Radiation Hazards (G.E.) D. Shielding Instruments 1. Gamma Scintillation Spectrometer 2. Neutron Dosimetgr'Deve]opment 3. Proton Recoil Spectrometer for Neutrons 4. He3 Counter for Neutrons 5. LiI Crystals for Neutrons 6. Neufirqn Spectroscopy with Photographic Plates " E. Shielding Materials 1. Preparation of High-Hydrogen Rubber 2. Development of Hydridés for Shields 2005 NDA 3001 3022 2001 3010 3010 3010 3010 3010 3006 Goodrich Company MHI II1. MATERIALS RESEARCH A. Liquid Fuel Chemistry 1. Phase Equilibrium Studies of Fluorides 2. Preparation of Standard Fuel Samples 3. Special Methods of Fue]-Purification 4. Evaluétion of Fuel Purity 5. Thermodyhamic Stability and Electrochemical Properties of Fuel Mixtures - 6. Hydro]ysis and Oxidation of Fuel Mixtures 7. Simulated Fuel for Critical Experiment 8. Stability of Slurries of U0, in NaOH 9. Phase Equilibria Among Silicates, Borates, etc. 10. Fuel Mixtures Containing Hydrides 11. Chemicél Literature Searches 12, So]utjon of Metals in Their Halides B. Liquid Moderator Chemistry 1. Preparation and Evaluation of Pure Hydroxides 186 9733-3 9733-3 9733-3 9733-3 9733-3 9733-3 9733-3 BMI BMI MHT 9704-1 3550 9733-3 ' Maiefiéchein, Schenck Johnson, Haydon Overholser, Cuneo Simon w Goldstein, Feshback, Stern Clifford, Simon Blizard, Hungerford, Simon, Ritchie, Meem, i Lansing, Cochran, B Maienschein Morgan Maienscheih Hurst, Glass, Cochran Cochran, Henry, Hungerford Cochran Davidson . ! Banus Barton, Bratcher, Traber Nessle, Truitt, Morgan, Love Grimés, Blankenship, Nessle Overholser, Sturm Overholser, Tofiol B]ankenship; Metcalf Patterson Crooks Banus Lee Bredig, Johnson, Bronstein Overholser, Ketchen *, 2. 3. FOR PERIOD ENDING MARCH 10, 1952 Electrochemical Behavior of Metal Oxides in Mol ten Hydroxides Moderator Systems Containing Hydrides Corrosion by Liquid Metals 1. w 4 5 6. 7 8 Static Corrosion Tests in Liquid Metals and Their Alloys Dynamic Corrosion Research in Harps Effect of Crystal Orientation on Corrosion Effect of Carbides on Liquid Metal Corrosion Mass Transfer in Molten Metals Diffusion of Molten Media into Solid Metals Structure of Liquid Pb and Bi Alloys, Mixtures, and Combustion of Liquid Sodium Corrosion by Fluorides 1. o -~ O\ W b . . . - . 10. 11. Static Corrosion of Metals and Alloys in Fluoride Salts Isothermal Static Corrosion Tests in Fluoride Salts Fluoride Corrosion in Small-Scale Dynamic Systems Dynamic Corrosion Tests of Fluoride Salts Corrosion of Meta]s.and Their Oxides in HF Magnetic Susceptibility Due To Fluoride Corrosion Fluoride Corrosion in Rotating Cylinder Apparatus Reaction of Metals with Fluorides and Contaminants Loop Tests.with Fluorides _ Dynamic'Cbrrosion Testing of Fluorides 'Equlllbrla Between Electropositive and Transition Meta]s in Halide Melts Corr051on by Hydrox1des 1. 2, Statlc CorrOS1on of Meta]s and A]ons in Hydrox1des Mass Transfer 1n Molten Hydr0x1des 'Ph s1céj'CHéhisbry of the Hydroxide Corrosion P enomenon Statlc Corr081on by Hydrox1des Static Corr031on by Hydroxides " Physical Properties of Materials 1. 