~ CENTRAL RESEARCH LIBRAI | DOCUMENT COLLECTION | DO HOT TRANSFER TO ANOTHER PERSON yii wish someons site 16 gee e doonmani B mwesme weith dosument aegd the Hheary wilt oo B b, : ORNL-1170 This document consists of 191 pages. Copy ~7 of 208 Series A. Contract No. W-7405, eng. 26 AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT for Period Ending December 10, 1951 R. Co Bl‘iant Director, ANP Project Edited by: W. B. Cottrell DATE ISSUED OAK RIDGE NATIONAL LABORATORY Operated by CARBIDE AND CARBON CHEMICALS COMPANY A Division of Union Carbide and Carbon Corporation Post O0ffice Box P Oak Ridge, Tennessee JAVTRTRRA 3 445sL 0360930 7 H M T C Gt O = CJ?:&:»—»-L‘L‘zzfljasni?i?-gaiag)rv INTERNAL DISTRIBUTION %, G. T. Felbeck (C8CCC) %Chemistry Library j_[ySiCS Librar‘y iplogy Library ey etdlkurgy Library rain®pg School Library entrafi%Files Cegter . Lafgpn . Hume@a(K—ZS) Laver&%fiY*lz) . o Briant Swartout . Cisar Snell ollaender | Steahly 5; Morgan fl§ Cardwekfi Kelley’ . King# Wiggirs e I PMEHENCISRE OO IROOMmME #4th Physics Library . VonderLage ™% 37. 38. 39. 40 . 41. 42, 43. 44. Jfigfi. 47, 48. 49, 50. 51. 52. 53. 54. 55. 56. 57. 58. 59. 60. 61. 62. 63. 64, %5-74. wpfi:?ssspgém>momcp>§fiwn>smfim ZECOMETENOrCYUORH TN E WD TZ RUIRE W % ORNL-1170 Progress . Stoughton Bruce Sawage Eister “Householder Graham Lyon . Keim Gall . Miller . Schroeder . Billington Blizard Clifford Clewett Callihan Livingston Manly Meem Susano Cottrell Breazeale . Grimes rasunas Poppendiek . Uffelman Cowen Reyling ANP Library Central Files (0.P.) 84-86.% 87-108. 109-118. 119. 120-127. 128. 129-131. 132. 133, 134-139. 140. 141. 142-146. 147-150. 151. 152. 153-156. 157. 158-161. 162-164. 165. 166-167. 168-171. 172. 173-1745 175-174" #18. 1794180 181§ 184, 185-193. 194-208. % Air Force Engineering Office $leneral Electric Co., Oak Rige fi&gonne National Laboratory# ORNL-1170 Progress EXTERNAL DISTRIBUTIONG Afped Forces Special Wea At¥pic Energy Commissiory Khaven National g, 0f Aeronautiy ric Company, Richland i Company Knollg Atomic Po Los #amos Laboratory Ma;fiéchusetts Insti e of Technology (Kaufmann) ind Laboratory % ational Advisory Comm National Advisory Commi New York Operations Of{ North American Aviation, Patent Branch, Washington Savannah River Operations Of % University of California Radidfaon Laboratory Westinghouse Electric Corporatigy Wright Air Development Center % Technical Information Service, (al tee for Aeronautics, Cleveland ece for Aeronautics, Washington iil iv Reports previously 1ssued ORNL- 528 ORNL-629 ORNL-768 ORNL-858 ORNL-919 ANP-60 ANP-65 ORNL-1154 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, 1931 Ending June 10, 1951 Ending September 10, 1951 TABLE OF CONTENTS FOREWARD PART I REACTOR THEORY AND DESIGN SUMMARY AND INTRODUCTION 1. CIRCULATING-FUEL AIRCRAFT REACTOR Airplane and Overall Arrangement Reactor ' Engine Radiators Shielding Accessory Systems LIQUID-METAL-COOLED ATRCRAFT REACTOR EXPERIMENT Core—Reflector—Pressure Shell Fluid Circuit Reactor Control Control system High temperature fission chamber Reactor dynamic computer Instrumentation ARE Building Faciliaty Remote-Handing Equipment REACTOR PHYSICS Reactor Calculations on IBM Equipment Circulating-Fuel Reactors Beryllium~oxide moderated reactor Water-moderated reactor Comparison of H,0- and Be0-moderated reactors Effect of delayed neutrons on kinetic response Circulating-Moderator Reactors Errors in Reactor Physics Calculations PAGE VW 00 ~1 ~3 4N 11 11 11 12 12 12 12 12 12 13 13 14 14 17 29 30 31 31 4. vi CRITICAL EXPERIMENTS Preliminary Assembly of Direct-Cycle Reactor Graphite Heactor NUCLEAR MEASUREMENTS The 5-Mev Van de Graaff Accelerator Measurements of the (n,2n) Reaction in Beryllium Time-of-Flight Neutron Spectrometer EXPERIMENTAL REACTOR ENGINEERING Pump Development Centrifugal pumps for figure-eight loops ARE centrifugal pump design Canned-rotor pump Frozen-sodium-seal pump Two-stage electromagnetic pump Electromagnetic pump cell development Seal Tests Frozen-sodium seal Frozen-fluoride seal Graphitar ring~--ketos tool steel gas seal Test Loops Calibration loop Sodium manometer loop Self-welding tests Materials tests Valve tests Heat-Exchanger Tests NaK to NaK heat exchanger Sodium to air radiator Liquid-Fuel Systems Instrumentation Level control and indication Flow measurement Pressure-measuring devices Full-Scale ARE Component Test Facilities PAGE 35 35 37 38 38 38 39 40 40 41 41 42 42 42 42 43 43 43 43 44 44 44 44 44 45 45 45 47 47 47 48 48 438 48 Fluoride Production Cleaning of Fluorides from Systems NaK Disposal ' Alkali Metals Manual PART I1 SHIELDING RESEARCH SUMMARY AND INTRODUCTION 7. BULK SHIELDING REACTOR Reactor Operation Mockup of the Unit Shield Mockup of the Divided Shield DUCT TESTS Air-Filled Duct Tests in Lid Tank Cylindrical ducts Annular ducts Liquid-Metal Duct Test in Thermal Column SHIELDING INVESTIGATIONS Tower Shielding Facility Proposal Circulating-Fuel Reactor Shields NDA Divided-Shield Studies PART 111 MATERIALS RESEARCH SUMMARY AND INTRODUCTION 10. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS Low Melting-Fluoride Fuel Systems LiF-KF-UF, LiF-NaF-KF-UF, LiF-NaF-RbF-UF, NaF-BeF,-UF, LiF-NaF-BeF,UF, Tonic Species in Fused Fluorides PAGE 49 49 50 50 53 54 54 55 59 65 65 65 72 72 73 13 74 T4 17 79 79 80 80 80 81 81 82 vii 11. viii Experimental procedures Results of electrolyses Homogeneous Fuels Uranium solubility in hydroxide-borate mixtures Uranium solubility in hydroxides with various additives Moderator-Coolant Development Preparation of pure hydroxides KOH-LiOH Hydroxide-fluoride systems Coolant Development RbF-LiF NaF-Bel, NaF-KF-RbF KF¥-LiF-RbF NaF-LiF-RbF Ternary systems containing BeF, Service Functions CORROSTION RESEARCH Static Corrosion by Fluorides Static Corrosion by Hydroxides Corrosion of uncoated metals Corrosion of coated metals Dissolution of metals in sodium hydroxide Static Corrosion by Fluoride-Hydroxide Systems Static Corrosion by Sodium Cyanide Static Corrosion by Liquid Metals Sodium on stainless steel Lithium on coated stainless steel Static Corrosion of Fuel Capsules in Sodium Dynamic Corrosion in Thermal Convection Loops Fluoride corrosion Hydroxide corrosion Fundamental Corrosion Research EMF measurements in hydroxides Polarographic study of sodium hydroxide Survey of the mass-transfer phenomena PAGE 82 82 83 83 83 84 84 85 85 85 85 86 86 86 86 87 87 88 89 89 90 98 100 102 103 103 103 103 105 106 107 107 110 110 110 111 PAGE 12. HEAT-TRANSFER RESEARCH AND PHYSICAL PROPERTIES 115 Natural Convecticn in Liquid-Fuel Elements 115 Heat-Transfer Coefficients 116 Heat transfer in fused hydroxides and salts 116 Heat transfer in molten lithium ‘ 117 Entrance-region heat transfer in a sodium system 118 Heat transfer in a circulating fuel system 120 Heat Capacity 120 Thermal Conductivity 120 Thermal conductivity of liquids 120 Thermal conductivity of solids 122 Thermal conductivity of diatomaceous silica powder 122 Density of Liquids 123 Viscosity 123 Falling-ball viscometer 124 Zahn type viscometer 124 Viscosity of fluoride mixtures 126 Vapor Pressure of Liquid Fuels 126 13. METALLURGY AND CERAMICS 128 Solid-Fuel-Element Fabrication 128 Effect of U0, particle size 129 Effect of rolling temperature 130 Effect of elimination of capsule 130 Preparation of tubular fuel elements 132 Welding Techniques 132 Cone arc welding 132 Resistance welding 132 Brazing Techniques 135 Effect of brazing time 135 Brazing of clad fuel elements 135 Nickel-palladium brazing alloy 135 Creep and Stress Rupture of Metals 137 Operation of creep and stress-rupture equipment 138 Creep-rupture tests of inconel 139 Creep of nickel Z 140 Stress-rupture tests 140 ix 14. Ceramics Laboratery Equipment Subcontract work RADIATION DAMAGE Irradiation of Fused Materials Pile irradiation of fuel and KOH capsules Cyclotron irradiation of fuel and KOH capsules In-Pile Circulating Loops Creep Under Irradiation Radiation Effects on Thermal Conductivity PART IV APPENDIXES SUMMARY AND INTRODUCTION 15. 16. SUPERCRITICAL-WATER REACTOR Analysis of Supercritical Water Reactor by NDA Stability Start-up Analysis of Supercritical-Water Reactor by ORNL Compressor-jet cycle Results ANALYTICAL CHEMISTRY Studies of Diatomaceous Earth Determination of Ni, NiO, and O in Alkali Hydroxides Available oxygen Metallic nickel Nickel oxide Studies of Ternary Alkali Fluoride Eutectic pH of aqueous solutions of alkali fluorides Composition of the euntectic Metallic impurities Corrosion of Metal Containers by Hydroxide Solutions Determination of Uranium Trifluoride Determination of Oxygen in NaK PAGE 141 141 141 142 142 142 143 144 144 145 149 150 150 150 150 151 151 151 152 152 153 153 153 153 154 154 155 155 155 155 155 PAGE Determination of Oxygen in Helium 156 Preparation of Oxygen-Free Sodium Samples 156 Clarity of Borated Water in Concrete Tanks 157 Analytical Services 157 17. LIST OF BEPORTS ISSUED 158 18. DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS AT ORNL 161 Reactor and Component Design 161 Shielding Research 164 Materials Research 165 Analysis of Other Nuclear Reactor Systems 172 19, TECHNICAL ORGANIZATION OF THE ANP PROJECT 173 10. 10. 10. 10. 10. 10. 10. x11 N e W N LIST OF TABLES TITLE PAGE Features of the Circulating-Fuel Aircraft Reactor 8 Performance of Engines at Design Point, Mach 1.5 at 45,000 ft 9 Design Values for the BeO-Moderated Circulating-Fuel Reactor 14 Design Values for the NaF-UF, - Cooled Water-Moderated Reactor 17 Design Values for the Water-Moderated Circulating-Fuel Reactors for Use with the NaF-BeF, and LiF-KF-NaF Fuel-Coolants 22 Comparison of BeO- and H,0-Moderated Reactor 29 The NaOH-Cooled and —~ Moderated Reactor 31 Reactivity Calculation Results on the Alkali Hydroxide Reactors 32 Critical Assembly Sizes, Composition, and Calculated Multiplication Constants 33 Summary of Promising Fluoride Fuel Systems of Low Uranium Content 30 The Pseudo-Binary System (LiF-NaF-KF)-UF, 81 The Pseudo-Binary System (LiF-NaF-RbF)-UF, 81 The Pseudo-Binary System (LiF-NaF-BeF,)-UF; 81 Solubility of Uranium in Hydroxide-Borate Mixtures 83 Effect of Various Additives on the Solubility of Uranium in Hydroxides 84 Container Filling Services 87 TABLE 11, 11. 11. 11. 11. 11. 11. 11. 11. 11. 11. 12. 12. 12. 14. 16. 16. 10 11 - b Q2 N e TITLE Summary of Corrosion by Ba{OH), at 816°C for 100 hr Summary of Corrosion by Sr(OH), at 816°C for 100 hr Summary of Corrosion by Sodium Hydroxide at 815°C for 100 hr Summary of Corrosion by Lithium Hydroxide at 816°C for 100 hr Summary of Hydroxide Corrosion of Clad Metal Specimens at 816°C for 100 hr Metal Content of Sodium Hydroxide at Function of Temperature Corrosion Data Obtained at 816°C Using Molten Sodium Cyanide for 100 hr Static Corrosion Tests in Sodium Analysis of Inconel Fuel Capsules in Sodium at 800°C Analysis of 316 Stainless Steel Fuel Capsules in Sodium at 800°C Analysis of Hydroxide From Nickel Thermal Convection Loops After Plugging Heat Capacities of Various Substances Data on Several Fused Salt Mixtures Vapor Pressure of the NaF-KF-UF, Eutectic Tests on Standard NaF-KF-UF, and KOH pH Values of Aqueous Solutions of Certain Alkali Fluorides at 25°C Summary of Service Analyses PAGE 90 91 02 92 99 102 104 105 106 . 106 113 121 124 127 143 154 157 X111 LIST OF FIGURES FIGURE TITLE PAGE 3.1 Total Uranium Inventory and Uranium in Core as a Function of the Volume Fraction Fuel Coolant in the Core for Be(0-Moderated Circulating-Fuel Reactors with 3-, 3 1/2-, and 4-ft-Diameter Cores 15 3.2 Total Uranium Investment in BeO-Moderated Circulating-Fuel Reactor as a Function of Reactor Core Diameter for Several Assumed Holdup Volumes External to Core 16 3.3 Percent Thermal Fissions as a Function of the Fuel-Coolant Volume Fraction in the 3- and 4-ft-Diameter Cores of the BeO-Moderated Circulating-Fuel Reactor 17 3.4 Fission Spectrum vs. Lethargy for the BeO-Moderated Circu- lating-Fuel Reactor with a 3 1/2-ft-Diameter Core 18 3.5 Absorption Spectrum vs. Lethargy for the Be(-Moderated Circu- Jating-Fuel Reactor with a 3 1/2-ft-Diameter Core 19 3.6 Leakage Spectrum vs. Lethargy for the Be(C-Mcderated Circu- lating-Fuel Reactor with a 3 1/2-ft-Diameter Core 20 3.7 Neutron Flux Spectrum vs. Lethargy for the BeO-Moderated Cir- cuwlating-Fuel Reactor with a 3 1/2-ft-Diameter Core 21 3.8 Total Uranium Inventory and Uranium in the Core as a Function of the Volume Fraction Fuel-Coolant ipn the 3-ft-Diameter-Core Water-Moderated Reactor with NaF-UF4 Fuel-Coolant 22 3.9 Total Uranium Inventory and Uranium in Core as a Function of the Volume Fraction Fuel-Coolant in the Core for the Water-Mod- erated Circulating-Fuel Reactor with 2 1/2-, 3-, and 3 1/2-ft- Diameter Cores 23 3.10 Fission Spectrum vs. Lethargy for the H,0-Moderated Circulating- Fuel Reactor with 3-ft-Diameter Core 24 3.11 Absorption Spectrum vs. Lethargy for the H,0-Moderated Circu- lating-Fuel Reactor with 3-ft-Diameter Core 25 X1iv FIGURE N~ Al =3 N e =2 . 12 .13 . 14 .15 . 16 ~ o B W N TITLE PAGE Leakage Spectrum vs. Lethargy for the H,0-Moderated Circulating- Fuel Reactor with 3-ft-Diameter (ore Neutron Flux Spectrum vs. Lethargy for the H,O-Moderated Cir- culating-Fuel Reactor with 3-ft-Diameter Core Total Uranium Inventory and Uranium in Core as a Function of Volume Fraction Fuel-Coolant in the Core for H,0-Moderated Circulating-Fuel Reactors with 2 1/2-, 3-, and 3 1/2-ft- Diameter Cores and LiF-NaF-KF-UF, Fuel-Coolant Fractional Change in Neutron Flux as a Function of Time for 0, 67, and 80% Decrease in Delayed Neutrons Effective Multiplication Cohstant as a Function of Volume Fraction of Stainless Steel in the Core for the 2 1/2-ft- Diameter-Core NaQOH-Moderated and - Cooled Reactor Mid-Cross-Section of Second Mockup of Direct-Cycle Reactor Sodium to Air Radiator Gamma Radiation Intensity for the Unit-Shield Experiments Thermal -Neutron Flux for the Unit-Shield Experiments Fast-Neutron Dosage for the Unit-Shield Experiments Dependence of the Weight of the Unit Shield on its Thickness Installation of Divided-Shield Mockup with Reactor in Position Shadow-Shield Experiment with Sodium Source Bulk Shielding Facility Fast-Neutron Data An Array of Four 2-in. Steel Conduit Ducts with Three 90° Bends 26 217 28 30 32 36 46 36 57 o8 60 61 62 63 66 Xv FIGURE 8. 10. 10. 10. 10. 10. 10. 11. 11. 11. 11. xvil 2 SN e W N TITLE PAGE Lid Tank Duct Test D-915, X Traverse (Vertical) and Z Center- line Measurements of Neutron Flux for Two Arrays of 2-in, Steel Conduit Lid Tank Duct Test D-10, Y Traverses in H,0O behind 54-in. Rubber Conduit (2 in. 1.d.) with Two Bends of Variable Radius Lid Tank Duct Test D-11, Y Traverses (Horizontal) in H,0 behind 52-in. Rubber Conduit (2-3/8 in. i1.d.) with Variable Bends Lid Tank Duct-Test Annular-Duct Rectangular Cross-Section, 1/8- in. Steel Plate Welded Lid Tank Test D-12, Y (Horizontal) and Z Traverses behind Amnular Duct The System KOH-LiOH The System RbF-LiF The System NaF-BeF, The System NaF-KF-RbF The System KF-LiF-RbF The System NaF-LiF-RbF Surface of 304 Stainless Steel after 100 hr of Exposure to Barium Hydroxide at 816°C Surface of 446 Stainless Steel Specimen after 100 hr of Ex- posure to Barium Hydroxide at 816°C : Surface of Inconel Specimen after 100 hr of Exposure to Barium Hydroxide at 816°C Surface of Nickel Z Specimen after 100 hr of Exposure to Barium Hydroxide at 816°C 67 68 69 70 71 85 85 86 86 86 87 93 93 94 95 FIGURE 11. 11. 11. 11. 11. 11. 11. 11. 11. 11. 11. 11. 11. 12. 10 11 12 13 14 15 16 17 1 TITLE Surface of Chromium Specimen after 100 hr of Exposure to Sodium Hydroxide at 816°C Surface of Inconel Specimen after 100 hr of Exposure to Lithium Hydroxide at 80G°C Surface of Inconel Specimen after 100 hr of Exposure to Rubidium Hydroxide at 800°C Effect of Exposure Time on the Corrosion of Inconel by Sodium Hydroxide at 800°C Effect of Exposure Time on the Corrosion of Inconel by Potassium Hydroxide at 800°C Effect of Temperature on the Corrosion of Inconel by Potassium Hydroxide in 100 hr Weight Loss of Inconel Specimen in Potassium Hydroxide for 100 hr as a Function of Temperature Surface of 304 Stainless Steel Coated with 3-mil Nickel Electro- plate after 100 hr of Exposure to Barium Hydroxide at 816°C Nickel-Clad Inconel (Nickel Sheet) after 100 hr of Exposure to Sodium Hydroxide at 816°C Nickel-Clad Inconel (Nickel Powder) after 100 hr of Exposure to Sodium Hydroxide at 816°C Nickel Thermal-Convection Loop Walls with Potassium Hydroxide Plugged Section Nickel Thermal Convection Loop with Sodium Hydroxide Coolant ' Deposit of Silver Crystals by Mass Transfer in Silver Capsule Containing Sodium Hydroxide Temperature Ratio in a 3-mm Tube Filled with Brine in Which Heat is Generated Uniformly PAGE 85 96 96 97 97 98 99 100 101 102 108 109 112 116 XxXvii FIGURE TITLE 12.2 Flow Diagram for Lithium Heat-Transfer Experiment 12.3 Thermal Conductivity of Diatomaceous Earth 12.4 Viscosity Apparatus 13.1 Effect of U0, Particle Size on UQ, Distribution 13.2 Effect of Rolling Temperature on UQ, Distribution 13.3 Effect of Elimination of Capsule on UQ, Distribution 13.4 Seamless-Tube Fuel Elements Formed by '"Rubberstatic'" Pressing 13.5 Transverse Section of a 0.015-in.-Thick Inconel Sheet Spot Welded to an 0.188-in.