2. Density of Liquids Viscosity of Liquids. 9733-2 MHI 2000 9201-3 2000 2000 2000 2000 2000 2000 2000 9766 9766 9201-3 9766 9201-3 9201-3 9733-3 2000 2000 3550 2000 2000 2000 9766 9204-1 9204-1 Bolomey, Nichols Banus Vreeland, Day, Hoffman Adamson, Reber Smith, Cathcart, Bridges Brasunas, Richardson Brasunas, Richardson Richardson Smith Bridges, Smith Vreeland, Day, Hoffman ‘Kertesz, Buttram, Smith, Meadows Kertesz, Buttram, Smith, Meadows Adamsén, Reber Kertesz, Buttram, Croft Tunnell Tunnel ] Overholser, Redman, Powers Smith, Cathcart, Bridges Brasunas, Richardson Bredig, Johnson, Bronstein Vreeland; Day; Hoffman Brasunas, Richardson, Smith, Cathcalt Cathcart, Sm1th Kertesz, Croft Jaffee, Craighead Cisar, Kap!lan Tobias, Cisar, Kaplan, Jones 187 ANP DIVISION QUARTERLY PROGRESS REPORT 3. Thermal Conductivity of Solids 9204-1 Powers, Burnett i 4. Thermal Conduct1V1ty of L1qu1ds 9204-1 c1a{boffié, Cooper e - i 5. Specific Heat of Solids and Liquids 92041 Powers, Blalock E | 6. Thermal Diffusivity 9204-1 - Tobias S | t,? 7. Electrical Resistance of Fluoride Salts 9201-3 Affel - | k E | 8. Viscosity of Fluoride Fuel Mixtures 9766 Kertesz, Knox | % é 9. Vapor Pressure of Fluoride Fuels | 9733-2 : Barton,‘Moore | 10. Vapor Pressure of BeF, BMI Patterson, C]egg G..'HéatlTranofer . 1. Connection in Liquid Fné] Elements | 9204-1 | Hamil ton, Redmond, - _ o ‘ . ' Lynch 2. Heat Transfer in Circulating-Fuel Reactor 9204-1 Poppendiek, Pa]mén 3. Heat Transfer Coefficients of Fluorides and 9204-1 Hof fman, Lones ; Hydroxides i 4. Heat Transfef Coefficient of Lithium 9204-1 'C]aiborné, Winn ; 5. BoilingrLiQuid Metal Heat Tranoferr o ' ' 9204;1‘ Farmer. o ! 6. Sodium Heat Transfer Coefficients in Short Tubes 9204-1 Harrison ’ 7 Heat Transfer in Special Reactor Geometrles 9204-1 Claiborne _ H. Flnoride Haqdling 1. Fluoride Production | © 9201-3 'Whito Courtney, Mann é.‘ Deaién of Spacial Fluoride Handling Equipment 9733-3 Grimes, Blankenshlp, B D _ - y I Nessle 3. Prétroatment.of Fluoride-Containing Systems 9201-3 Mann, Wischhusen 4. Inspection of Components of Fluoride Systems 9201-3 Reber, Mann 5. Filling Techniques for Fluoride Systems 9201-3 Mann, Wischhusen 6. _Protroaping“of Fluoride Systems 9201-3 McDonald, Affel B ,'7. Decontamination of Fluoride Systems ) 9901-3 .Mann, Wlschhusen | i :“8. >F]uor1de Salvage and Disposal o ' 9201-3 ' Mann, Courtney, wnitex' f 1. L1qu1d Metal Handllng ’ '_ i : 1. Equlpment Cleaning Techniques 9201-3 | Mann E E 2. Continuoua.and Batch Sodium.éurification | | 9201-3 Mann h ; % 3. Samp]inéjTeohniques | 9201;3 | Mann, Blake]y E 4. Sodium Vapor Trapping 9201-3 Mann | 5. xLiquidlMooa] Salvage and Disposal ' 9201-3 _ Deven1sh Mann 6. L1qu1d Meta] Safety Equlpment 9201-3 ereven1sh Mann 7. B]anket Gas Purlflcatlon 9201;3 | Mann J. Dynamic Liquid Loops 1. Operationrof Convection Loops 9201-3 Adamson; Reber 2. Opefation.of-Figure-Eight Loops 9201-3 Cough]en 188 "y FOR PERIOD ENDING MARCH 10, 