-0.d. Inconel Tube of 0.025-1n. Wall Thickness 13.6 Section of a Spot Weld Joining Two Stainless Steel—Clad Fuel Plates 13.7 Longitudinal Sections of a Type 316 Stainless Steel Tube-to- Header Joint Nicrobrazed at 1120°C in a Dry Hydrogen Atmos- phere 13.8 Transverse Section of a Stainless Steel-~Clad Fuel-Element Butt Nicrobrazed to a Stainless Steel Sheet 13.9 Transverse Section of an Incomel Tube-to-Header Joint Brazed with a 60% Pd - 40% Ni Alloy in Dry Hydrogen at 1270°C for 20 min 13.10 Stress-Rupture Time for Inconel Sheset 13.11 Time-Elongation Curve for Nickel Z Xxviii PAGE 118 123 125 129 130 131 133 134 134 136 137 138 139 140 FOREWORD This is the quarterly progress report of the Aircraft Nuclear Propui- sion 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, Y-7405, Eng 26, but such research for the national ANP program as 1s performed by Laboratory personnel. The report is divided into four parts: I, Reactor Theory and Design; 11, Shielding Research; III, Materials Research; and IV, Appendixes. As each of these parts may be regarded as a separate entity, each has a separate "Summary and Introduction” which pre- cedes the part in this report. SUMMARY AND The search for a nonoxidative high- temperature fluid other than sodium which would be suitable as a reactor coolant has lead to the proposed use of fused fluoride salts containing uranium (Sec. 1). The resulting cir- culating-fuel reactor would have the important advantage of eliminating a heat-transfer stage within the reactor core. Preliminary design studies of such a reactor indicate that a 3.5-ft- “diameter beryllium oxide—moderated circulating-fuel reactor will produce around 350 Mw at amaximum temperature of 1500°F, The design point of the aircraft incorporating this reactor is Mach 1.5 at 45,000 ft. Performance and weights of the airplane, reactor, shield, engines, and radiators are being explored. Although the Aircraft Reactor Experiment (Sec. 2} was originally intended as a prototype of the sodium- cooled quiescent-liquid-fuel aircraft reactor, substantial portions of the ~equipment would be applicable to the circulating-fuel reactor. Consequently, construction, design, and procurement of equipment have proceeded on the original schedule. The building facility itself is 80% complete and mest components are on their design or delivery schedule. Reactor physics calculations (Sec. 3) have been devoted primarily to the staties, and to some lesser extent the kinetics, of the circulating-fuel reactor. The minimum uranium invest- ment 1in the beryllium oxide—moderated circulating-fuel reactor has been determined as 69 1lb of which 26 lb 1is in the core. (The uranium investment in a water-moderated circulating-fuel reactor is somewhat lower because of the better moderating capacity of water, but design considerations appear to favor the use of Be0,) The percent INTRODUCTION thermal fissions of this BeO-moderated reactor is 52, 1.e., the reactor lies between epithermal and intermediate. Kinetic studies, which have shown the practicality of the conservation of delayed neutrons, imply that the design of the reactor must provide for as high a fuel volume as practical, possibly even at the expense of the uranium investment. A second mockup of the General Electric "direct cycle” reactor has been assembled (Sec. 4)}. The critical mass of this assembly was 90 1b of uranium. The cross-sections of iron and beryllium are being determined for refined reactor physics calculations (Sec. 5). Measurements of fast neutrons in the 5-Mev Van de Graaff give an average total cross-section for iron which varies from 2.5 barns at 0.6 Mev to 3.2 barns at 3.6 Mev. Upper limits for the (n,2n) cross-section has been determined in one case to be < 0.56 barn and in another to be < 0.26 barn. Final adjustment are being made on the time-of-flight neutron spectrometer with which resolutions of at least 1.2 pusec per meter are anticipated. The development of reactor plumbing and associated hardware, while largely concerned with that for a sodium- cooled reactor, has been redirected to the requirements of the circulating- fuel reactor (Sec. 6). Valves, pumps, seals, heat exchangers, and related equipment are being developed for both fluids for operation at reactor temper- atures, i.e., up to 1500°F. However, extensive development with the fluoride fuels has first necessitated a program for their manufacture, purification, and handling; hence, to date the experimental work with fluorides has ANP PROJECT QUARTERLY PROGRESS REPORT been somewhat limited. A frozen- fluid-seal centrifugal pump which was successful with sodium is being modi- fied for use with the fused fluorides. A centrifugal pump with a gas seal, however, has already been used to pump the fluorides satisfactorily for short periods. Valve tests in sodium systems show significant increase in torque with time although several combinations cf metals appear to resist self-welding. A NaK to NaK heat exchanger has now operated for 550 hr with a maximum temperature of 1200°F. 'FOR PERIOD ENDING DECEMBER 10, 1951 1. CIRCULATING-FUEL AIRCRAFT REACTOR R. W. Schroeder, The objective of the Oak Ridge National Laboratory in the national Aircraft Nuclear Propulsion Program is the development of high-performance reactors, 1.¢., for supersonic propul- sion. This implies the use of liquid- conolant systems which require small core sizes and have good heat-transfer characteristics. The specific objective of the ORNL-ANP project is, accordingly, the exploitation of nonoxidative high- temperature fluids. This line of research avoids not only the high pressures associated with some cycles but alsc the oxidation inherent to others. A sodium-cooled guiescent- liquid-fuel reactor was the Laboraea- tory’'s first considered proposal as a reactor with potentialities for super- sonic flight. However, the almost prohibitively difficult task of assuring the safety of a sodium~cooled and water-shielded reactor, as well as the limitations of a unifunctional coolant, has enhanced the search for a more versatile, and less inflammable, coolant. As a result of this search, OBNL has turned toward the use of fused fluoride salts (generally a ternary or quaternary sSystem composed of uranium fluoride and a mixture of two or three alkali1 fluorides or beryllium fluoride) -—— not for just the heat-transfer medium, however, but as the reactor fuel as well. In addition to the advantageous physical properties of the fused salts (although not so good as sodium from the stand- point of heat transfer) and their noninflammability in air and water, an important advantage of such a cir- culating-fuel reactor 1s that 1t eliminates a heat-transfer stage within the reactor core. The advantage to be gained from separating core and heat exchanger ANP Division cannot be overemphasized. Although it is perhaps possible to conceive of a small and at the same time a well- designed 500 Mw heat exchanger, the requirements of these entities are so different that i1t is difficult te design the heat exchanger into the reactor without severely penalizing both. Furthermore, whereas the sodium- cooled reactor required considerable departure from conventional design practice to remove approximately 200 Mw from a 3-ft core, it now appears feasible, from a fabricaticnal and fluid~flow standpoint, to remove 350 Mw from a 3%-ft core employing circulating fuel. The c¢irculation of fissionable material external to the core dirscts attention to design studies of the entire system -— rveactor, radiator, engine, etc. Although the circulating- fuel system has not been studied suffi- ciently long to ensure the performance of the resulting aircraft, the more outstanding problems of this cycle have been appreciated. The preliminary design studies which have been initiated for the exploration of a supersonic airplane application of a circulating- fuel reactor are encoursaging. However, performance and weights of the air- plane, reactor, shielding, engines, radiators, and other components are sco interrelated that i1t was found necessary to make studies of the overall arrange- ment. The studies have not yet been completed, and it is not possible to to draw conclusions at this time. However, the arrangement being studied may be cursorily described. ' AIRPLANE AND OVERALL ARRANGEMENT The airplane visualized is designed for a speed of Mach 1.5 at 45,000 f¢, ANP PROJECT QUARTERLY PROGRESS REPORT with a gross weight of approximately 350,000 1b, an L/D ratio of approxi- mately 6.5, and a wing loading of approximately 70 1b/ft?, A divided shield 1s employed with gamma and neutron shielding about the crew com- partment and neutron shielding about the reactor. Six turbojet engines are arranged 1n the fuselage in a circle aft of the reactor-shield assembly, and circulating fuel isducted directly from the reactor to the engine radiators, thereby eliminating the weight and temperature loss associated with the use of i1ntermediate heat exchangers. REACTOR Water-moderated and solid-moderated reactors have been explored, and tentatively, both types appear to be feasible. However, the water-moderated and -reflected reactor would, for this application, require a water- cooling system capable of disposing of approximately 30,000 kw with a small radiator temperature difference and a small air temperature rise. The small temperature difference would require very large radiaters, and the small air temperature rise would entail very large cooling air flow requirements. Further, the use of water as a moderator involves a potential hazard associated with rapid steam formation if a failed fuel tube should permit abrupt mixing of fuel and water. Accordingly, the studies outlined here are based on the use of BeO as amoderator and reflector. Moderator heat is removed by the cir- culating fuel and is employed usefully in the propulsion cycle. By proper allocation of coolant flows and coolant-tube surface areas, it is possible to maintain maximum coolant- tube wall temperatures only slightly in excess of the circulating-fuel maximum temperature. To permit re- moving reflector heat with a high radiator temperature difference, the design under study postulates cooling the reflector with a non-fuel-bearing fluoride mixture. This reactor con- forms to the specifications given in Table 1.1. Table 1.1 Features of the Circulating-Fuel Aircraft Rezactor Size 3.5-ft sphere Moderator BeO Moderator ~65% Fuel ~ 33% Structural ~2% Design point Power ~ 325,000 Btu/sec Reactor fuel inlet temp. "~ 1000 °F Beactor fuel outlet temp. ~1500°F ENGINE A circulating-fuel type reactor should be inherently capable of supply- ing propulsive power by means of a turbojet cycle or either of several vapor cycles. The relative superiority of these cycles has not yet been com- pletely established, but, pending com- parisons, the use of the turbojet cycle has been presumed. Preliminary optimization studies have indicated that a compression ratio in the region of 6.1, with a turbine inlet tempera- ture of approximately 1250°F, 1is desirable. Further increases 1n the compression ratio would diminish the radiator weight at the expense of engine weight, and further increases in turbine inlet temperature {(with fixed reactor conditions), would favor turbojet weight to the detriment of the radiator log mean temperature differ- ence and radiator weight. More rigorous evaluations of these component weights will be required before the turbojet specifications indicated here can be accepted with any degree of fimality. FOR PERIOD ENDING DECEMBER 10, 1951 ° Installational convenience favors the use of asmall number of relatively large engines, whereas engine develop- ment and availability considerations favor the use of a larger number of smaller engines., Tt is difficult to predict at this time the size of engine that could be made available when an airplane of the type studied would require these engines, but 1t appears that the validity of the overall study is relatively insensitive to the number of engines presumed. Therefore the use of six engines conforming to the general specifications listed in Table 1.2 has been assumed. Table 1.2 Performance of Engines at Design Point, Mach 1.5 at 45, 000 ft Thrust 8200 1b Maximum diameter 61 in. Compression ratio 6.1 Turbine inlet temperature 1250°F Equivalent sea-level airflow 665 lb/sec Calculations have indicated a specific impulse of approximately 30 and overall efficiency of 30%. This latter 1s defined as the ratio of mnet thrust horsepower to reactor thermal hersepower. RADIATORBS Radiators with heat-transfer capac- ity per unit volume 1n excess of the best obtained to date will be desired for any ligquid-cycle supersonic nuclear-powered airplane. Arrangements are being made to receive the advice and consultation of established heat- exchanger manufacturers relative to the design of radiators for such appla- cations. This advice will not be available for at least several months, however, and 1t was necessary to achieve a preliminary radiator design in order to permit related phases of the study to continue. The design conceived of, which may be far from optimum, involves a radial grouping of rectangular-shaped banks about the engine centerline between the com- pressor and turbine. Liquid-conveying tubes are passed through, and normal to, closely spaced sheet fins. The flow pattern contemplated 18 counter- current, as dictated by the tempera- tures of the two fluids. The arrange- ment entails the use of dividing baffles between adjacent banks, and the 1nstallation of by-pass valves in these baffles will permit controlling turbine inlet temperature while a substantially constant liguid tempera- ture 1s maintained. SHIELDING Shielding calculations currently are 1n process, and it 1s not possible to make a guantitative description of the shielding at this time. Initial studies have indicated (see "Circu- lating-Fuel-Reactor Shields™ in Seec. 8), however, that the circulating fuel in the radiators and the associated plumbing may remain unshielded provided that sufficient neutron and gamma shielding is placed about the crew compartment, and alsec provided that the payload, and possibly other com- ponents, are protected by local shielding. It currently i1s contemplated that the reactor shield will include only hydrogenous material, ACCESSORY SYSTEMS Accessory circuits are required to permit cooling the reflector and shield and to provide power for pumps and general aircraft accessory demands. As it appears that many of the acces- sories are favored by the use of variable-speed drives, the overall ANF PROJECT QUARTERLY PROGRESS REPORT accessory system contemplated involves the use of individual pneumatic tur- bines for power supply. The energy 1s supplied, and reflector cooling 1is achieved, by passing compressor bleed- of f air through the reflector radiators and then to the accessory turbines and to parallel propulsive nozzles. The accessory turbines are controlled by me ans of variable-area discharge nozzles, which provide some net thrust after imparting energy to the accessory turbines. Preliminary analysis of this system indicates that 1t not only provides reflector cecling and acces- 10 sory power adeguately, but that the reflector cooling system, when con- sidered as an open Braytom cycle power plant, has a favorable cycle efficiency and power-weight ratio. The use of 1ndividual accessory drive turbines permits supplying the acces- sories with power and speed control without special mechanical, hydraulic, or electrical transmission systems. The shield-water radiator is cooled by low-pressure compressor bleed-off air, and the =2ir is then discharged through a propulsive nozzle. FOR PERIOD ENDING DECEMBER 10, 1951 2. LIQUID-METAL-COOLED AIRCRAFT REACTOR EXPERIMENT R. W. Schroeder, Aircraft reactor cycles other than the sodium-cooled stationary liquid- fuel cycle have been(!”? and are (Sec. 1) under active study. A large portion of design effort has, therefore, been diverted to study of these alternate cycles., These i1nvestigations have indicated a high probability that substantial portions of the equipment, designed for the original sodium-cooled reactor experiment, would still be usable 1f the ARE were to be a prototype of one of these other cycles. Ac- cordingly, design and procurement of the building and equipment have been continued on the original basis. At the present time the facility under contract te the Nicholson company is approximately 80% complete. All major fluid-circuit components have been designed and are on order or under construction locally, and all core and pressure shell components are either scheduled for fabrication or have already been completed. Instru- mentation, electrical circuits, and remote-handling equipment have been partially designed, and 1tems requiring long fabricational time spans have been ordered. CORE~ REFLECTOR—PRESSURE SHELL Approximately 90% of the BeO moderator and reflector blocks have " been received., The inconel pressure- shell billet has been shipped from International Nickel Co. to Lukenweld, and the shell is scheduled for December rolling at Lukenweld. The fuel- (. F. Hemphill, R. W. Schroeder, and H. R, Wesson, “Circulating-Moderator-Coolant Reactor: OBNL,” Atircraft Nuclear Propulsion Project Quar- terly Progress Report for Period Ending September 10, 1951, OBNL-1154, p. 188 (Dec. 17, 1951). ANP Division element tubing has been drawn by Superior Tube Company and the 1.235-1in. coolant tubes are in process at Superi- or. Miscellaneous sheet and bar stock for the core have been received. Detailed core manufacturing drawings have been made and retained pending material availabilicy. FLUID CIRCUIT The fluid-circuit design for the original NaK system {(the use of Nak was a modification, to eliminate pre- heating, of the use of sodium) is approximately BO0% complete, and all major components have been released for procurement or local manufacture. These include (1) inconel pipe, (2) dump tanks, (3) all heat exchangers, (4) helium blowers for heat disposal loop, (5) Vickers variable-speed hydraulic drive system for helium blower drive, (6} space-cooling—heat- exchanger-——fan combinations, (7) con- trol-rod cooling rotary blowers, and (8) NaK purification assembly. REACTOR CONTROL E. S. Bettis Research Director’s Division Because of the proposed modifi- cations in the ARE design, work on the control system for this reactor has necessarily slowed down. However, satisfactory design of most of the components of the control system had already been attained.. The high- temperature fission chamber failed at T00°F and is being redesigned. Develop- ment of the reactor dynamic computor has continued. 11 ANP PROJECT QUARTERLY PROGRESS REPORT Control System. Detailing of the control rcom, console, 1nstrument racks, and interconnecting cables has continued. Discussion of the manu- facture of the hardware incorporating these designs was held with possible outside vendors. Several companies have indicated willingness to contract for these 1tems, have been made. but no commitments High-Temperature Fission Chamber. The high-temperature fission chamber was completed and tested. It failed at about 700°F because the insulators were inadequate. A new chamber, correcting thisdifficulty, 1s presently under construction and will be ready for test in about six weeks. No difficulty 1s anticipated in providing a satisfactory fission chamber for use in the ARE. Reactor Dynamic Computer. The reactor dynamic computer has progressed to a point where 1t will be ready for use in about six weeks. The converter, from d-c to digital notation, and the multiplier have been tested and debugged. INSTRUMENTATION About 5% of the instruments required for the ARE have been received, and an additional 75% are on order and should be received within four months. Ten percent will probably be constructed by the Laboratory because of unavaila- bility from industrial concermns. The remaining 10% are in a deferred status 12 pending the outcome of experimental engineering investigations currently in progress. REMOTE ~-HANDLING EQUYPHMENT Remote-handling equipment design 1is in process, and some components have been ordered. Demonstrations have been made simulating installation, removal, and reinstallation of the core dome, involving a sequence of welding, cutting, scarfing, and re- welding. These demonstrations will continue and will employ the automatic cutting equipment intended for the ARE. Remote removal of the core dome after power operation of the reactor will be accomplished by means of rubber-bonded abrasive cut-off wheels, operating under water, driven by a Vickers hydraulic system and guided by peripheral tracks. This equipment has been designed, some components have been requisitioned, and other components have been released to Y-12 shops for fabrication. ARE BUILDING FACILITY The building shell 1s virtually complete; walls and roofing are 1in place. Painting, plumbing, and internal partitioning are in process. ORNL engineering of special internal features is approximately 80% complete and should be ready for transmittal to Nicholson via AEC in approximately four weeks. FOR PERIOD ENDING DECEMBER 10, 1951 3. REACTOR PHYSICS Nicholas M. Smith, Significant studies by the ANP Physics Group have been the surveys of the circulating-liquid-fuel designs. These surveys have been conducted using bare-reactor theory (i.e., a condition under which space and lethargy variables are separable) and for several parametric variations. For a given moderator these parameters were core diameter and the volume ratio of circulating fuel to that of the moderator. The rather extensive bare-reactor calculations were made possible by having the calculation set up entirely on the Card Frograming Computor IBM equipment. Considered first are the values of these parameters leading to a minimum fissionable material in- ventory in the overall system. When the core diameter 1s fixed, 1t 1is seen that as the fuel-coolant volume fraction increases, the critical mass increases but that the uranium density in the fuel-coolant at first de-~ creases, then increases. The result is that there is a minimum in the uranium investment in the overall system. This minimum usually re- sults 1n a reactor with about half thermal fissions, 1.e., the borderline between the epithermal and the intermediate. It 15 known from kinetic studies that the ratio of fuel in the reactor to that in the overall system should be as high as is practical 1in order to conserve delayed neutrons. Thus, in comparison to the condition of minimum critical uranium mass, the condition of minimum overall uranium mass results in a higher reactor-to- system volume ratio and favors the kinetic behavior. The kinetic be- havior can be favored to an even greater extent by choosing a fuel- a reactor on- Jr., ANP Division coolant volume ratio greater than that resulting in a minimum uranium inventory. Since the curves of total uranium inventory vs. fuel-coolant volume ratio havea very broad minimum, it is possible to design for a greater fuel-volume ratio without greatly in- creasing the uranium requirement. Thus it becomes advisable to design the reactor with as high a fuel volume ratio as practical — say, one in- creasing the uranium requirement 5% over that of the minimum —~ in order to favor the kinetic response. The conditions of near-minimum uranium mass and of favorable kinetic behavior thus result in an increase of the neutron spectrum, yielding a reactor just barely in the intermediate class. The trend of the nuclear design to one of the intermediate class is thus inescapable. Very satisfactory agreement within theory and critical experiment 1s reported for intermediate uranium- graphite and uranium-water assemblies. The agreement gives corroboration to the explanation that the discrepancy between theory and experiment for the uranium-beryllium assembly is caused by omission of a significant physical phenomenon, possibly the (n,2n) re- action. REACTOR CALCULATIONSON IBM EQUIPHMENT F. C. Uffelman W. C. DeMarcus Phyllis Johnson Uranium Control and Computing Department The extensive bare-reactor calcu- lations were made possible by having the calculations set up entirely on the CPC-IBM equipment. During the past quarter the IBM equipment was i3 ANP PROJECT QUARTERLY PROGRESS REPORT also employed in the calculation of cross-sections for cores and re- flectors and kinetic calculations of various reactors. The division of effort 1s 1temized as follows: 1. Reactors. During the period from September 1 to November 16, 1951, the IBM section com- pleted calculations on 29 EPLA (end-point linear approximation) reflected reactors, the total to 211 since the start in February. Programing was 1nitiated and completed for twe series of bare-reactor cal- culations, hydrogenous nonhydrogenous. Twenty-nine hydrogenous and 43 nonhydrog- enous calculations were made. bringing and 2. Cores and Reflectors. During this same period average cross- sections were calculated for 117 cores and reflectors, and "constant" variations were made for two cores and reflectors. 3. Kinetic Calculations. Programing completed for the dual space-time varilable kinetic calculations and three such calculations were made, one for time running to 15 sec and the other two only up to 4 sec each. One calculation, with time the only variable, was also per- formed. was CIRCULATING-FUEL REACTORS C. B. Mills, ANP Division The circulating-fuel reactor under study consists simply of a spherical moderating core with circular fuel- coolant tubes penetrating the length of the core into the reflector. The reflector is a spherical shell of core modearator material of 6 in. re- flector savings. The moderator 1is expected to be either beryllium oxide 14 If BeO 1s utilized as a moderator, 1its cooling will be provided by the circulating fuel itself, making the "reflector" also a neutron multiplying region. If water is so employed, 1t will be 1nsulated from the circulating fuel by a thin layer of, say. Si0,, and the moderator cooling will be accomplished by circulating the water 1n a secondary system, with 1ts separate heat ex- changer. or light water. A series calculations have been made to determine optimum size and composition. Reflector savings were included by adding a constant 15.24 cm of core material plus 1.13 ¢cm of extrapolation distance to the variable core radius. A series of a maximum of 27 calculations 1s suffi- cient to determine the optimum size and shape with the variable of re- actor core diameter, volume fraction of fuel-coolant or moderator, mass, and total uranium Thermal base effects, median energy for fission, and neutron absorption, fission leakage, and flux lethargy (or energy) distributions are obtained in the same group of calculations. The reactor design values used 1n this series are given in Table 3.1. Re0-Hoderated Reactor. of bare-reactor uranium inventory. Table 3.1 Design Values for the Be@-Moderated Circulating-fuel Reactor CORE COMPONENTS VOLUME FRACTIONS Fuel-coolant 0.10 0.20 0. 30 0.35 |0.40 (NaF-UF;) Structure 0.015| 0.015] 0.015 1 0.015(0.015 (1inconel) Moderator (BeD) 0.855] 0.755| 0.655 {0.650(0.555 Yoid (assumed) 0.03 | 0.03 { 0.03 |0.03 [0.03 FOR PERIOD ENDING DECEMBER 16, 1951 These volume fractions gilven 1in Table 3.1 were used for threes reactor core diameters: 3, 3%, and 4 {t. The BeQ reflector thickness was 6 1in. The uranium mass reactivity coefii- cient, Nk/k T Am/m, was examined 1in the process of the determination of the minimum critical mass. The value of the coefficient decreased by about 20% in changing the volume coolant fraction of the coolant from C.1 to uranium inventory as a fanction of volume fraction fuel-coolant are given in Fig. 3.1 for 3-, 3%-, and 4-ft reactor cores. The total uranium inventory was a minimum of 66 1b at 0.28 volume fraction of fuel 1n the 4-fr-diameter core. The total uranium inventory with the fuel-coolant holdup in cubic feet 1n the heat ex- changer as a parameter is given 1in Fig. 3.2 as a function of reactor 0.4. The critical mass and total core diameter length. SECRET ANP-PHY-223 DWG, 13565 120 - 100 \ REFLEGCTOR SAVINGS =6in. FUEL-COOLANT: NaF-UFg4 EXTERNAL SYSTEM VOLUME T T T T T IS ASSUMED TEO BIE 15|fi3 : LN \\ - 3-ft DIAMETER TEJTAL S 80 D - i URAN UM — e ‘\\W\\ 3/é$f DIAMETER INVENTORY W > cqn S = 4-ft DIAMETER = 60 < = = < € 40 4(fleAMETER URANIUM / . -3 /o-ft DIAMETER I N — /fif’ 3-ft DIAMETER] CORE e O . 0 040 0.20 0.30 0.40 Q.50 0.60 0.70 VOLUME FRACTION FUEL-COOLANT IN CORE Fig. 3.1, Total Uranium Inventory and Uranium in Core as a Function of the Volume Fraction Fuel Coolant in the Core for BeO-Moderated €Circulating-Fuel Reactors with 3-, 3%-, and 4-ft-Diameter Cores. 15 ANP PROJECT QUARTERLY PROGRESS REPORT SECRET ANP-PHY-224 DWG. 13566 1 ! T T T ] - 40 AN FUEL-COOLANT VOLUME FRAGCTION=035 uly30 ™ FUEL-COOLANT: NaF -UF4 = N o ] 1ip] R, zZ \\\; \\N‘Nmm —~ \\ 30 112 = _ 5100 <~ - 3 Z 90 \\_\ - 25 ft HOLDUP VOLUME _| a - x e OUTSIDE THE |, 80 20 f1° REACTOR GCORE 2 of — O Nh""""*—-«....,_,_____ ] 3 T o 15 f17 v 1l 50 * 30 3.25 35 3.75 4.0 4.25 4.5 4.75 REACTOR CORE DIAMETER (ft) Fig. 3.2. Total Uranium Investment in Bed-Moderated Circulating-Fuel Reactor as a Function of Reactor Core Diameter for Several Assumed Holdip Volumes Externa}l to Core, The percent thermal fissions as a function of fuel-coolant volume per- centage 1s shown in Fig. 3.3 for the 3- and 4-ft-diameter cores. Apparently the ANP reactors are in the epithermal to intermediate energy range of neutron energy distribution. Fis- sioning, absorption, and leakage spectra for the 3%-ft-diameter core are given in Figs. 3.4, 3.5, and 3.6, respectively. The percent thermal fissions for this reactor 1s 59, the percent thermal absorption 1s 41.6, and the percent thermal escape 1s 8§, all at lethargy of 18.6. The neutron flux distribution (Fig. 3.7) is high 16 and rather uniform over all lethargy intervals above thermal. The poisoning effect of stainless steel for this reactor been estimalted as 6% change in k__. . for a change in volume fraction of stain- less steel of 0.01 in the vicinity of a total volume fraction of 0.028. has The set of data 1s not yet suffi- ciently complete to do more than specify the region of design interest. Apparently the reactor will have approximately 35% fuel-coolant and 65% BeD moderator, with a structure FOR PERIOD ENDING DECEMBER 10, 1951 SECRET ANP-PMHY-225" OWG. 135667 100 90 Y 8o § -\~§§: /p~4‘fl-COREIMAMETER “ 70 P ® ~ “ 60 S~ o NN O N Q O 0.08 0.06 0.C4 0.02 O SECRET ANP-PHY-226 DWG. 13568 COMPOSITION BY VOLUME FRACTION FUEL-COOLANT (NaF-UR, ) STRUCTURE (INCONEL) MODERATOR (BeO; VOID 59% THERMAL FISSION AT v=18.6 18 16 14 12 10 8 LETHARGY {v) 0.300 0.C15 0.655 0.030 Fig. 3.4. Fission Spectrum vs. Lethargy for the Be0D-Moderated Circulating- Fuel Reactor with a 3k-ft-Diameter (ore. LY0d34 SSHUI0Md ATHILVVYAD 123af0Md JINV 61 ABSORPTION PER UNIT LETHARGY PER 1, INITIAL NEUTRONS 0.36 G.32 0.28 0.24 0.20 0.46 0.12 0.08 0.04 SECRET ANP-PHY -227 DWG, 13569 COMPOSITION BY VOLUME FRACTION FUEL-COOLANT (NaF-UF ) 0.300 STRUCTURE (INCONEL) 0.015 MODERATOR (BeO) 0.655 VOID 0.030 44697 THERMAL ABSORPTION Al w=18.6 18 16 14 12 10 8 6 4 2 0 LETHARGY () Fig. 3.5. Absorption Spectrum vs. Lethargy for the BeO-Moderated Cir- culating-Fuel Reactor with a 3%-ft-Diameter Core. IS6T ‘0T HAGWADHAG INIGNT AOIWAd HO4 0 0.18 0.6 0.4 02 040 0.08 0.06 0.04 0.02 LEAKAGE PER UNIT LETHARGY PER v, INITIAL NEUTRONS Fig. 3.8. Leakage Spectrum vs. COMPOSITION BY VOLUME FRACTION FUEL-COOLANT (NaF-UF,) 0.300 STRUCTURE (INCONEL) 0.015 MODERATOR (BeO) 0.655 VOID | 0.030 i8 16 14 12 10 8 6 LETHARGY () Fuel Reactor with a 3%-ft-Diameter Core, SECRET ANP - PHY-228 DWG, 13570 Lethargy for the BeO-Moderated Circalating- L1H043Y SSIU90Hd ATHRLYVND 1JAfodd dNV ié SECRET ANP-PHY-229 DWG. 13574 COMPOSITION BY VOLUME FRACTIONS FUEL-COOLANT (NaF-UF,) 0.300 STRUCTURE (INCONEL) 0.015 MODERATOR (BeO) 0.655 37 VOID 0.030 RMAL AT LETHARGY ‘ u =86 NEUTRON FLUX ¢ {(¢) PER 7, INITIAL NEUTRON 18 e i4 12 {0 8 6 4 2 O LETHARGY () Fig, 3.7. Neutron Flux Spectrum vs. Lethargy for the BeO-Moderated Cir- culating-Fuel Renctor with a 3%-ft-Diameter Core. IS6T ‘O HIAWFDAG SNIONZ QOIVAd HOd ANP PROJECT QUARTERLY PROGRESS REPORT 160 SECRET ANP—PHY~23! DWG,. 13572 | 1 LIGHT WATER , DENSITY=09667 g/cc 140 \\\EXTERNAL VOLUME =15 ft. | FUEL-COOLANT : NaF-UF,4 | | 120 \ o N o 100 & N TOTAL URANIUM INVENTORY = \\ 80 s = P < 60 (el o 40 { URANIUM IN THE CORE 20 0 O {0 20 30 40 50 VOLUME FRACTION PERCENT COOLANT IN CORE Fig. 3.8. Total Uranium Inventory and Yranium in the Coreas a Function of the Volume Fraction Fuel-Coolant in the 3-ft-Diameter~Core Water- Moderated Reactor with NaF-UF4 Fuel-Coolant, BeF,-NaF-UF, Fuel Coolant. Re- actor design values for this coolant in the water-moderated core are given in Table 3.3 and are applicable to each of the 2%-, 3-, and 3Y%-ft-dia- meter cores. The uranium mass reactivity coeffi- cient, Ak/k + Am/m, decreased by only 30% for values of volume fraction of fuel-coolant in the core from 0.3 to 0.8. The critical mass and the total uranium inventory 1in the core as a function of the volume fraction fuel- coolant are given in Fig. 3.9 {for a 2%-, 3-, and 3%-ft-diameter reactor 22 Table 3.3 Design Values for the Water-Hoderated Circulating-Fuel Reactors for Use with the Na¥F-Bel, and LiF-KF-NaF Fuel-Coolants CORE COMPONENTS VOLUME FRACTIONS Fuel-coolant Moderator (H20) Structure (inconel) Void (assumed) 0.30 0.6038 0.0100 0.0862 0.45 0.4681 0.0150 0.0669 0.60 0.3325 0.0200 0.0475 0.80 0.1255 0.027 0.0495 FOR PERIOD ENDING DECEMBER 16, 19531 SEGRET ANP—PHY-235 120 DWG. {3573 EXTERNAL SYSTEM VOLUME 15 #° FUEL-COOLANT Be F2~NGF-UF4 WATER TEMPERATURE 1i83° F 100 LIGHT WATER ; DENSITY=0.9667 g/cc 24 ft | 80 :\\‘1‘ B =~ : TOTAL URANIUM |— ~ 4 £ \ L3V ft INVESTMENT N 23 & D Vel " w \\‘~mw_"1j;/7 ™ = 2 = = 3l ¢ > 40 x 420 12 1 DA, -~ .~ |43 ft DA URANIUM - // le2' 1t DA, j N CORE 20 e R B oy O : O 0.20 .40 060 0.80 100 1.30 .50 VOLUME FRAGTION FUEL~ COOLANT Fig. 3.9. Total Uranmium Inwentnry and Urapium in {ore 28 a3 Function of the Volume Fraction ¥Fuel-Coolant in the Core for the Water-Moderated Circulating- Fuel Reactor with 2%-, 3-, core. The total uranium inventory was a minimum of 60 1lb at 0.60 volume fraction of fuel in the 3-ft core. The minimum critical mass with water moderation 1s realized with a smaller core than that required with beryllium oxide moderation, primarily because water 15 a better moderator. ' absorption, and leakage the 3-{t diameter Fissioning, spectra for all and 3%-~-ft-Diameter Cores. water-moderator cores with 0.60 volume fraction fuel-coolant are given in Figs. 3.10, 3.11, and 3.12, respec- tively. The percent thermal fissions for this reactor 1s 86.4, the percent thermal absorption 1s 86.4, and the percent thermal escape 1s 14.6, all at a lethargy of 19.6. This 1s, therefcre, an epithermal reactor. The neutron-flux distribution (Fig. 3.13) has a high peak in the high- energy (low-lethargy) range. Ve FISSIONS PER UNIT LETHARGY PER 7, FISSION NEUTRONS 018 016 012 010 0.08 0.06 0.04 0.02 SECRET ANP—PHY —244 DWG, 13574 OMPOSITION OF CORE BY VOLUME FRACTION FUEL-COOLANT (NaF - BeF, - UF,) 06000 MODERATOR (H,0) 0.3325 STRUCTURE (INCONEL) 0.0200 VOID | 0.0475 864 % THERMAL FISSIONS AT - v=196 18 16 4 12 10 8 6 4 2 O LETHARGY () Fig., 3.10¢. Fission Spectrum vs. Lethargy for the flzo-Moderated Circulating- Fuel Reactor with 3-ft-Diameter Core. 1H043Y SSAUT0Ud ATHALEVNO LDALOUd dNV Se ABSORPTIONS PER UNIT LETHARGY PER Ve FISSION NEUTRONS SECRET ANP—-PHY-242 0.36 DWG, 13575 032 OMPOSITION OF CORE BY VOLUME FRACTION 0.28 FUEL-COOLANT (NaF - BeF, - UF,) 06000 MODERATOR (H,0) 0.3325 STRUCTURE (INCONEL) 0.0200 VOID 0.0475 0.24 0.20 0l6 012 87 % THERMAL 008 ABSORPTION AT v=19.86 0.04 18 6 4 12 {0 8 6 4 2 O LETHARGY (u) Fig. 3.11. Absorption Spectrum vs. Lethargy for the H20=Moderated Cir- culating-Fuel Reactor with 3-ft-Diameter Core, IS6T ‘0T WIIWADHG INIANT aoryiad Ho4 9¢ LEAKAGE PER UNIT LETHARGY PER v, INITIAL NEUTRONS 0.18 0.16 G114 Q12 010 0.08 C.06 0.04 C.02 SECRET ANP—-PHY-243 DWG, 13576 ! T fi P J T Y | | 7 L | EEEEEEEEE | *% 4‘ _I J 1 { T % COMPOSITION OF CORE BY VOLUME FRACTION ZL FUEL-COOLANT (NoF~BeF - UF )Y 08000 ; MODERATOR (H,0) 0.3325 | |STRUCTURE (INCONEL) 0.0200 - WOoID j 0.0475 | T % | ) | | | | - ; [ | L | - — — = + i [ I S i' — | ' 46 % - = THERMAL ] ESCAPE | — S | | — ATu=196-4 | ! —— — | T R ! | i8 16 14 12 {0 8 4 2 LETHARGY () Fig., 3.12. Leakage Spectrum vs. Lethargy for the HQQ-MO&erated Circulating- Fuel Reactor with 3-ft-Diameter Core, 140438 SSTUO0HA ATHILYYNO LDAf0Ud JNV La NEUTRON FLUX ¢ (¢) PER UNIT LETHARGY PER 7, FISSION NEUTRONS SECRET ANP-PHY-244 DWG, 13577 COMPOSITION OF CORE BY VOLUME FRACTION FUEL— COOLANT (NaF -BeF, - UF,} 06000 MODERATOR (H,0) 0.3325 STRUCTURE (INCONEL) 0.0200 VOID 0.0475 26 % THERMAL FLUX AT v =196 18 16 14 12 10 8 6 4 2 0 LETHARGY (¢) Fig. 3.13. Neutron Flux Spectrum vs. Lethargy for the Hzo-Moderated Cir- culagting-Fuel Reactor with 3-ft-Diameter Core, TS6Y ‘01 HIAGWADAA ONIOGNA GOIVAdL HO4 ANP PROJECT QUARTERLY PROGRESS REPORT The fuel-coolant density value used for this designset was 2.1 g/em’. Evaluation of the fuel-coolant density reactivity coefficient as well as of the reactivity coefficients for structure and moderator will be com- pleted soon. An estimate of the water-moderator temperature coeffi- cient, made to determine the effect of water cooling stability on re- activity, was found to be: The stainless steel poisoning effect was determined as a loss of 2.55% in k_., per 0.0] increase in volume fraction of stainless steel. A change from 0.015 to 0.030 in stainless steel volume fraction changed k_,, 1in the water-moderated reactor with 35% fuel-coolant from 1.024 to 0.987. Naf-KF-LiF-UF, Fuel Coolant. The reactor design values for this coolant in the water -moderated reactor core Ak/k i k : “Zé" =-7.8 X 10°% per degree Fahrenheit.zf: ;?;i same as those given in Ta- SECRET ANP~PHY— 239 DWG. 13578 200 S I B LiF-NaoF-KF-UF4 FUEL COOLANT VOLUME OF THE SYSTEM OUTSIDE | . THE CORE: 5.0 ft° 160 b—n LIGHT WATER: DENS!TY=Q.9/667 e g/cc i _ Z3 > ft DIAMETER TOTAL o ] —. = ~ e e _4«-2%ft DIAMETERURANIUM w T~ *:_-;-q;‘:i.m:_; o “‘"\ INVEN TORY 2120 3 ft DIAMETER Y | = | | = ‘ - = % 80 > 3% ft DIAMETER AT URANIUM bt _ 1 3 ft DIAMETER | N 40 ST o 2% ft DIAMETER CORE ™ e Ffl_flfl_fl,,+w-” 2 O 0.20 0.30 040 050 060 Q70 080 090 VOLUME FRACTION FUEL-COOLANT IN GCORE Fig. 3.14. Total Uranium Inventory and Uranium in Core 28 2 Fuanction of Volume Fractionm Fuel-Coclant in the Core for Hzfi-MQGerated Circulating-Fuel Reactors with 2%-, 3-, and 3%-ft-Diameter Cores and LiF-NaF-KF-UF, fuel-Coolant. 28 FOR PERIOD ENDING DECEMBER 10, 1951 Figure 3.14 presents the summary graphs for the system, showing critical mass values for the cores and in- dicating the minimum uranium require- ments for the systems to be 127 lb for the 2¥%-ft-diameter core, and 122 1b for the 3- and 3¥%-ft-diameter cores. The optimum volume fraction for fuel- coolant in the core is mear 0.50, with a very small sensitivity to this parameter indicated by the small rate of change between 0.30 and 0.60 volume fractions. This reactor has approximately 80% thermal fissions. Comparison of H,0- and BeD-Moderated Reactors. These reactor series serve to indicate the regions of interest for possible designs. No attempt has been made to compute any one design in detail because of the sensitivaty of reactivity calculations to shape and constituents. The relative ac- curacy of the results depends not only on the usuvual assumptions but also on the proximityof the estimated physical constants {(such as the fuel-coolant density) to their actual values, as well as on the structural details. Some of the differences between the water-moderated and the Be0O-moderated reactors are given in Table 3.4. More detailed comparisons may be made directly from the various spectra that are presented for both types of reactor. In addition to the i1tems compared in Table 3.4, the poisoning effect of stainless steel is somewhat worse for the BeO-moderated reactor. On the other hand, there is a great deal more escape of fast neutrons from the H,0-moderated reactor. However, no important "physical"difference between the reactors has appeared to date, so that design considerations must be given the greater weight 1in choice of a reactor type. Attention 1is called to the large increase in uranium inventory caused by the use Table 3.4 Comparison of Be(- and Hzo-Moderated Reactor ITEM COMPARED BeO REACTOR H,0 REACTOR Fuel NaF-UF, NaF-BeF,-UF, Core diameter (ft) 3.5 to 4.0 3 Fuel-coolant volume fraction 0.36 0.71 Critical mass(® (1b) 26 to 30 26 External fuel volume (estimated) (ft3) 15 15 Total uranium inventory(®) (1bh) 69 to 72 63 Ratio of fuel in core to total fuel 0.37 to 0.42 D.41 Percent thermal fissions 52 to §5 69 MOSEZ:E?r dg;sity-temperaturg reactivity coeffi- 0.6 x 10°° 7.8 x 10°° At 183°F -9.93 x 10°°F -14 x 1075 At 1500°F -9.85 x 10°° (e necessary heterogencity of structure will increase these values sowewhat. B3 N3 ANP PROJECT QUARTERLY PROGRESS REPORT of large amounts of potassium in the fuel-coolant, It is also very im- portant to minimize the structure volume fraction in any of the above reactors. Effect of Delayed Neutrons on Kinetic Response (C. B. Mills, ANP Division). The importance of delayed neutrons for controlling the speed of response of a stationary-liqguid-fuel ARE (Aircraft Reactor Experiment) was determined by repeating a calculation in which an increase in inlet coolant temperature of 25°C is 1introduced stepwise and then maintained and a control rod is actuated to 1insert or remove 0.00025 unit of reactivity per second. Two-thirds and four-fifths of the delayed neutrons were removed from the kinetic response equations, and the twe curves for (¢ - ¢0)/¢0 vs. time were compared with the curve for no loss of delayed neutrons. The results in Fig. 3.15 1ndicate a rate of change of flux three times as great as normal when the fraction of delayed neutrons is decreased by 80%. For a 67% decrease the rate of change of flux 1is almost twice as great as normal. SECRET ANP—PHY-246 g 3600 DWG. 13579 @ NITIAL SLOPE OF THE FLUX VS. TIME CURVES — ¢-¢ d 0 PERCENT LOSS X 3200 4, OF DELAYED g 1 NE UTRONS E:¢ 2800 00 80 < O 0.07 67 =i 0.03 0 - . 