1952 3. UOa;NaOH Slurry Loop 4. Operation of Thermal Convection Loops K. Materials Analysis and Inspection Methods 1. Analysis of Fluorides for Metallic Corrosion Products 2. Analysis of Alkali Fluoridé Eutectics 3. Trace Impurities in Alkali Fluorides Determination and Removal of Boron in Diatomaceous 4. Earth 5. Metallic and Oxide Impufities in Alkali Hydroxides 6. Chemical Methods of Fluid Handling 7. Meta]]égraphié Ekafiinations 8. Identification of Compounds in Solidified Fuels 9. Preparat1on of Tested Spec1mens for Examination - 10. Identification of Corrosion Products from Dynamic Loops 11. Assembly and Interpretation of Corrosion Data from Dynamic Loop Tests . L. Radiation Damage 1. Liquid Compound Irradiations in LITR 2. Liquid Compound Irradiations in Cyclotron 3. Liquid Compound Irradiations in MTR 4. Fluoride Fuel Irradiation in Berkeley Cyclotron 5. LiquidrMeta] Corrosion in X-10 Graphite Pile Loops 6. Stress Corrosion and Creep in LITR Loops 7; Creep of Metals in X-10 Graphite'Pi]e and LITR 8. Thermal Conduct1v1ty of Metals in X-10 Graphite Pile and LITR 9, Diffusion of Fission Products from Fuels 10. Neutron Séeégfufi'of LITR 11. Irradiation of Water (G.E.) M. Strength of Materials 1, Creep Tests in Fluoride Fuels 2., Creep and Stress-Rupture Tests of Metals in Vacuum and in Fluid Media BMI 2000 9201-2 9201-2 9201-2 9201-2 9201-2 9201-3 2000 9733-2 9733-2 9733-3 9733-3 9201-3 3005 9201-3 3025 NAA 3001 3005 3001 3025 3001, 3005 3001 3005 3550 9201-3 2000 Simons Cathcart, Bridges, Smith ' White, Talbott .White, Boyd, Ross White White, Ross Whi#e Mann, B]akeiy Gray, Krofise, Roeche Barton, Anderson Nessle, Truitt, Didlake Hoffman, Blankenship Blankenship, Blakely Adamson Keilholtz, Morgan, Webster, Robertson, ~Klein, Kinyon Keilholtz, Feldman, Sturm, Jones Keilholtz, Klein, Kinyon Pearlman Sisman, Bauman, Carroll, Brundage, Parkinson, Ellis, Olsen Sisman, Bauman, Carroll, Brundage, Parkinson, Ellis, 0]sen Wilson, Zukas, Davis Cohen, Templeton Keilholtz, Morgan, Webster, Robertson, Klein, Klnyon Sisman, Trice, Lewis Taylor Adamson, Reber Oliver, Woods, Weaver 189 e i b . ANP DIVISION QUARTERLY PROGRESS REPORT 3. High-Temperature Cyclic Tensile Tests ' 2000 ' Olivef, Woods, Weaver 4. Tube Burst Tests 9201-3 ' Adamson, Reber 7 ¥ 5. Tube Burst Tests 2000 Oliver, Woods, Weaver 6. Relaxation Tests of Reactor Materials 2000 "QOliver, Woods, Weaver - 7. Creep Tests of Reactor Materials (G.E.) 2000 'O]iver, Woods,:Weaver N. Metals Fabrication Methods 1. Welding Techniques for ARE Parts 2000 . Patriarca, Slaughter 2. Brazing Techniques for ARE Parts 2000 Patriarca, Slaughter 3. Molybdenum Welding Research | BMI Parke 4. Molybdenum Welding Research ' MIT Wul ff 5. Resistance Welding for Mo and Clad Metals RPI Nippes,.Safage 6 Welds in the Presence of Various Corrosion Media 2000 Vréélénd, Patriarca, _ _ Slaughter 7. Nondestructive Testing of Tube-to-Header Welds 2000 Patriarca, Slaughter 8. Basic Evaluation of Weld-Metal Deposits in Thick 2000 Patriarca, Slaughter