2400 ™ wJ L 5 80% LOSS OF . = 2000 DELAYED v > NEUTRONS 7% LOSS OF D ooz DELAYED - © 1600 NEUTRONS - 800 0% LOSS OF < DEL AYED tl 400 NEUTRONS - T e - 0 1.0 2.0 3.0 40 TIME (sec) Fig. 3.15. 67, 30 Fractional Change im Neutron Flux and 88% Decrease in Delayed Neutrons. as a Function of Time for 0, FOR PERIOD ENDING DECEMBER 10, 1951 These results were then extrapolated to the case of kinetic response of the circulacing-fuel reactor. The factors i1im the kinetic calculations described above areabout the same for the stationary and the circulating fuel reactors with the single ex- ception of fuel expansion in the fuel tubes. However, for times less than 0.5 sec there 15 very little daif- ference in the rate of change of flux between the case with fuel expansion, as it occurs in the stationary fuel and the case with no expansion. 1t appears therefore thatif a disburbance such as the one specified above is introduced, the kinetic response within 1 sec thereafter 1s not sensi- tive to fuel expansion and the circu- lating-fuel-reactor flux will follow a rate-of-change curve very similar to that shown on Fig. 3.15. The loss 0of delayed nentrons from the core up to 80% of these there for the stationary-fuel case should not in- crease the kinetic response rate of the liquid fuel reactor by more than a factor of three times that for the reactors with stationary fuel for which kinetic response data are available. CIRCULATING-MODERATOR REACTORS B. T. Macauley C. B. Mills ANP Division Tentative designs for alkali hy- droxide ~~moderated reactors been computed to determine the re- actor criticality constants. These are the 200- and 400-Mw sodium hy- droxide —moderated and -— cooled reactors. The design data are given in Table 3.5. have Reactor criticality constants for the 200~ and 400-Mw sodium hydroxide reactor and for a 200-Mw potassium hydroxide reactor are given in Ta- ble 3.6. Table 3.5 The HadB-CTooled and ~ Moderated Reactor (July 12, 1951) Reactor power (Mw) 200 400 Reactor core diameter {ft) 24 K4 RBeactor core material {volume fractions) Stationary liquid fuel 0.0477 0.0148 (NaF-UF4¥ Fuel~-element structure 0.1766 0.1252 Baffle structure 0.0257 Moderator~-coolant 0.7500 0.8600 Apparently a larger reactor will be required if 1t 1s necessary to use KOH. However, the critical mass for a 3k-ft-diameter KOH-umoderated reactor is only 70 1b. The higher critical mass of the smaller core (see table} i1s mecessitated by the greater leakage of that core. More complete data and analysis of these hydroxide reactors will be presented in the next report. Figure 3.16 shows the sensitiveness of this reactor to the stainless steel used for structure. Evidently stain- less steel 1s an 1mportant reactor poison. ERRORS IN REACTOR PHYSICS CALCULATIONS C. B. Mills, ANP Division The value of the present method¢(!'?? of reactor calculations lies in the large amount of accurate information {1)p, K. Holmes, The Muitififigug Method as U;gd)by the ANP Physics Group, -38 (Feb. 15, 1951). M. J. Nielsen, Bare Pile Adjoint Solu- téggj ORNL, Y-12 site, Beport Y-¥F10-18 {Qct. 27, 19507, ANP PROJECT QUARTERLY PROGRESS REPORT Table 3.6 Reactivity Calculation Resulis on the Alkali Hydroxide Reacters NaOH, 200 Mw NaOH, 400 Mw KOH, 200 Mw Core diameter (ft) 24 3% 2% Critical mass (1b) 52 77 220 Ne/k S .47 .48 0.065 (at 60 1b mass) (Dm/m ) 235 Percent thermal fissions 55.5 17.1 42.9 Median energy for fission (ev) 0.04 Thermal 0.15 Ak/k - . 209 (Ap/p) moderator Dk/E —_—rT T -0.221 (AP/P) struc ture Ak/fk - = radius) 0.5347 AR/R SECRET obtained, 1its speed, and 1ts rel- ANF—-PHY 245 . | 20 OWG. 13580 atively small cost. Almost any re- ' \\\\ actor can be computed by this method \\\ 1if the cross sections of the con- stituents are known. 110 - ™ \ \\\ Results of a comparison between S uncorrected bare-reactor calculations .00 - : | . and experiment show a surprising consistency for the very simple COMPONENTS BY VOLUME FRACTION EFFECTIVE MULTIPLICATION CONSTANT 0307 LIQWD FUEL (NaF-UFg) 0.0477 ~ FUEL ELEMENT STRUCTURE (5.5)0.1766 [~ BAFFLE STRUCTURE (S.S.) 0.0257 | MODERATOR=- COOLANT 0.7500 0.80 0 005 040 015 020 0.25 VOLUME FRACTION STAINLESS STEEL IN THE GORE Fig. 3.16. Effective Multiplication Constant as a Function of Volume Fraction of Stainless Steel in the Core for the 2%-ft-Diameter-Core NaQH-Moderated and - Cooled Reactor. 32 multigroup method used. Consistency of multiplication constant deter- mination is about *1% with an average error of -2% for water-moderated reactors of any size of uranium con- -1% for the moderated reactor experiment, and -10% for the beryllium-moderated critical experiment. The error 1n the beryllium calculation is beang investigated in some detail and seems to be due to the omission of a signif- icant physical effect [the (n,2n) reaction, for instance] and to known centration, carbon- FOR PERIOD ENDING DECEMBER 10, 1951 Table 3.7 Critical Assembly Sizes, Composition, and Calculated Multiplication Constants Water-Moderated Reactors {(a Right Cylinder, with Extreme Variation in NH/NU) CALCULATION DIAMETER HEIGHT MASS OF U233 k., {core No.) {cm) (cm) Ny/Ny23s (kg) CALCULATED Reflected 106 25. 40 22. 40 328.7 0.893 0.970 RBeflected 107 38.10 44.30 999.90 1.31 0.988 Bare 1089 25.40 34.00 52.9 7.90 0.984 Bare 110 25.40 32.30 43.9 8.80 0.985 Carbon Critical Experiment (a Cube) CALCULATION AREA LENGTH | . MAsS OF U*®® k5 {core No.) (in. x in.) (in.) Ny2as | N ”;1 (kg) CALCULATED Bare 146 51 x 51 44.111 0.725 770 | 36.8 44.6 0.991 Beryllium Critical Experiment CALCULATTON AREA LENGTH | Njjas (Ng, |Na, | Mass oF u?%° kg (core No.) (in. x in.) (in.) (kg) CALCULATED Bare 60 21 x 21 23.22 3.08 1074 .36.5 19.3 0.903 *x 10°%° per cec. simplifications in the conservation equations. Reactors that have had both critical-experiment test and bare-reactor calculations are de- scribed in Table 3.7. The calculations on the four water- moderated reactors were based on the theory of Goertzel and Selengut(?®) described in the last quarterly re- port.¢*> Two bare and two reflected (3)G. Goertzel, Criticalit { drog en- Moderated Reactors, ORNL, ANP, TAB-53 uly 25, 1950) . 43¢, B. Mills, * The Sodium Hydroxide Re- actor,” Aircraft Nuclear Propuls;on Project Quarterly Progress Report for Period Ending ?; {fmber 10, 1951, 0 L-1154, p. 70 (Dec. 17, critical experiments were computed using the modified age equations described in detail elsewhere.(%? The two reflected-reactor calculations used a correction on the core radius equal to the slowing-down length of neutrons in water. Bare-reactor multigroup calcula- tions were made on the four reactors with the multiplication constant (keff) resulting as shown in Ta- ble 3.7. Note that these reactors were selected to give the worst 533, w. Nebster, Numerical Technique for Criticality Calculations on Hydrogen-Moderated fleacters, OBNL, Y-12 site, Beport Y-F1l0-466 (Aug. 20, 1951). 33 ANMP PROJECT QUARTERLY PROGRESS REPORT possible case for results, as 1s an- dicated by the large changes in critical mass and atomic density of U?35, The experiments are described in detail in report K-343.(%®) It was necessary to use a variable buckling [B?*(u)] in the age equation of the small size of the Constant buckling results -0.0208. because reactors. were lower by Ak The relative importance of hy- drogen moderation was determined by ignoring moderation by all other (%¢. b. Beck, A. D. Callihan, J. W. Merfite, and R. L. Murrey, Critical ¥ass Studies. Part III, C & OCC Beport K-343 (Apr. 19, 1949}, 34 atomic constituvents. This lowered k,,. of Ok = -0.0247. A Fermi age equation solution of this problem without consideration of the effects of hydrogen moderation separately results k .o of Ak ., = 0.192. in an error 1n ff The details of the leakage and fissioning lethargy distribution were presented previously.(7) Note that the percent thermal fission values change from 94% for core 107 to 41.5% for core 110, 1.e., from thermal to intermediate reactor fission distri- butions for equally good results. (7)Figures 3.26 and 3.27, ORMi.-1154, op. cit., pp- ¥5 and 76. FOR PERIOD ENDING DECEMBER 10, 1951 4. CRITICAL EXPERIMENTS A. D. Callihan, Physics Division A second wmockup of the General Electric direct-cycle (air-water cycle) reactor was completed, and the critical mass of the assembly was found to be 90 1b of uranium. The inhomogeneities of the mockup are serious, however, and the extrapo- lations of the critical mass of this assembly to that of the aircraft reactor are not yet reliable. Considerable data have been taken on the uranium graphite assembly, but the results have not been reduced. PRELIMINARY ASSEMBLY OF DIRECT-CYCLE REACTOR At the request of the General Electric Company and with its co- operation, studies are being made of the proposed direct-cycle nuclear reactor for aircraft propulsion. This reactor, as presently conceived, will be air cooled and water moderated, with stainless steel as one of the principal structural materials, Beryllium will be the reflector. One series of experiments was outlined in the preceding report(!) and has been described more fully elsewhere,{(?) During the period reported here a second mockup has been assembled and made critical. The core is an approxi- mate right cylinder, 36 in. long and 51 in. in diameter, the periphery of which i1s not smooth because the (1)a. p. Callihan, “Critical Experiments,”’ Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1351, ORNL-1154, p. 79 (Dec. 17, 1951). : (233, A. Hunter, Report of Critical Experi- ments for a Water Moderator, CA-2, G.E. Report DC-51-9-11 (Sept. 12, 1951). smallest structural unit is 1% in. square. The axis of the cvlinder is horizontal. The mid-cross-section is shown in Fig. 4.1. The hydrogenous material, a methacrylate plastic, is placed in horizontal layers, 1 in. thick. These layers are separated by an open structure of stainless steel and uranium about 2 in. thick. This structure consists of six sheets of steel, each 0.017 in. thick and 36 in. Jong. The edges of each piece are bent under in order to separate adjacent stacked sheets by about 0.25 in., simulating the air ducts of the reactor. The upper three of each group of six vertically stacked pieces are inverted, 1.e., the plane surfaces of the center pair are in contact. The 3-in,- diameter 0.010-in.-thickenriched uranium metal disks are placed, with planes horizontal, in the interspace between the central steel sheets. {In some instances, because of insuf- ficient large disks, four 1¥%-in.- diameter disks are used.) The center- to-center spacing between disks 1s 3.6 1n. 1n the direction of the cylinder axis and 3 in. in the hori- zontal direction perpendicular to the axis, the latter being fixed by the dimensions of the square aluminum tube matrix into which the reactor com- ponents are built. The lateral sur- face of the core is enclosed by a 6-1in.-thick reflector of beryllium; the ends of the reactor, including the beryllium jacket, have a graphite reflector 6 in, thick, Insufficient available beryllium necessitated the substitution of graphite as the end reflector, The void fraction in the core is 0.57 and the H/U?*3° atomic ratio is 225. The fuel and moderator inhomogeneities are seriocus. The critical mass 1s 41 kg of uranium or 38 kg of U?3%, 35 ANP PROJECT QUARTERLY PROGRESS REPORT o - Iy == ——— S == o— - By ;fl;g Fig. 4.1. Mid-Cross-Section of Second Mockup of Direct-Cycle Reactor. 36 FOR PERIOD ENDING DECEMBER 10, 1951 GRAPHITE REACTOR In the time available prior to the inception of the program described above, a series of measurements was made on the uranium-graphite assembly described in an earlier quarterly The results, obtainable have not been re- report, (3) from these data, alized. (3icritical Assembly of Graphite Reactor,” ORNL-1154, op. cit., p. B80. 37 ANP PROJECT QUARTERLY PROGRESS REPORT 9. NUCLEAR MEASUREMENT Data for gamma-ray and neutron yields from proton bombardment of some materials has been extended using the 5-Mev Van de Graaff accelerator. This accelerator has also been used to measure the total cross-section of iron from 0.6 to 3.6 Mev and (p,n) thresholds for F'? and Na??. Upper limits have been set for the (n,2n) cross-sections 1n beryllium in an experiment i1n which a neutron source was surrounded witha beryllium sphere. Final adjustments are being made on the time-of-flight spectrometer for neutrons. THE 5-Mev VAN DE GRAAFF ACCELERATOR H. B. Willard, Physics Division The gamma-ray and neutron yields from the proton bombardment of Li7, Be®, and F!9 have been extended to approximately 5.4 Mev. These curves are to be printed in the Physics Division quarterly report for period ending December 20, 1951. The (p,n) thresholds for F!% and Na2? have been determined to be 4,253 £ 0.005 and 5.09]1 £ 0.010 Mev, respectively. The total cross-section of ordinary iron for fast neutrons has been measured from 0.6 to 3.6 Mev with a resolution of 35 kev. Unresolved levels occur every 100 kev, with indications that with better resolution many more would be observed. The average cross-section varies from 2.5 barns at 0.6 Mev to 3.2 barns at 3.6 Mev. These total cross-section data have not yet been reduced to specific cross-sections, which are used inshield calculations. 38 MEASUREMENT OF THE (n,2n) REACTION IN BERYLLIUM E. D. Klema, Physics Division The data on the (n,2n) reaction in beryllium discussed in the previous quarterly progress report{!) have been reduced by G. B. Arfken to an average (n,2n) cross-section for the spectra of the two sources used. The calcu- lations were made under the assumption that a neutron which makes any sort of collision [(n,n), (n,a), or (n,2n)] in the beryllium sphere is lost as far as producing the (n,2n) reaction 1is concerned, The absorption of neutrons by the beryllium sphere was estimated from the data obtained with the carbon spheres. The upper limits for the (n,2n) cross-sections as determined by this experiment are as follows: o.(n. 2n) Po-a-B source, < 0.25 barn Po-a-Be source, Sn,2n) < 0.56 barn The upper limit of this cross-section, although appreciably lower than that reported by Houtermans, (2) is still so high as to be capable of introduc- ing a significant error in the critical mass of a reactor assembly employing beryllium, In reality, this cross- section is the suspected cause of the discrepancy in the calculated (disre- garding the cross-section altogether) and actual critical mass of the (I)E. D. Klema, ‘“Measurement of the (n,2n) Reaction in Beryllium,’ Aircraft Nuclear Propulsion Project Quarterly Pragress Report for Period Ending September 10, 1951, OBNL-1154, p. 97 (Dec. 17, 1951). (DF, 6. Houtermans, *“On the (n,2n) Reaction in Beryllium with Neutrons of a Polonium-Beryllium Source,” Nachr. Akad. Wiss. Gottingen, Math. - physik. Klasse IIb Abt. p. 52 (1946), translated by W. K. Ergen, April 20, 1951 (Y-F20-12). FOR PERIOD ENDING DECEMBER 16, 1951 beryllium-reflected graphite-uranium critical assembly.(3) . TIME-OF-FLIGHT NEUTRON SPECTROMETER G. S. Pawlicki, ORINS E. C. Swith, Physics Division The spectrometer 1s essentially completed, and calibration aund adjust- ment of the i1nstrument are in progress. The best resolution to be esxpected with this rotor is 1.2 pusec per meter, which 1s comparable to that of the best operating spectrometers. Another rotor has been ordered whiech should improve this resolution by a factor of 2. Details of the apparatus will be found in the Physics Division quarterly reperts for June 20 and September 20. (3)(3. B. Mills and N. M. Swith, Jr., ‘“Calcu- lation for the Critice) Experiment,” Aircraft Nuclear Propulsion Project Quarterly Progress Repart for Period Ending June 10, 1951, ANP-65, p. 71 (Sept. 13, 1951). ANP PROJECT QUARTERLY PROGRESS REPORT 6. EXPERIMENTAL REACTOR ENGINEERING H. S. Savage, ANP Division The activities of the Experimental Engineering Group now center around the vigorous pursuit of the technology of both sodium and the fused fluorides to provide fundamental engineering data through the development of actual components: valves, pumps, seals, heat exchangers, etc. Consideration of the advantages of a circulating- fuel-coolant reactor for aircraft application necessitated a program for manufacturing fluorides and for determining their effects on container materials and reactor components. Approximately 400 1b of eutectic LiF-KF-NaF has been produced for test purposes. A pump incorporating a gas seal has satisfactorily pumped fused fluorides up to 1300°F for 35 hr. Also, fluoride transfer through small tubes has been repeatedly demonstrated, as has successful filtration through a 10-pu sintered stainless steel filter. Thus far, fluorides appear to reduce oxides inside the system and to be severely corrosive only in the presence of oxygen. Equipment is being assembled as rapidly as possible for conducting self-welding and pump-shaft-seal experiments with fluorides. Continued work with ligquid-metal systems during the quarter resulted in a pump incorporating a frozen sodium seal operating approximately 450 hr with sodium at temperatures up to 1100°F with a pressure across the seal of 25 psig. In other equipment, a seal test rig operated with sodium at 1500°F sealing successfully against 107 psig. Electromagnetic pump-cell development reached the stage at which sodium was pumped at temperatures up to 1300°F for approximately 160 hr; test-loop heater limitations prevented 40 the attainment of higher sodium temper- atures. Self-welding tests were continued with sodium, and equipment 1s being modified to continue work of this nature with fluorides. Valve testing was continued to evaluate certain commercially available varieties for use in liquid-metal systems. In some cases instrumentation needs for fluorides are radically different from those required for liquid-metal systems; this program has thus undergone change compatible with the changes 1in ARE. Null-type pressure- measuring devices were belng tested with both ligquid-metal and fluoride systems at the end of the quarter. Preliminary work on heat exchangers and analytical methods for sodium were continued during the period, as were cleaning procedures for both fluoride and ligquid-metal systems. Planning of full-scale ARE component testing facilities were being continued along only such lines as would allow the facilities to be used for testing either liquid-metal or fluoride systems. PUMP DEVELOPMENT The centrifugal pump employing a gas seal of graphite set against a flat sealing face has operated with NaK and subsequently with the fluoride eutectic LiF-KF-NaF. The latter test system was operated for 35 hrat 1300°F and was terminated because of a leak outside the pump. The design of the centrifugal pump of the above type for the liquid-metal (NaK) ARE has been completed. The 1V-hp canned-rotor pump, as rebuilt with high-temperature (500°F) windings, has been satis- factorily pretested with water, but some difficulty has arisen in canning FOR PERIOD ENDING DECEMBER 16, 1951 the rotor. The 6-hp canned-rotor pump operated poorly because of an inef- ficient suction inlet, which is being redesigned. The centrifugal pump with a frozen-sodium seal operated with no indication of seal failure for 450 hr with fluid temperatures up to 1100°F, Limited development of electromagnetic pumps has continued, and two new design cells have been tested. Centrifugal Pumps for Figure-Eight Loops (W. G. Cobb, ANP Division). The centrifugal pump for liquid metals(1) was operated on two occasions with NaK as the pumped fluid. 1In the first test pumping was continuous for 12 hr and the NaK temperatures reached 820°F be fore a leak necessitated a in a welded joint shutdown. The second test also was conducted with NaK, and temperatures reached 940°F during the 12-hr pumping period. Excessive gas leakage through the shaft packing and gasketed joint under the stuffing box necessitated termination of this test. The pump and system were cleaned, reassembled, and provided with heaters and insulation to allow preheating to approximately 1000°F. The system was then filled with the eutectic LiF-KF- NaF. Flow, metered by a sharp-edged orifice with pressures being indicated by gas trapped in bulbs connected to the orifice taps, was approximately 4.8 ft/sec (about 6 gpm). Level control was accomplished by means of electrical probes and gas pressure directly in the pump body. The test system ran continuously for 35 hr, pumping at temperatures up to 1300°F. Considerable encouragement was derived from the fact that gas leakage through the shaft seal was negligible through- out the run, and a2 small but definite (1)“C€ntrifugal Pumps for Figure-Fight Loops,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending Junz 10, 1951, ANP-§5, p. 168 (Sept. 13, 135i). signal was observed from the electro- magnetic flowmeter. Moreover, smooth operation was obtained with a gas-held level control. Although several small leaks in the system developed, omne of which eventually caused termination, the pump itself operated satisfactorily throughout the run. ARE Centrifugal Pump Design. (W. G. Cobb, L. V., Wilson, and J. F. Haines, Consultant, ANP Division). The design and detailing of a centrifugal pump employing NaK for the ARE has been completed, General design features are compatible with those listed previously(?? with the exception of level-control features. Avoiding the necessity for simultaneously develop- ing a satisfactory pump and a level control device appeared advisable; this pump is therefore to be located in the fluid circuit on the same level as the expansion tank. Therefore, the ligquid level in the pump 1s determined by the level in the expansion tank, and sufficient shaft overhang has been incorporated to compensate for level variations occurring in the surge tank. A gas seal of graphite against a flat sealing face has been in- corporated in the design, and tempera- ture control of bolted joints and pump body above the casing is to be effected by circulating gas. Also, provisions for internal cooling of the shaft by circulating gas have been included. All bearings are external to the gas space above the coolant being pumped, and a circulated lubricant serves the bearings as well as the graphite seal. All parts except the impeller are to be of fabricated and machined con- struction. Impellers of Worthite have been received from Worthington Pump Corporation, . G. Cobb and J. F. Haines, “ARE Pump Design,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending Sep{;mber 10, 1951, CORNL-1154, p. 16 (Dec. 17, 1951). i i—l ANP PROJECT QUARTERLY PROGRESS REPORT Canned-Rotor Pump (M. Richardson, Reactor Experimental Engineering Division), The NaK test loop discussed in the previous report 1is being mod1fied to include a 30-kw heater and a vertical double-tube heat exchanger in which water flows as a falling film down the 8-ft length of the inside 2-in. tube and the sodium-potassium alloy rises in the annulus formed by a 3-in. pipe. The two canned ¥%-hp motors, wound with 500°F Class H windings, have been tested in water and found to operate satisfactorily. Attempts are being made to can the to protect them from high temperatures. Stainless steel has been used and found to operate satisfactorily, but the increase in gap caused by the 0.010-1n. wall of stainless steel causes the current load to become excessive, Some magnetic material will be tried next which will protect the rotor and reduce the stator-rotor gap. rotor elements COTTrosi1on at The 3- to 6-hp canned-rotor pump was operated for a short period but the performance was poor. It 1s believed that this poor performance was caused by the turbulence and thrust load which, in turn, were caused by the poor design of the suction inlet to the pump. This 1is being corrected by straightening out the sharp bend at the suction face and placing a baffle at the inlet to reduce the pressure on the shaft at this point, A sleeve bearing will be introduced within the motor can which will protect the can from damage caused by misalignment but will allow the rotor bearings draulically. to operate hy- A new wire 1nsulating material which is manufactured by the Bentley- Harris Insulating Company of Coshocton, Pa., would allow the canned-rotor pumps to operate at 1200°F. This 42 insulation will be 1included in future pumps. Investigations are underway to determine the possibility of making the rotors of suitable material, which wi1ll make canning unnecessary. Frozen-Sodium-Seal Pump (W. B. McDonald, ANP Division). Testing the frozen-sodium seal under dynamic conditions was effected in an 1sothermal test loop containing a modified Worthington centrifugal pump. This test ran approximately 450 hr at sodium temperatures up to 1100°F, alimitation imposed by silver-solder pump-casting repairs. Flows up to 40 gpm were attained, and the pressure differential across the seal was 25 psig. No indication of seal failure was observed during the test. Two-Stage Electromagnetic Pump (J. H. Wyld and A, L. Southern, ANP Division). Some parts of the two- stage electromagnetic pump have been received from the main shop; however, the increase in emphasis on fused fluorides has necessitated postponing plans for immediate testing. Electromagnetic Pump Cell Develop- ment (A. G. Grindell, Engineering and Maintenance Division). Two types of electromagnetic pump cells designed by ANP personnel for high-temperature liquid-metal applications were given preliminary tests during the period. The first of these featured integral construction of the throat and elec- trodes from type 316 stainless steel. Initial tests indicated that severe local overheating occurred. The second cell, employing nickel lugs inserted into the stainless steel pump throat and welded in place with inconel filler, operated satisfactoraly for approximately 160 hr, pumping sodium at temperatures up to 1300°F. Test-loop heater limitations prevented raising the sodium temperature above the 1300°F value. FOR PERIOD ENDING DECEMBER 10, 1951 SEAL TESTS W. B. McDonald, ANP Division Many problems arising in connection with pump development for liquaid metals and fluorides may be more properly classified as seal problems. A program is underway in which various types of seals are evaluated on pumps as well as in simulated eguipment. Seals currently being developed are the frozen-sodium seal, the frozen- fluoride seal, and the graphitar ring-—tool steel gas seal. The frozen- sodium seal has been successfully tested for 1150 hr with pressure across the seal of up to 107 psig. Only limited success has yet been ex- perienced with the two other seals. Frozen-Sodium Seal. During this quarter one phase of testing the frozen-sodium seal was successfully completed., Using the seal-testing device described in a previous re- port,¢?? the frozen-sodium seal operated for a total of 1150 hr at sodium bath temperatures up to 1500°F. Maximum pressure across the seal with no indication of failure was 107 psig. Just prior to termination, however, the pressure was raised to 150 psig, resulting in a small amount of sodium being forced out of the system around the shaft. Pressure was further in- creased to 200 psig in an effort to make the seal fail completely, but the leakage observed appeared to be no worse than at 150 psig. Leakage was at no time sufficient to cause a serious sodium fire; rather, it was characterized by sparks emerging from around the shaft. A design has been completed in which the refrigeration coil, used in the tests above to freeze the sodium, heas been replaced by copper cooling fias attached to the stainless steel sleeve. (3)%'. B. McDonald, “Frozen-Sodium Seal,” ORNL-1154, op. cit., p. 21, Calculations indicate that a section approximately 5 in. long and contain- ing 3%-in.-diameter fins spaced 1/8 in. apart are sufficient to freeze sodium around the shaft by convective cooling only. This radiant-cooled seal has not been tested. Frozen-Fluoride Seal. Design and assembly have been completed on a device with which to determine the feasibility of a frozen-fluoride seal. The principle involved is identical to that of the frozen-sodium seal in which the liquid is solidified in an annulus between a sleeve and a rotating shaft. (3) The first test has been completed in which the shaft, extending through a finned cooling sleeve into a pot containing eutectic NaF-KF-LiF-UF, at 1200°F, was driven by a 5-hp motor. Pressure across the seal was 20 psig. The test lasted for 19 hr hbefore a heater failure and a leak caused termination. Postrun inspection revealed that material was deposited on both the shaft and sleeve, and some scoring at the bottom end of the shaft was observed. Whether the attack on the shaft 1s chemical or mechanical was undetermined, but the material deposits were found to be magnetic. Graphitar Ring—Ketos Toel Steel Gas Seal. Testing of a graphitar ring—Ketos tool steel gas seal was continued during the period. Thas type of seal, consisting of a graphitar ring sandwiched between a stationary and a rotating member both made of Ketos tool steel full hardened, was tested in the presence of air and of helium to secure control data for comparison with runs made in the presence of NaK vapors. Successful sealing was obtained in these i1nstances with differsntial pressures of 5 psig. 43 ANP PROJECT QUARTERLY PROGRESS REPORT Three tests were conducted with the seal operating in the presence of vapor from NaK bath at 600°F, a seal temperature of 500°F, a spindle speed of 1300 rpm, and a differential pressure of 5 psig. The first test failed after 24 hr, and postrun ex- amination indicated that the graphitar had been attacked. Subsequent tests operated under similar conditions with no apparent attack on the graphitar ring for 70 and 170 hr, Failure in both cases appeared to be due to eccentricity in the seal test device. As the period ended, a fourth test had been in progress for 220 hr with no indication of failure. Ap- parently the graphitar in the first test was faulty, respectively. TEST LOOPS A. G. Graindell, Engineering Maintenance Division A calibration loop for checking electromagnetic pumps and flow measur- ing devices is now being used routinely for this purpose. The sodium manometer loop for measuring flow rates has not proved successful. €alibration Loop. The calibration loop was designed for testing electro- magnetic pumps and flow measuring devices. During the previous period this loop was used for reproducing manufacturer’s performance curves for a type G-3 General Electric electro- magnetic pump. This work was con- tinued i1nto the present period with the result that two electromagnetic pump cells designed by ANP personnel were tested. HResults of these tests are included in the section "Electro- magnetic Pump Cell Development™ (see above). Sodium Manometer Loop. During the previous quarter, the sodium manometer loop, designed for developing and 44 testing flowmeters, was used for test- ing an electromagnetic pump. First attempts to determine sodium flow in the system resulted 1n plugging of the small tubes connecting the flow nozzle to the primary manometer. Efforts to prevent this plugging by using NaK in place of sodium as the working fluid failed because NaK shorted electrical probes 1n the level control device. De-emphasis on liguid metals systems in favor of fused salts caused this project to be placed on a standby basis. SELF-RELDING TESTS The self{-welding of metal to metal contacts in a high-temperature fluid stream, as might occur in reactor pumps or valves, is being examined. Tests of five pairs of metals under static load in sodium for 100 hr have been completed. Stellite against stellite and inconel against zirconium exhibited no welding. Test of valves in sodium systems indicated a signifi- cant increase 1n opening torque with time. The valve had a stellite face and a 316 stainless steel seat. Materials Tests (G. M. Adamson, Metallurgy Division)., Several ma- terials were examined to determine their self-welding properties at high temperatures in the presence of liquid sodium. A series of three cylinders with diameters of 1/8, 1/4, and 3/8 in. were machined from one material to be tested, and three corresponding cylinders 1/2 in. in diameter were machined from either the same or different material. After being polished, pairs of these were placed end to end in the desired liquid bath under a load of 75 1b for 100 hr. The results, obtained in sodium at 1500°F, were: Few points of welding but broke during handling Inconel-inconel (120 grit) FOR PERIOD ENDING DECEMBER 10, 1951 Inconel-inconel Welding hed just started (polished) but pieces could be hand] ed Inconel - zirconium No welding Few points of welding on two smallest cylinders 347 stainless stee]l — 347 stainless steel Stellite-stellite No welding in 65 hr Valve Tests (W, C., Tunnell and K. W. Beber, ANP Division). Valve testing experiments were begun during the gquarter to determine the suitability of commercially obtained valves for use in liquid-metal systems. The first test was conducted on a Y%-in. Powell bellows valve with a stellite face and a 316 stainless steel seat, The valve was installed in the hot leg of a thermal-convection loop which was filled with sodium and heated to 1500°F. The valve was to be left open for 24 hr and then closed for 24 hr with a torque of 20 ft-1b. The torgue required for opening the valve varied widely during the 323 hr of the test, although the valve was operable at all times. Postrun inspection of the valve awaits availability of suitable inspection equipment. : A second test was started in which two %-in. Powell bellows valves were placed in parallel]l in the test loop in order that the valves could be alternately opened and closed on 24-hr cycles, thereby permitting convection fiow of the sodium at all times. The valves in this test also had stellite faces and 316 stainless steel seats and were closed with 35 ft-1b torque. Torgques required to open each ranged from 20 to 53 ft-1b. The time of closure was increased from 24 to 48 hr, and the torque required to open the first valve was 76 and 44 ft-1b after the first two cycles, while the second valve required 46 ft-1b after one cycle, Valve tests were still under- way at the end of the period. The study of bellows type valves, with the bellows surrounded by con- ventional packing material, has been suspended owing to the successful operation of the conventional valves and the emphasis on fluorides. HEAT-EXCHANGER TESTS A. P. Fraas and M. E. LaVerne ANP Division A heat exchanger with NaK as both the primary and secondary fluid has operated for 550 hr at temperatures up to 1200°F. The agreement between actual and expected heat transfer 1is good at high (11 gpm) flow rates but is poor at low (5 gpm) flow rates. A heat exchanger (radiater) with sodium as the primary fluid and air as the secondary fluid has been designed and built. NaK to NaK Heat Exchanger. The NaK heat exchanger figure-eight loop has completed 550 hr of operation at temperatures up to 1200°F and flows of 5 and 11 gpm. Increased tempera- ture could be achieved only by reducing the flow rate. Present plans call for continuing testing until 1000 hr of operation has been completed. Thus far there has been indicataon of neither any increase in pressure drop across the heat exchanger nor any loss in heat-transfer performance. This has been reassuring since the heat-exchanger matrix was made of 1/8- in.-o.d. tubes spaced 0.041 1n. apart, and it had been feared that clogging might be a preblem. All experimental features of this loop have proved quite satisfactory; for example, the nicro- brazed electromagnetic pump cell, by-pass filter, venturi, and null-type pressure units have given no trouble. Heat-transfer data at high flow rates agree well with values computed from Lyon’s formula, but data at low 45 ANP PROJECT QUARTERLY PROGRESS REPORT CLASS I FIED { N = = — 46 ir Rad iator. FOR PERIOD ENDING DECEMBER 10, 1951 flow rates give heat-transfer co- efficients as much as 40% below the theoretical values, Sodium to Air Radiator. A high- performance air-radliator core section suitable for use 1n a J-53 turbojet engine has been designed and con- structed (Fig. 6.1). The core matrix consists of 0.010-1n.-thick stainless steel fins spaced 15 to the inch on 3/16-in.