Plates , 9. Evaluation of the Cone-Arc Welding Téchnifiue 2000 Patriarca, Slaughter E ; 10. Development of High-Temperature Brazing Alloys Wall- Peaslee * . , Colmonoy » 11. Evaluation of the High-Temperature Brazing Alloys 2000 Patriarca, Slaughter _ & 0. New Metals Development 1. Mo and Cb Alloy Studies | 2000 Bomar, Coobs 2. Heat Treatment of Metals 2000 Bomar, Coobs P. Solid Fuel Element Fabrication 1. Solid Fuel Element Fabrication ' 2000‘ Bomar, Coobs 2. Diffusiofi—Corfosion in Solid Fuél Elements 2000 Bomar, Coobs 3. Determination of the Engineering Properties of 2000 Bomar, Coobs : Solid Fuel Elements ; 4. Electroforming Fuel-Tube to Header Configurations Gerity Graaf ; , Michigan 5. Electroplating Mo and Cb Gerity Graaf Michigan ‘ 6. Carbonyl Plating of Mo and Cb 2000 Bomar 5 7. Rdlling”of Ffie] Plate Laminates (G.E.) 3012, Bomar, Cunningham, - : 2000 Leonard : L o ! ' ? 4 Q. Ceramics and Metals Ceramics | ; 1. BeQ Fabrication Research Gerity Graaf % : : : Michigan - 2. Metal Cladding for Be0 ' ' Gerity Graaf ; , Michigan _ ? 3. 7B4C Control Rod Development | 2000 Bomar, Coobs : 190 } ~ FOR PERIOD ENDING MARCH 10, 1952 4. Hot Pressing of Tungsten Carbide Bearings 2000 Bomar, Coobs 5. High-Temperature Firing of Uranium Oxide to 2000 Bomar, Coobs Produce Selective Power Sizes - : ‘ 6. Development of Cr-UO2 Cermets for Fuel Elements 9766 Johnson, Shevlin 7. Ceramic Coatings for Stainless Steel 9766 White 8. Ceramic Valve Parts for Liquid:Metals and Fluorides 0.S.U. Shev]in 9. Application of Ceramic Materials to Reactors 9766 Johnson 10. Crucible Development for High Temperatures Norris Wilson, Doney ¢ Electric Lab. IV. TECHNICAL ADMINISTRATION OF AIRCRAFT NUCLEAR PROPULSION PROJECT AT OAK RIDGE NATIONAL LABORATORY PROJECT DIRECTOR R. C. Briant* ASSISTANT DIRECTOR FOR COORDiNATION A, J. Miller* ASSOCIATE DIRECTOB FOR ARE J. H. Buck Assistant for Coordination B. T. Macauley Administrative Assistant L. M. Cook Project Editor W. B. Cottrell PROJECT DIRECTORY SECTION NUMBER STAFF ASSISTANT FOR PHYSICS W. K. Ergen* Shielding Research E. P. Blizard IT B,C,D,E Reactor Physics W. K. Ergen* I E,F Critical Experiments A. D. Callihan I G Nuclear Measurements A. H. Snell 11 A STAFF ASSISTANT FOR RADIATION DAMAGE A, J. Miller* Radiation Damage D. S. Billington III L STAFF ASSISTANT FOR GENERAL DESIGN A. P. Fraas* General Design A. P. Fraas* I A STAFF ASSISTANT FOR ARE E. S. Bettis* ARE Design R. W. Schroeder I B,D Reactor Control E. S. Bettis* 1 C Special ARE Projects ' unassigned I K STAFF ASSISTANT FOR ENGINEERING RESEARCH R. C. Briant* Experimental Engineering H., W. Savage I H,1,J,K, 11T D,H,I,K Heat Transfer Research H. F. Poppendiek III F,G 191 . ol s, e it diibkie i iy Ceramics STAFF ASSISTANT FOR METALLURGY Metallurgy STAFF ASSISTANT FOR CHEMISTRY Chemistry Corrosion Research Chemical Analyses o Dual capacity, 192 = = T. N. McVay Manly* Manly* oo Grimes* o = . Grimes* Kertesz w. w. F. C. D. Susano IT1 111 IIT. IT1 ITI C!D}E)J’M’N’ 0,P,Q A,B D,E -