-0.d. stainless steel tubes. The latter have been placed in a square spacing on 2/3-in. centers. Parts for the core-element test section have been completed and are ready for nicrobrazing, Preliminary testing 1is to be effected with air supplied by a blower extracting the heat which 1s supplied by electrically heated Nak pumped through the radiator. LIQUID-FUEL SYSTEMS E. Wischhusen and D. R. Ward ANP Division The fluid dynamics of the ligquid- fuel system of the liquid-metal-cooled reactor have been studied from mockups of the fuel system. The first two mockups(*’ were primarily glass systems with a colored fluid and yielded use- ful qualitative information on such problems as filling and emptying rates, valving requirements, bubble formation and migration, and flushing methods. A third mockup of the static-fuel system was intended, but the design change to a circulating-fuel system and the difficulty in transferring molten fluorides has directed attention to their handling. The fuel-transfer apparatus consists essentially of two tanks, one in which the raw material 1s introduced and a second to which the molten fluorides g, L. Meem, ‘““Bulk Shielding Reactor,” ORNL-1154, op. cit., p. 83. are transferred through a filter by gas pressure, Fuel lines are 347 stainless steel while all other com- ponents are of 316 stainless steel. The filter is of 5-u stainless steel. After the first transferral, a viewing window was provided in the charge tank. The fuel has been successfully filtered and transferred from one tank to the other., The unfiltered fluid is generally dark, the filtered fluid rather bright green and uniform in appearance. At the transfer tempera- ture, 1400°F (the melting point of the fuel, NaF-KF-LiF-UF_, is 960°F), the fuel has a watery consistency and a concave meniscus. Postrun examination of the equipment revealed a black layer of uranium oxides on the bottom of the supply tank and on the inlet side of the filter. Portions of the system wet by the fuel mixture exhibited a rather bright satinlike appearance, indicating the removal of welding scale and oxides. Subsequent transfer resulted in a leak in a connecting fitting and after that a plugged transfer line. Examination of the filter now shows the presence of bundles of dendritic metal crystals which are being analyzed. INSTRUMENTATION J. F. Bailey and P. W. Taylor ANP Division A. G. Grindell, Engineering and Maintenance Division Instruments capable of giving adequate indication and control of either liquid-metal or fluoride systems must be developed., Work is being carried out on the development of level-control, flow-measurement, and pressure-measurement devices for the ARE. Spark plugs appear satisfactory for level control sincemolten fluorides conduct electricity. Venturi methods 47 ANP PROJECT QUARTERLY PROGRESS REPORT appear to be best suited for the measurement of the flow of fluorides. The null-balance type pressure-measur- ing devices are suited for both liguid- metal and fluoride systems., Level Control and Indication. Work on instrumentation for level control and indication for liquid-metal systems centered around the testing of resist- ance type probes. An experimental loop for testing level control and indicating devices was fabricated in which a resistance type probe immersed in sodium was connected to a Brown potentiometer through a Wheatstone bridge. As the sodium level fluctuated, the resulting resistance change of the probe produced a signal which both indicated and controlled the level. This method proved unsatisfactory since the probe resistance constantly changed as a result of sodium buildup. [imited experience with fused fluorides indicates that sparkplug type level controls are adequate for laboratory-scale systems. Solid- state fluorides are nonconductors, which eliminates the problem of elec- trical shorting of probes previously encountered with sodium,. Flow Measurement. The shift in emphasis to fused fluorides caused a change in flowmeter development. Electromagnetic flowmeters had received major consideration for use with liguid metals, but the comparatively poor electrical conductivity of fused fluorides as compared to container materials made this means for detect- ing flow appear impractical. On the other hand, venturi methods of flow measurement appear to be quite practi- cable, having virtue in their sim- plicity of construction and small head loss under operation. Since this method involves differential pressure measure- ment, null-balance type pressure- measuring devices are being examined. 48 Pressure-Measuring Devices. Seek- ing suitable methods for measuring pressure i1n liquid-metal systems led to the modification of null-balance type devices to meet specific needs. The device in use, as modified, operates on the principle of balancing pressure on either side of a bellows and employing electrical methods for establishing approximate bellows The bellows 1s essentially filled with trapped gas which transmits circu- lating-fluid pressure. This pressure 1s, in turn, balanced by gas pressure applied external to the bellows. The vent valve enables the pressure of the system to be "bracketed" between two expansion or position. volume values within 0.1 psi, the pressure required to expand the bellows suf- ficiently to make electrical contact which 1s indicated by a signal light and/or a buzzer. Although these pressure-measuring devices were designed and fabricated primarily for use with liquid-metal systems, the only foreseeable objection to their use with systems containing fused fluorides 1is the methods used for joining the bellows to bellows heads. The search for suitable methods for making these connections centers around heliarc welding, resistance seam welding, and nicrobrazing. Thus far, two devices have been assembled using nicrobrazing for joining bellows and heads and have been installed on the gas-seal fluoride pump. FULL-SCALE ARE COMPONENT TEST FACILITIES H. P. Kackenmester and G. A, Cristy ANP Division Plans for full-scale testing facilities for ARE components to be located in Bldg. 9201-3 were reviewed in relation to providing them by late winter or early spring, 1952, For FOR PERIOD ENDING DECEMBER 16, 1951 the present, utilities and services are being provided sufficient to handle fluorides and other equipment capable of being safely operated without test cell enclosures. Plans and specifications have been developed for construction of an equipment train in which fluid-flow measuring instruments can be calibrated and pumping characteristics can be determined by weight-rate measurement of fluid flow. Equipment for the calibration system will be capable of handling either liquid metals or molten salts. Jnquiries to manufacturers indicate that low-frequency (60-cycle) induction heating apparently offers a versatility along with higher heat- input rates to vessels contalning either liquid metals or molten salts than can be achieved by radiant- or conductive-heating equipment. Also, investigations of fume-collecting equipment show that several types of relatively inexpensive commercial equipment are available which will collect 99%% of condensed dispersoids of liquid wetals or liquid salts 1in large guantities (20,000 to 50,000 cfm) of air in temperature ranges of 0 to 2000°F. FLUORIDE PRODUCTION L. A. Mann, ANP Division The decision to make engineering- scale tests on eutectic salts of LiF-NaF-KF activated earlier plans for assembling and fabricating production type equipment and storage contalners for this fluoride mixture. Liaison with the Liguid Fuels Chemistry Group was maintained in the effort to ensure adequate quality and purity of product. Early runs were made by charging c.p. salts 1nto a melt tank 1n proper proportions for producing the eutectic mixture, purging air from the system, and melting at 1825°F under 5 psi of helium. Material was forced from the melt tank into the receiver through a 5-4 sintered 316 stainless steel filter by helium gas pressure. After three melts, succeeding runs were made uvnder approximately a 500-u Hgvacuum, Also, 316 stainless steel scrap was put into the chamber to "precondition" the mixture., Thus far, seven melts have been made, totaling approximately 400 1b of eutectic fluorides. The addition of desired amounts of UF, to certain batches has been accomplished in the receiving tank. The entire contents are agitated by bubbling helium through the mixture. CLEANING OF FLUORIDES VFROM SYSTEMS R. Devenish, ANP Division With the advent of fused fluorides, the problem of removing the residue from containers in asafe and efficient manner arose. Sodium and potassium fluoride constituents are comparatively water soluble and hence present less difficulty; however, lithium fluoride, being only slightly soluble in water, adhered to the container walls. The best method found thus far for removing this material i1is with hot 25 vol % nitric acid solution. Aluminum nitrate for this application proved unsatisfactory. In collaboration with the Health Physics group the following pre- cautions have been temporarily es- tablished for working with fluorides: 1. Dust respirators should be worn by personnel when charging the fluorides of lithium, sodium, and potassium i1nto production furnace pots. 2. Assault masks withM-11 canisters should be used when working with 49 ANP PROJECT QUARYTERLY PROGRESS REPORT molten fluorides in open con- talners. NakK DISPOSAL R. Devenish, ANP Division A method was formulated for dispos- ing of used NaK which previously had presented difficult storage problems. In this procedure NaK was mixed with graphite powder and kerosene, and the resulting slurry was oxidized off with steam. A continuous fire was observed until all the NaK was consumed. 50 ALKALY METALS MANUAL R. Devenish, ANP Division The Alkali Metals Manual setting forth safety considerations to be observed when working with alkali metals was distributed during the period. Also, a Y-12 plant procedure covering handling of liguid metals was prepared by joint effort of the Y-12 Alakli and Liguid Metals Safety Committee and the Plant Procedures Group. This procedure was submitted to the Y-12 plant superintendent for approval, SUMMARY AND INTRODUCTION E. P, Blizard, Physics Division The aircraft unit shield design as specified over one year ago by the Lid Tank optimization experiments has heen verified in the Bulk Shielding Facility (Sec. 7). The calculations are now complete and show a slightly lighter shield (127,700 1b actual weight) than was originally predicted. The experi- ments in this shielding reactor have been started on the divided shield. A long period of detailed measurement will ensue, however, before a divided shield weight will be available. The most significant progress in the shielding art has been the air- duct theory, which grew directlyout of a series of tests in the Lid Tank (Sec. 8). Measurements of the neutron at tenuation in air ducts agree to within 20% of the calculated dose. For these duct sizes (2 in. 1.d.) with bends, no correction 1s necessary for streaming and the net effect of the ducts on the shield weight is of the order of 1 to 2 tons. This favorable result with air ducts implies that liguid-metal ducts would have a negligible effect on total shield weight., The divided shield mockup, mentioned above, cannot provide shield weights with the degree of accuracy possible with the unit shield because of the uncertainty associated with the trans- mission of gamma rays in air. Conse- quently, a proposal has been submitted to the AEC for a "Tower Shielding Facility” with which air scattering, as well as the effect of this scat- tering on the crew shield, may be determined (Sec. 9). Shielding weights and thicknesses have been estimated for the circulating-fuel reactor and do not appear to be significantly different from those for the sodium- cnoled reactor in spite of the circu- lating fissionable material. 53 ANP PROJECT QUARTERLY PROGRESS REPORT . BULK SHIELDING REACTOR 7 L. Meem G. Cochran . P. Haydon M. Henry B. Holland H. E. Hungerford . Johnson . Leslie . Maienschein G. M. McCammon E. J. F. OR® T. N. Roseberry Physics Division During this quarter themain efforts at the Bulk Shielding Facility have been devoted to analyzing and reporting the results of the measurements on the unit shield mockup and inpreparing for and installing the divided shield mockup. REACTOR OPERATION Preliminary measurements on the temperature rise of the water flowing through one of the reactor fuel elements have been made, and the results look reasonable. However, to obtain an accurate value for the heat released per fission, an especially insulated fuel element is being ordered and the measurements will be repeated. Since the presence of the borated water in the unit shield mockup altered the power distribution i1n the reactor, it was necessary to redetermine the power with neutron-flux measure- ments. {!? The total power of the reactor against the borated shield was only 3% less than the value found previously with the reactor im water without the shield. However , the power generated in the front row of fuel elements was 16% less with the reactor against the shield than with no shield. These new power measurements were use to calculate(?’ the leakage from (1)g. B. Johnson, Power Calculations of the l{géi)Shidd Reactor, ORNL CF-51-9-112 (Sept. 18, (2)]. L. Meem and H. E. Hungerford, Calculations E; geaka ¢ from the Bulk Shielding Reactor, OBNL 1-10-94 (Oct. 5, 1951). 54 the front face of the reactor into the shield. The leakage L from a reactor is glven by z L =F P(Z) e %/N 4z where F is a factor converting from the power produced to the appropri- ate type of radiation dosage escaping, Z is the distance into the reactor from the surface, P(Z) is the power per unit volume as a function of Z, Z, an@ Z, are'appropriate limits of integration, and A is the relaxation length of the escaping radiation. P(Z) was determined from the above- mentioned experiment and a value of A= 9,2 cm. The factor F need not be evaluated as will be seen later. The leakage from the Bulk Shielding Reactor was found to be L =F x 0,735 watts/cm? with reactor operating at a power of 10 kw. Taking into account the attenu- ati1on of a small amount of water and aluminum between the reactor face and FOR PERIOD ENDING DECEMBER 10, the inside surface of the shield, the leakage from the reactor into the unit shield was L, = F x 0,607 watts/cn’ at 10 kw power. MOCKUP OF THE UNIT SHIELD The results of the unit shield experiments are presented in detail in ORNL-1147.¢?) The experimental curves for gamma rays, thermal neutrons, and fast neutrons are shown in Figs. 7.1, 7.2, and 7.3. In one experiment (Exp. 3) measure- ments were made starting from the surface of the last lead layer and proceeding outward in the water of the pool along the centerline. In another experiment (Exp. 4) a tank of borated water was placed around the shield as described in an earlier pProgress report,(q) and measurements were made ocutward from the wall of the tank. As the boron concentration was increased, gamma measurements were repeated, as shown in Fig. 7.1, until a concentration of 0.4% was reached. The weight of the unit shield as specified by the Shielding Board(®’ was 134,000 1b for a 200-Mw reactor in the shape of a 3-ft right square cylinder with a dose of 1 r/hr al- lowed at the crew, 50 ft away. The power density of the aircraft reactor was assumed to be constant, (3}J. L. Meem and H. E. Hungerford, Unit Shield Exfieriments at the Bulk Shielding Facility, ORNL-1147 (in preparation) (4)A. S. Kitzes, “Mock-up of Unit Shield,” Aircraft Nuclear Propuls;on Prafifct Quarterly Progress Report for Period Endlng arch 10, 1951, ANPZ50, p. 158 (June 19, 1951 (siafl?art of the ANP Shteld:ng Board, NEPA- OFNL, 53 (Oct. 16, 1950). 1951 and accordingly the leakage is calcu- lated to be L = a F x 3,07 x 10* watts/cm? with reactor at 200 Mw. It 1s convenient for the purpose of these calculations to define a dosage unit D as follows: The D unit 1s the maximum dosage rate of nuclear radia- tions that may be taken by military persennel during a 25-hr flight of a nuclear-powered aircraft. Accordingly, 1D = 1 r/hr of gamma radiation or 0.1 rep/hr of fast neutrons The tolerance dosage of 1D is not to be exceeded in any combination of fast neutrons and gamma rays. If 1D is the tolerance dosage at the crew compartment, then the allowed dosage, D_, at the surface of the shield measured at the Bulk Shielding Reactor 1s where S is the crew separation = 50 ft, R 1s the outside radius of shield required, L and L are the leakages defined above, . is the radius of the unit shield reactor = 33 cm, and rajfi the radius of the aircraft reactor = 58 cm. 55 93 GAMMA RADIATION INTENSITY, R/HR/WATT 1073 0™ o7 BORATED WATER WITHIN TANK : ALUMINUM TANK EXP 4 ONLY ( ) (EXP. 4 ONLY) EAR OF UNIT SHIELD MOCK-UP CURVE 4 ; GURVES 2,3 WITHOUT ALUMINUM '3 TANK 4,AND 5 EXP.3 EXP 4 50 cc STD 10'2 GHAMBER o . 900 cc 10'0 CHAMBER A 900 ¢c 10'® CHAMBER 0 G-M GOUNTER 10 130 150 170 190 2i0 DISTANGE FROM REAGCTOR, cm SECRET DWG. {2817 y —INTENSITIES BEHIND Pb—H,0 UNIT SHIELD MOCK-~UP (EXP. 3) y —INTENSITIES BEHIND Pb-B- H,0 UNIT SHIELD MOCK-UP WITH VARIOUS CONCENTRATIONS COF BORON AS INDICATED (EXP. 4) 230 250 270 Fig. 7.1. Gamma Radiation Intensity for the Unit-Shield Experiments. 290 LHOdAY SSAUN0Ud A THILYVAD LDAfoUd dNV 10 103 Toka 0—4 THERMAL NEUTRON FLUX, NV, /WATT Fig. 7.2. 80 FOR PERIOD ENDING DECEMBER 10, 1951 SECRET DWG. 12818 REAR OF ALUMINUM TANK REAR PERIPHERY OF EXP 4 ONLY) UNIT SHIELD MOCK-UP BORATED WATER WITHIN TANK “(EXP 4 ONLY) - ~ THERMAL NEUTRON FLUX-—1"~ ““BEHIND Pb- Ho0 UNIT SHIELD THERMAL NEUTRON FLUX - "BEHINDG Pb-0.4% B-Hp,0 - UNIT SHIELD MOCK-UP~~ (EXE 4) 2N, FISSION COUNTER INDIUM FOIL. MEASUREMENTS 8 IN. BF; COUNTER 2 IN. BF, SINGLE CHAMBER COUNTER 2 IN. BF, DOUBLE CHAMBER COUNTER 100 120 140 160 180 200 220 240 260 DISTANCE FROM REACTOR, cm Thermal -Neutron Flux for the Unit-Shield Experiments. 57 ANP PROJECT QUARTERLY PROGRESS REPORT 58 FAST NEUTRON DOSAGE, MILLIREP/HR/WATT 10 S, 5] W g, A 10 SECRET DWG. 12816 | L L] __REAR PERIPHERY OF UNIT SHIELD MOGK-UP /REAR OF ALUMINUM TANK (EXP. 4 ONLY) ] = ) ! T —~ BORATED WATE WITHIN TANK TT(EXP. 4 ONLY) o~ DATA TAKEN BEHIND Pb-H,0 B e e — UNIT SHIELD MOCK~UP (EXP 3) T N DATA OBTAINED BEHIND Pb-04%B- HoO UNIT SHIELD MOCK-UP (EXP. 4) \O\ e | NPT FAST NEUTRON oo - N#“ DOSAGE BEHIND ~F— S - N UNIT SHIELD (EXP 3) ~ffo oo - ] p\%\ & """"""""""" N |5y """"""""" ' *\\* T = ] \\ : Tl . ] """""""" N B _______ e, - O\_O_ TN . o I . e - \O u_;fikf ''''''''' i o Frrie ] \E\_ I T FAST NEUTRON DOSAGE .. 1 e +— —UNIT SHIELD (EXP. 4) | T - BEHIND 0.4 9 BORATEVV";A\_W.- 80 Fig. 7. 3. 100 120 140 160 DISTANCE FROM REAGTOR, cm 180 Fast-Neutron Dosage for the Unit-Shield Experiments. FOR PERIOD ENDING DECEMBER 10, 1951 For these calculations the 3-ft square cylindrical aircraft reactor isclosely approximated by a 3.8-ft sphere. The significance of each term in this expression 1s: (S/R)? is the attenuation due to reactor- crew separation, inverse-square-law Lu/La is the ratio of leakages of the two reactors, and ru/ra is a factor which represents the geometrical attenuation in a shield with spherical surfaces. Using this formula and the data of Fags, 7.1, 7.2, and 7.3, the shield thickness was determined and the weight of the aircraft shield was found to be 129,200 1b. Since the insertion of a layer of lead in water has anegligible effect on the neutron attenuation,; and since the effective relaxation length for gamma rays in lead at this position in the shield has been established as 3 cmat the Lid Tank, 1t 1s interesting to calculate the effect of adding or substracting lead in the last layers of the shield. This 1s done by holding the total neutvon plus gamma dosage constant but varying the fraction of the dosage taken in either neutrons or gamma rays. When lead is subtracted, the apportioned gamma dosage goes up and a thicker outside layer ol water muist be added to bring the apportioned neutron dosage down, and vice versa. The results of these calculations are shown inFig. 7.4, where the shield weight 1s plotted against both the shield thickness and the outside radius of the shield around the Bulk Shielding Reactor. The percentage of the dosage taken in neutrons is also shown. In Shield No. 1, the shield actually measured as weighiang 129,200 1 (mentioned above), 44% of the dose is taken 1n neutrons and the rest in which was gammas. For ease of construction of the mockup all lead layers were made with l-in. thicknesses. Actually, the shield specified by the Shielding Board was only 0.4 in. thick in the last lead layer. If this excess lead is peeled off and water added, the weight goes down to 127,600 1b, as shown under Shield No. 2 in Fig. 7.4. This is the minimum weight of shield which is possible with the present lead spacing and boron concentration. This shield is identical with that predicted by the Shielding Beoard except it has 6 cm less water. The agreement in weight 1s within 5%. All the combined errors in measurement and calculation from the bulk shielding data do not add up to an error of 5% in shield weight. It 1s estimated that these combined errors would amount to less than a ton. The important result of these exper iments 1s not that an aircraft shield will weigh 127,600 1b, since this is an ideal shield. The necessary engineering will undoubtedly increase 1ts weight, while current Lid Tank experiments iadicate that the addition of more boron to the water with a subsequent reoptimization of the lead spacing may decrease the weight by as much as 10%. The real significance of these tests 1s that the methods of calculation used by the Shielding Board have been confirmed. Given the proper specifications about the reactor and airplane, it is probable that the weight of an engineered unit shield can be calculated from existing data to within a few tons. MOCKUP OF THE DIVIDED SHIELD The divided-shield mockup has been installed in the pnol and measurements are now underway. A photograph of the installation with the reactor in place 2% ANP PROJECT QUARTERLY PROGRESS REPORT DWG 12962 THICKNESS OF UNIT SHIELD, cm SECRET 120 124 128 132 136 140 144 148 152 156 160 164 168 172 ‘ MINIMUM THEORETICAL SHIELD THICKNESS f.—{NEUTRON DOSE =100% TOTAL DOSE) { } REMOVE Pb FRGM NEXT l ADD Pb TO | INNER Pb LAYER tOUTSIDE Pb LAYER . - | (LAYER 17) - 100 N 57— | REMOVE Pb FROM OUTSIDE f’ W ]\ i Pb LAYER (LAYER 19) SHIELD WEIGHT —el /. 90 Q I VARIATION WITH DISTANGE 3 FOR 1 R/hr TOTAL DOSE = 2\ /N S 148 i l \, {USE SCALE TO LEFTj//// 80 o AN 70 & l \ N //7 o 144 1 p 4 60 4 \ N | NEUTRON DOSE L/ g | N 44% OF TOTAL DOSE A 50 W o \ 5| FoR SHIELD NO. 1 7 & LJ % l _____ . :‘\ . J l A1 /1 e 40 < v ~ w NEUTRON DOSE 0 z ~| 25% OF TOTAL DOSE o ;7// — 30 g 1 \\ ‘\(%Ef/’ FOR SHIELD NO. 2 V4 S o §es YMFE\ L 20 3 = : : THE PREDIGTED SHIELD — = 0 = z l (SAME AMOUNT OF Pb ‘; e L b L] W 2132 -eh——1 AS SHIELD NO. 2) ] 4 _ Tl g = z | J WT.= '34’300 ib PERCENT OF TOTAL NEUTRON DOSE 3 1 + B t 1= VARIATION WITH DISTANGE : = SHIELD NO. | .y | [ {USE SCALE TO RIGHT) $ 128 - (ME ASURED) etk ] }-- — { R/hr TOLERAMNCE /’/{- | wr=t29200 | t—7T 1 1 | } 4+ | 1 | 4+ i ot b ] ] <"SHIELD NO.2 (OPTIMIZED) | { R/hr TOLERANGE 124 T WT. =127,700 1b [t | 120 l 178 182 186 190 194 198 202 206 210 214 218 222 226 230 OUTER RADIUS OF SHIELD, cm Fig. 7.4. Dependence of the Weight o is shown in Fig. 7..5. A report on the gamma-ray spectrometer being used for the gamma-ray measurements 1s being prepared.(fi) In the last quarterly report{’? an exper iment was described in which the gamma radliations from the reactor were used as a source to observe the effec- tiveness of a lead slab as a shadow shield. The objective was to obtain information which would be of help in (G)F. C. Maienschein, Multiple Crystal Gamma- Ray Spectrometer, ORNL-1142 (in preparation). (7)epy1k Shielding Facility,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1951, ORNL-1154, Fig. 5.2, p. 86 (Dec. 17, 1951). 60 f the Unit Shield on its Thickness. designing the lead shadow disk within the hydrogeneous reactor shield. This problem is common to many divided shield designs. The reactor plus the capture gamma rays 1t produces 1in the water proved to be too diffuse a source for definitive conclusions. Conse- quently, the experiment has been repeated(8> during this quarter using a can of radioactive sodium which 1s small enough to simulate a point source. The layout of the experiment and the shadow patterns observed behind the lead slabs are shown in Fig. T7.6. (S)H. E. Hungerford, Experiment V-A at the Bulk Shielding Facility — Shadow Shield Measure- ments with aNae>? Source, ORNL CF-51-11-95 (Nov. 15, 1951). ? FOR PERIOD ENDING DECEMBER 10, 1951 Fig. 7.95. Installation of Divided-Shield Mockup with Reactor in Position. 61 ANP PROJECT QUARTERLY PROGRESS REPORT SECRET DWE. 1298 T T 1717 1 1 1 1 1 117 17T 1771117 11 1T 71T 1073 @ @ 60 & 10—4._- @ I 1 3 | | 10-2 — & & l 8 N 5 | @ 25 @ 20 p 15 > HORIZONTAL OISTANCE FROM REAR OF SHADOW SHIELD, cm w o 3 o 5 J/‘ /4/ |0-3 — 7 ® 0 — Pb SLAB Iwz in. J 2in, / A PICTOGRAPHIC VIEW OF GAMMA RADIATION INTEN— SITIES EXISTING ALONG VERTICAL TRAVERSES H Pb SLAB xl,/z n. MEASURED BEHIND THE SHADOW SHIELD GURVE 1 VERTICAL TRAVERSE 5cm FROM SHIELD CURVE 2 VERTICAL TRAVERSE 15cm FROM SHIELD CURVE 3 VERTICAL TRAVERSE 35cm FROM SHIELD CURVE 4 VERTIGAL TRAVERSE 60cm FROM SHIELD CURVE 4 CURVE 4 TO SAME SCALE BASE AS GURVE 3 SGALES TO RIGHT OF EACH GURVE GIVE RADIATION N INTENSITIES IN r7hr L -I—’UP LINE A LINE CONNEGTING POINTS OF MINIMUM A 5 INTENSITY P c LINE B LINE CONNEGTING POINTS OF MAXIMUM z F——=< 60 cm INTENSITY a3 yay 4,1 LINE C SHADOW LINE OF SHIELD FROM GENTER Qs /. OF Na SOURCE 15¢cm <[ E &9 2 cl o 5cm e e m v SCALE DIAGRAM OF SHADCW 23 SHIELD EXPERIMENT WITH 2 SODIUM SOURGE B SHOWING WHERE TRAVERSES WERE MADE ANG THE RELATION OF LINES A,B, AND G TO SOURCE SOURGE @ /et e bl O 20 40 60 8¢ 100 120 140 160 180 200 220 240 VERTICAL DISTANCE FROM GENTERLINE, ¢m +* Fig. 7.6. Shadow-Shield Experiment with Sodium Source. 62 FAST FLUX {A/sec/cm@/watt ) 10 5 LTI boo—e FOR PERIOD ENDING DECEMBER 10, 1951 SECRET OWG. NO. 13194 T T T T ....... B UP® FAST FISSION GHAMBER, NG BORON OR e - CADMIUM LINER, NO Pt SHIELD " ,,,,,,, S~ — = e U FAST FISSION CHAMBER WITH BORON-CADMIUM T = S LINER, 3 INGHES OF Pb 10 cm IN FRONT OF GOUNTER N T @ DOSIMETER — | A SULFUR THRESHOLD DETECTOR — | ] e b 20 40 60 80 100 120 140 160 180 200 DISTANGE FROM REACTOR {cm) Fig. 7.7. Bulk Shielding Facility Fast-Neutron Data. 63 ANP PROJECT QUARTERLY PROGRESS REPORT . 9 Measurements are now available! ) on the fast-neutron flux from the Bulk Shielding Reactor using the dosimeter, a U?3% fission chamber, and sulfur threshold detectors. The (®JH. E. Hungerford and R. G. Cochran, Fast Neutron Mecsurements at the Bulk Shielding Facil- ity, ORNL CF.51-11-96 (Dec. 10, 1951). 64 results of the three types of measure- ment are shown in Fig. 7.7. "Fast- neutron flux" is in itself a nebulous quantity and 1s defined differently for each of the three detectors used. The agreement obtained was nevertheless, quite gratifying and gives added confidence to all previous results using the fast-neutron dosimeter. FOR PERIOD ENDING DECEMBER 19, 1951 8. DUCT TESTS C. E. Clifford, Physics Division Following the decision of General Flectric to build an air-cooled reactor, 1t was decided that the Lid Tank could do some useful experiments to aid in the design of the air ducts which penetrate the water-reactor shield. As these experiments pro- ceeded, the results indicated that the problem of penetrating a water shield with ducts, even though only filled with air, was not so difficult as had been previously supposed. Once a theoretical evaluation of the experi- ments has been made, the prediction of the transmission through sodium-filled only minor ducts would i1ntroduce complications. AIR-FILLED DUCT TEST IN LID TANK A. Simon J. D. Flynn T. V. Blosser Physics Division Two types of air ducting are under consideration: (1) small round pipes with various bends, and (2) annular ducts with various bends. In the annular ducts the flow is confined to the region between two large cylinders. Bends are effected by displacing the center of the ducts i1in fractions of 1ts diameter with both ends remaining approximately fixed. Cylindrical Ducts. Considerable work was done on the round pipes since they were believed to be the most favorable arrangement, The size of the ducts investigated was chosen on the basisof pressure-drop calculations made by the G.E. group. This indi- cated that the pressure drop due to an array of approximately 200 small circular ducts on the order of 2 to 3 in. in diameter would not be excessive. Design studies by the G.E. ANP group have indicated that the ducting system 1s not radically changed i1n contem- plated supersonic applications, The measurements were concentrated on the 1nlet air ducts, since these ducts would face the crew compartment and therefore would probably have the greatest effect on the shield weight due to the larger attenuation required in that direction. After some experi- mentation it became apparent that an array of 2-1n. ducts on a 3.7-1n. triangular lattice, four of which are shown in Fig. 8.1, could be used with- put seriously increasing the radiation transmitted by the shield. This array was built up of 2-1in.-1.d. electrical conduit, the walls of which were 1/16-in. steel. Standard 90° elbows were used which were bent onm a 10-1in. radius. Each elbow had two 5-1n. straight sections, so that when two were joined by means of a standard coupling, a 10-in. straight section resulted. The assembled elbows were held in place by wooden spacers and the whole array was mounted in the Lid Tank on a plywood table. Neutron Transmission. The effect of this duct array on the transmitted radiation is shown in Fig. 8.2. This shows the thermal-neutron-flux distri- bution in the water beyond an array of 15 ducts, three rows high and five rows wide. It was possible to measure only thermal neutrons in these tests because of time limitations and the necessity of taking a large number of data. However, 1t is felt that these thermal measurements can be readily interpreted in terms of fast-neutron dose by correlations obtained from measurements made 1in water of both thermal -neutron and fast-neutron dose under the same conditions. All that is required is that at some point 1in the thermal neutron measurements the 65 ANP PROJECT QUARTERLY PROGRESS REPORT 8 SOURCE PLATE ‘ SECRET | B FHOTO 11054 _284n.mA. VR e chi 3 2 2in-1D STEEL CONDUIT | (O.4-in. WALLS) g Fig. 8.1. An Array of Four 2-in. in excess of 6 cm after correction for geometrical attenuation. Measurements were taken both with two elbows joined and with three elbows joined. The two-elbow array penetrated approximately 40 in. of the shield and the three-elbow array penetrated 60 in. of the shield. slope be An integration of the vertical traverse for the two-bend case should 66 3 90° BENDS Steel Conduit Ducts with Three 90° Bends, give the transmission of an infinite array. The measured dose agrees within 20% with the dose calculated by taking account of only the reduced density due to the voids. The infinite array 1s the only geometry in which the true reduced density effect can be calculated. For this duct size 1n these bends no correction 1s necessary for streaming, a very encouraging result. As a consequence the effect FOR PERIOD ENDING DECEMBER 10, 1951 SECRET DWG. 13013 R1 21, ,=9.224 x 104 FOR 7=113.8 cm O-o0 S o O o O o / x TRAVERSES MADE AT z=#{3.8cm o 2 90° ELBOWS ® 3 90° ELBOWS z CENTER LINE MEASUREMENTS AT y=0 ® (¢ IN WATER © ¢ IN WATER BEHIND DUCTS WITH 2 90° ELBOWS / RELATIVE THERMAL NEUTRON FLUX, counts/min (121/2'in. BF; COUNTER) o Mo ¢ OF DUCT ARRAY WITH \ RESPECT TO «x 50 60 80 100 120 140 160 180 Z OR x (cm) 10 Fig. 8.2. Lid Tank Duct Test D-915, X Traverse (Vertical) and Z Centerline Measurements of Neutron Flux for Two Arrays of 2-in. Steel Conduit. 67 ANP PROJECT QUARTERLY PROGRESS REPORT RELATIVE THERMAL NEUTRON FLUX, counts/min (42‘/2-in. BF3 TWIN COUNTER, z=153.8 cm) -100 -80 -60 —-40 -20 SECRET DWG. {3016R1 STRAIGHT TWO 30° TWO 45° TWO 60° TWO 90° DUCT CROSS SECTION 20 40 60 BO {00 120 yi{cm) Fig. 8.3. Conduit (2 in. Lid Tank Duct Test D-10, i.d.) with Two Bends of of the ducts on the shield weight is very small — of the order of 1 to 2 tons — since the only additional shielding required is a conical patch along the periphery of the ducting system. Transmission with Variable Bends. From the point of view of pressure drop it was desirable to determine the minimum amount of bending required in the duct to limit transmission to the 68 Y Traverses in H20 behind 54-in. Rubber VYariable Radius. simple reduced density effect. To this, experiments were in the Lid Tank with tubing of 2 i.d. o.d. determine undertaken flexible 2-3/8 1in. in. and Figure 8.3 gives the results obtained by introducing two bends of various degrees in a 54-in. section of rubber conduit. The data indicate that even two bends of only 30° are more than sufficient to give a geometrical FOR PERIOD ENDING DECEMBER 10, 1951 133.8¢cm) RELATIVE THERMAL NEUTRON FLUX, counis/min (1244 in. BF; TWIN COUNTER, ~ -80 -60 - 40 -20 Fig. 52-1in. 8§.4. Lid Tank PDuct Test D-11, Rubber Conduit (2-3/8 in, i.d.) attenuation down the duct that 1s larger than the attenuation of the surrounding shielding material (water). Since large duct sizes and fewer bends reduce the pressure drop, the next experiment was done ona 2-3/8-1in. rubber conduit 52 1in. long with only one bend. Results of this experiment are presented 1n Fig. 8.4. The experi- ment was conducted by measuring first the transmission of the straight duct in the water beyond the end of the SEGRET DWG. 13015 Rt STRAIGHT DUCT 136 in. BEND 23/g in. BEND 33, in. BEND 7 Y% in. BEND 10" in. BEND 14, in. BEND o 20 40 60 80 100 Y Traverses {(Horizeontal) in #,0 behind with Variable Bends. duct, and then by measuring the trans- mission as the center of the duct was displaced by multiples of 1ts diameter from the centerline of the straight duct. As the center of the duct was displaced, the transmitted intensity was reduced. The first measurement, in which the center was displaced by one-half the diameter, gave a reduction of approximately a factor of 2, cor- responding closely to the reduction 1in the area of the source which could be seen from the center of the other end 69 ANP PROJECT QUARTERLY PROGRESS REPORT SOURCE PLATE 7’6 cm {6 cm —] SECRET DWG. 13021 Ry AIRPL ANE . V REACTOR ///A gom —=| - ///////// i HEADER ,_,/»—x\ CROSS SECTION OF 61em CYLINDRICAL SHELl——= f 105.6¢cm / 6icm 0.5 ¢cm - ' 457 glem x) ‘]L;~-4?.8 cm—s T~ | 12 5 crn CURVATURE NOT IN MOCK-UP AS TESTED 2’yx160m Fig. 8.5. Lid Tank Duct-Test Annular-Duct Rectangular Cross- Section, 1/8-in. Steel Plate Welded. 70 FOR PERIOD ENDING DECEMBER 1¢, 1951 SEGRET DWG, 13012R1 Zz TRAVERSES y =0 ¥y =623 ¥y =90 ¥ =100 y TRAVERSES zZ =22.4 zZ=1470.4 FAST NEUTRON DOSAGE BY DOSIMETER MEASUREMENTS - z=1{75, y=+16 8.666 x 1072 mrep/hr SOURCE PLATE (NOT PLOTTED) RELATIVE THERMAL NEUTRON FLUX, counts/min {8-in. BFz; COUNTER]) -0 O 20 40 60 80 100 {20 CENTER LINE DISTANCE (cm) Fig. 8.6. Lid Tank Test P-12, Y (Horizontal) and Z Traverses Behind Aonular PDuct, 71 ANP PROJECT QUARTERLY PROGRESS REPORT of the duct. When the center was dis- placed onediameter, entirely eliminat- ing a line of sight transmission down the duct, the largest decrease was noted, approximately a factor of 40. Further displacement of the center gave smaller decreases until, finally, beyond a 30° bend no further decrease was noted. Calculations for this bend indicated that the intensity was about what would be expected for the reduced density shield. Theoretical Duct Attenuation. On the basis of these dataa simple theory was derived which would predict the geometrical attenuation of the duct to within 20 to 40% for the various configurations tested. The theory is based on the picture of virtual sources created at the bends of the duct with the neutrons emerging with an isotropic distribution from the walls. Since these sources are small, the geometri- cal attenuation is Jarge for duct lengths that are many times the diameter of the duct (in comparison to the exponential attenuation of the sur- rounding water). A fuller account of this theory will be reported after further experimental correlation. Annular Ducts. One quick experi- ment was done on a mockup fabricated by G.E. to simulate a segment of an annular duct (Fig. 8.5). The measure- ments taken on this mockup are pre- sented in Fig. 8.6. These data have not been subject to analysis as yet, 12 but seem to indicate that theattenu- ation of high-energy neutrons {(which contribute more effectively to the dose) is greater than that of the lower energy neutrons. An analysis by G.E. of these data indicates that such a duct could be used, although it results in a sizable increase in the shield weight, The principal reason for using this type of duct is to in- crease the accessibilityof the reactor faces for control mechanisms. Liquid-Metal Duct Test in Thermal Column (F. J. Muckenthaler and M. K. Hullings, Physics Division). With the thermal column facility (previously referred to as the "Duct Test" or " "Duct Test Facility") the measurements carried out over the last year, on the 6- and 8-in. liguid-metal coolant ducts supplied by KAPL for all the interesting combinations, have been completed. The analysis of these data is in process and a complete report will be written. The experimental program in this facility is now ailmed at obtaining further experimental cor- relation with the simplified theory of air ducts. Neutron transmission will be measured for a set of small, flexible, air-filled ducts 1n water, in which the length, the diameter, the degree of bending, and the source size will be varied. Preliminary measure- ments have been completed on a 3-1n. duct 60 in. long for five bends from 0 to 90°. FOR PERIOD ENDING DECEMBER 10, 9. SHIELDING E. P. Blizard, Present shielding facilities do not permit a divided-shield mockup which incorporates the effects of crew-compartment shielding and air scattering on the radiations emerging from the reactor shield. Recause of the importance of this effect on air- craft shield weights, a proposal has been submitted to the AEC for a new shielding facility, called the "Tower Shielding Facility," with which air scattering and a crew shield mockup can be effected. This facility could be constructed at an estimated cost of about $694,650. Shielding weights and thicknesses for circulating fuel reactors have been estimated for application to configurations as they are developed, In general, these shields are pertur- bations of the divided shield for the sodium-cooled reactor. Divided-shield design studies are continuing at NDA in their effort for optimization of this shield weight. TOWER SHIELDING FACILITY PROPOSAL E. P. Blizard H. L. F. Enlund C. E. Clifford J. L. Meenm A. Simon Physics Division It was recognized in the Shielding Board Report{!) that the calculations leading to divided-shield weight were not on so firm a ground as those leading to the weight of unit shields. Unit shields have subsequently been mocked-up in both the Lid Tank and the Bulk Shielding Facility, and the (1)Report of the ANP Shielding Board, NEPA- ORNL, ANP-53 (Oct. 16, 1950). 1951 INVESTIGATIONS Physics Division resultant weights were in excellent agreement with those predicted by the Shielding Board. A divided-shield 1is now being mocked-up in the PBulk Shielding Facility. However, the information to be obtained from this mockup 1s only the intensity and direction of the radiations emerging from the reactor part of the divided shield. One of the larger uncertainties in divided-shield weights 1s the assumed scattering of the complex spectra emerging from the reactor shield and the effect of thais radiation at the crew shield after having been air-scattered. Since an increase of 1 1b in shield weight entails an increase of 2% 1b in aircraft weight,{(2) considerable effort is justified in specifying shield weights as closely as possible. Full-scale shielding experiments, including air-scattering and crew- shield effects, guickest, optimizing the divided shield. comparison, offer the cheapest, and most accurate method of (By an aircraft design based on calculations would be appreciably heavier because of the reguired conservatism.) Consequently, a proposal(3) for a Tower Shielding Facility to effect full-scale shield- ing experiments has been submitted to the AEC by the Oak Ridge National Laboratory. The proposal 1s to mount a small MTR type water-cooled reactor and shield mockup in a tower with a crew shield mockup supported 50 ft above the reactor. The tower would be tall (2)Report of the Technical Advisory Board to the Technical Committee of the Aircraft Nuclear Propulsion Programr, ANP-52 (Aug. 4, 1950). (3)E. p. Blizard, C. E. Clifferd, and A. Simon, Proposal for Divided Shield Expertiments, ORNL CF-51-11-158 (no date). 73 ANP PROJECT QUARTERLY PROGRESS REPORT enough so that ground scattering would be unimportant, A preliminary cost estimate indicates that the equipment, including tower, security fence, about $694,650. reactor, and roads, building, would cost CIRCULATING-FUEL REACTOR SHIELDS E. P. Blizard and F. H. Murray Physics Division The current designs of circulating- fuel reactors present new problems in divided-shield design. These are being attacked on a cursory basis i1n order to approximate shield weights for the power-plant configurations as they are developed. A preliminary estimate of the required shielding thickness has been derived for the Reactor Design Group.*) Tentative conclusions based upon the layout of the preliminary design show that the crew shield will require about 13 in. (see footnote 5) of plastic on all sides (somewhat less on the front) for shielding of delayed Mg, p. Blizard, Preliminary Estimate of CFR Shielding, ORNL CF-51-11-139 (Nov. 26, 1951). (S)Subsequent more detailed calculations indicate this dimension to be closer to 18 in. 714 neutrons, There appears to be no reason that some of this shielding could not be removed from the reactor shield, but some other modification may be reqguired, for example, the location of gamma shielding around the crew, No calculations have been made concerning the gamma shielding, but for the present it seems safe to assume that the exposed fuel will not consti- tute a source much worse than some cases already considered, namely, the sodium coolant discussed in ANP-53,(1!) It 1is also apparent that the divided shield for the circulating-fuel reactor is not fundamentally different from, but merely a perturbation of, the shield design for the sodium-cooled reactor. NDA DBIVIDED-SHIELD STUDIES Nuclear Development Associates, Inc., are continuing their analysis of divided-shield design. The ANP-53 appear to be adequate 1n although the details of design The final include designs weight, will undoubtedly change. report of their work will specific recommendations for the full- scale divided-shield experiments now contemplated by ORNL. SUMMARY AND Recent interest in the high tempera- ture circulating-fuel reactor has emphasized the need for liquid fuels or low (below 450°C) melting point with relatively low (4 to 8 1b/ft®) uranium concentration. Five fluoride systems fulfilling these requirements have been studied. Research on the chemistry of these and other high-temperature liquids, such as, the development of liquid moderators, self-moderating fuels, and the investigation of fluoride mixtures as possible reactor coolants, are reported im Sec. 10. Several ternary systems of alkali fluorides and beryllium fluoride are shown to be suitable coolants and are also of potential value as solvents for UF, in the development of fuels., Static corrosion tests of fluoride fuels i1indicate that these fluids are readily contained at 1500°F by inconel and any one of a number of stainless steels (Sec. 11). Dynamic corrosion tests in inconel have been similarly successful, but the two stainless steel loops in which the fluorides have been circulated have plugged, presumably as a result of mass transfer. The static and dynamic corrosion of these fluoride mixtures, which are of primary interest to the circulating-fuel reactor, are, except for the dynamic test 1n stainless steel loops, reason- ably satisfactory. Considerable research has been undertaken in attempts to contain the alkali hydroxides in nickel and the structural metals. Static-corrosion tests of hydroxides have shown, however, that only nickel, copper, and the more noble metals will withstand the corrosive action of these media. Furthermore, dynamic- corrosion tests of hydroxides in nickel have indicated severe plugging, also, as with the fluoride in stainless steel, the result of mass transfer. Fundamental studies of this phenomenon INTRODUCTION indicate that oxygen may play an important role in mass transfer, at least in hydroxide-nickel systems. Heat-transfer research (Sec. 12) has been concerned primarily with the determination of heat-transfer coeffi- cients of various systems, including those of the fluorides, hydroxides, and liquid metals. Although satis- factory data have not yet been obtained, mathematical solutions have been derived for the temperature structure in forced-convection fluoride systems. Excellent agreement has been obtained between theory and practice for the natural convection i1n a brine solution simulating the fuel elements; the peak temperature 1s substantially reduced from that computed assuming conduction only. HRoutine measurements are now being made on the heat capacity of materials and the density of high- temperature fluids. Some data are also available on the thermal conductivity of materials and the vapor pressure and viscosity of the fluoride salts. The metallurgical processes involved in the construction and assembly of a high-temperature core, including welding of fuel tubes, fabrication of solid fuel elements, and creep and stress rupture of metals, are currently under investigation {Sec. 13). The fabrication of solid-fuel elements has been further refined by the evalu- ation of the effects of the particle size of the UQ, powder, rolling temper- ature, and elimination of the capsule during hot rolling. Satisfactory spot welding and brazing of these fuel- plate laminates has also been demon- strated. The creep and stress~rupture testing of reactor fuel tubes in sodium has been initiated. Stainless steel tubes have withstood hoop stress up to 2600 psi for 1000 hr at 1500°F in sodium, whereas similarly stressed inconel tubes failed. [ ANP PROJECT QUARTERLY PROGRESS REPORT Although no pronounced radiation damage effects have yet been observed in irradiation of the constituents of the aircraft reactor in the X-10 graphite pile or the Y-12 cyclotron, the one irradiation of a fuel capsule in the higher-flux LITR showed an increase of corrosion products in the fuel, an increase in the decomposition of the fuel, and an 1ncrease 1n 78 corrosion of the capsule (Sec. 14). It is not yet possible to draw any conclusions regarding the significance of radiation damage to the aircraft reactor until confirmatory experiments are completed. A second irradiation of inconel 1in the X-10 pile at temper- atures up to 575°C has not confirmed the decrease in thermal conductivity reported earlier. FOR PERIOD ENDING DECEMBER 190, 10. Warren Grimes, Research in the ANP Chemistry Group has been concerned almost entirely with studies of high-temperature liguids for use as fuels, moderators, and/or heat-transfer fluids for an aircraft reactor. The general prop- erties required of such liquids along with progress 1in development of such materials have been described 1n previous reports (see p. iv for list). The research effort i1s at present concerned with fuel and woderator development with somewhat less emphasis on the heat-transfer fluid (coolant) development program. Hecent interest in a high-tempera- ture circulating-fuel reactor has emphasized the need for liquids of extremely low melting point with relatively low uranium concentration (4 to 8 1b of uranium per cubic foot), Five systems fulfilling these require- ments have been studied, but the program for the development of self- moderating fuels has not produced materials more satisfactory than those previously described.!) Several moderator-cooclant systems have been examined, but this program has not been vigorously pursued pending the demonstration, 1n design studies and corrosion research, of the appli- cability of these systems. Investi- gation of fluoride mixtures as possible reactor coolants have shown several ternary systems of the various alkal1 fluorides and beryllium fluoride to be suitable. (Dy, p. Blakely, G. J. Nessle, L. Bratcher, and C. J. Barten, “Phase Studies of Fluoride Systems,” Aircraft Nuclear Propulsion Project guarterly Progress Report g:r Period Ending une 10, 1951, ANP-65, p. 84 (Sept. 13, 1951). 1351 CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS Materials Chemistry Division LOW MELTING-FLUORIDE FUEL SYSTEMS J. P. Blakely L. M, C. J. Barton Materials Chemistry Division Bratcher Phase equilibrium studies utilizing the technique of thermal analysis(?+3) were directed initially toward develop- ment of fluoride mixtures with high uranium content and melting polnts below 550°C. A previous report(*’ listed nine systems which promised to fit these gqualifications. However, only three of these systems show low melting points in the range 0 to 2.5 mole % UF, and promise to be of use in circulating fuel reactors. This relative scarcity of suitable fuel mixtures along with the desirability of lower melting liquids for such reactors has prompted additional study of systems containing low concentrations of UF,. In addition 1t has been necessary to re-examine 1n greater detail the low-uranium regions of some of the known systems. At present the five systems listed in Table 10.1 are those which appear promising as circulating fuels. The examination of several of these listed systems is far from complete, and 1t (2)¢c, J. Barton, R. E. Moore, J. P. Blakely, and G. J. Nessle, “Low-Melting Fluoride Systems ~-- Thermal Analysis,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period {ggé?g August 31, 1950, OBNL-858, p. 110 (Dec. 4, (3)R. E. Moore, G. J. Nessle, J. P. Blakely, and C. J. Barton, "Low-Melting Fluoride Systems,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Endinf December 10, 1950, ORNL-919, p. 242 (Feb. 26, [951). Oy, R Grimes, “Chemistry of High-Temperature Liquids,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending fggffuber 10, 1951, OBNL-1154, p. 154 (Dec. 17, 19 ANP PROJECT QUARTERLY PROGRESS REPORT Table 10.1 Sunmary of Promising Fluoride Fuel Systems of Low Uranium Comntent LOWEST URANIUM CONC. OF | MELTING SYSTEM MELTING AT POINT COMPOSITION OF LOWEST OR BELOW 500°C FOUND MELTING MIXTURE COMPONENTS (mole % UF,) (°c) (mole %) LiF-KF-UF4 0 - 71 470 40 LiF, 55 KF, 5 UF4 LiF-NaF’-KF‘UF4 0 -6 450 97 (LiF-NaF-KF), 3 UF4 LiF-NaF-RbF-UF, | 0 - 4 425 97.5 (LiF-NaF-RbF), 2.5 UF, NaF-BeF,-UF, 0 - 12 330 47 NaF, 50 BeF,, 3 UF, LiF-NaF-BeF,-UF, 0 - 15 255 30 LiF, 20 NaF, 49 BeF,, 1 UF, is possible that continued study will result in compositions of lower melting point. In addition, it should be emphasized that the search for such fuels was initiated quite recently and it 1is likely that the list will be extended considerably in the future. As i1t is probable that four-component systems will be required for melting points below 350°C, such research will consequently bhe relatively slow, Examination of Table 10.1 indicates that BeF, is a component of the lowest melting systems. Indeed, 1t seems certain that BeF, and LiF will be components of any fluoride system melting below 350°C. The various fuel systems studied are described briefly under appropriate headings below. LiF-KF-UF, Re-examination 1in detail of the low uranium concentration region of this system indicate that the diagram previously presented(s) (5)Figure 13.2, ORNL-1154, op. cit., p. 158. 80 1s substantially correct. The lowest melting point found in this system was 470 + 10°C at 5 mole % UF,, 55 mole % KF, and 40 mole % LiF while the melting point at l mole % UF, would be approxi- mately 485°C. LiF-NaF-KF-UF, A few mixtures of the pseudo- blnary system obtained b¥ treating the LiF-NaF-KF eutectic(®: as the single component have been studied. The data, shown in Table 10,2, indicate that the addition of several mole percent of UF, to this eutectic does not raise the melting point appreciably. This system will be of definite interest and will receive additional study 1f it appears that melting points of the order of 450°C can be tolerated. LiF-NaF-RbF-UF,. A limited number of studies in which this system has been trecated as a pseudo-binary with (8)F, P.Hall and H. Insley, Phase Diagrams for Ceramists, American Ceramic Society, Inc., 1947, (M)A, G. Bergman and E. P. Dergunov, “Fusion Diagram of LiF-KF-NaF,” Compt. rend. acad. sct. 531 753 (1941); Chem. Abstr. 37, 823, FOR PERIOD ENDING DECEMBER 19, Table 10.2 The Pseudo-Binary Systen(“’ (LiF-NaF-KF)-UF, UF, (mole %) MELTING POINT (°C) 0 455 1.1 450 5.0 485 10 597 15 615 20 625 30 630 (2)42 mole %LiF, 11.5mole % NaF, 46.5mole % KF. the alkali fluoride ternary eutectic as one component have been performed. The data in Table 10.3 show that small additions (up to 2.5 mole %) UF, do not appreciably affect the melting point, but additions of larger amounts markedly 1ncrease this property. While the melting points obtained are slightly lower than those of the system immediately above, 1t is not likely that the slight benefit will Jjustify the use of RbF. Table 14.3 The Pseudo-Binary System(fl) (LiF-NaF-RbF) -UF, UF, (mole %) MELTING POINT (°C) 0 444 2.5 425(%) 5 615 10 7124 20 660 30 547 40 600 (242 mole % LiF, 6 mole % NaF, 52 mole ZRbF. (®)This mixture showed signs of supercooling. 1951 NaF-BeF,-UF,. The equilibrium diagram for this system has been presented in a previous report.(®’ A re-examination of the NaF-BeF, system (see section on Coolant Development below) and additional data at low uranium concentrations has resulted in some changes in the contour lines in this portion of the diagram. This system is still under study and a revised diagram will be presented later. Difficulty in establishing true melting points at high BeF, concen- trations was encountered as before as a result of the extensive supercooling of such mixtures. The minimum melt- ing point so far established 1is 330 + 10°C at 3 mole % UF, and 50 mole % BeF,. LiF-NaF-BeF,-UF,. Some low-melting mixtures from the LiF-NaF-BeF, ternary system (see section on Coolant Develop- ment below) have been treated as pseudo-binary systems with UF,. The most encouraging of these results are are shown in Table 10.4. Table 10.4 The Pseudo-Binmary System(“) (LiF-NaF-BeF,) -UF, UF, (mole %) MELTING POINT (°C) 0 245-265 1 255 2.5 245-270 5 405 10 490 20 612 30 675 40 720¢¢8) (4)30 mole % LiF, 20 mole % NaF, 50 mole %BeF,. (6) his sample showed definite supercooling. (8)Figure 4.4, ANP-65, op. cit., p. 91l. 81 ANP PROJECT QUARTERLY PROGRESS REPORY These data show that this system offers the lowest melting point avail- able to date. Although some of these mixtures were stirred at tempevratures well below 300°C, it is possible that extensive supercooling occurred. It 1s not possible to give any estimates of the viscosity at these low tempera- tures although such studies are being made. IONIC SPECIES IN FUSED FLUORIDES M. T. Robinson Materials Chemistry Division Determination of the electrolytic transference numbers in fused mixtures of UF, and alkali fluorides has been attempted in an effort to determine the nature of the 1onic species 1in these systems. In advance of the experiments i1t was thought that the uranium existed in solution as U*’ or as some species of negative complex ion, e.g., UF~ Electrolytic trans- ference experiments should distinguish clearly between these two possibilities but would be less likely to determine, with certainty, details of the nature of species differing sllghtly from those mentioned, e.g., UF*’ or UF, Experimeantal Procedure. The electrolytic vessel consists of a graphite cup acting as the anode 1in which 1s suspended a graphite rod cathode. The anolyte and catholyte are separated by a diaphragm of platinum sponge. The apparatus 1is sealed, evacuated, provided with an inert atmosphere, and maintained at temperature 1n a molten lead bath. The electrolytic current provided by two 6-v storage batteries is maintained at about 0.3 amp for 90 min, after which the apparatus is cooled. The anolyte and catholyte are remcoved separately, dissolved in HNO,;-A1(NO,;), solution, and analyzed for U, F~, and alkali metal ions. 82 Electrolyses have been performed at both 750 and 900°C on a binary mixture containing 28 mole % UF, and 72 mole % NaF. From the chemical analyses before and after the experi- ment, and from a knowledge of the quantity of electricity passed, 1t is possible to derive the transference numbers of the various i1oms present. Results of Electrolysis. It appears that both sodium and uranium are deposited at the cathode and that fluorine is evolved at the anode. No uranium (from anodic formation of UF) has been detected in the vacuum lines of the apparatus. A deposit of NaF has been observed 1in the quartz tube con- taining the electrolysis cell. This is probably due to volatilization of sodium and subsequent reaction with evolved fluorine. The apparent deposition of both metals 1s 1n agree- ment with the reported partial reduction of UF, by metallic sodium.(?’ The data indicate that no appreciable quantity of uranium is transferred into or out of either compartment. If com- plex ions of the type UF:' are assumed to be present, only the assumption that n = 4 yields results i1in which the transference numbers of all i1ons are However, the 4 does not mean that uranium 1s present as un-ionized UF,, but simply that no detectable transfer of uranium takes place. This result is 1n excellent agreement with the theory of Frenkel('%) of the electrical conductivity of fused salts. Briefly, Frenkel states that the conductivity of a fused salt may be taken as due only to the smallest 1on present. The 1onic radii of the Na®, F~, and U** ions are 0.98, 1.33, and between 0 and unity. conclusion that n = (19)5. Frenkel, Kinetic Theorg of Liquids, pp. 439-445, Oxford New York (937, J. Katz and E, Rab1now1tch, The Chemistry of Uranium, Part 1, N.N.E.S., Div. VIII, Vol. 3, p. 125, McGraw- Hill, New York, 1951, FOR PERIOD ENDING DECEMBER 10, 1951 0.97, respectively.(!?) Since the sodium and uranium(IV) ions are about the same size, 1t appears that uranium is complexed with fluoride in some manner, although the precise nature of the complex canmot be clucidated from the presently available data. It 1s 1intended to extend the work to another solution of UF, in NaF and to two solutions of UF, in KF. Since the radius of the potassium ion 1s about the same as that of the fluoride ion, it may be possible to throw more light on the nature of the uraniferous ion from measurements on mixtures containing KF. HOMOGENEOUS FUELS J. D. BRedman and L. G. Overholser Materials Chemistry Division The solubility of uranium trioxide in mixtures of sodium hydroxide and sodium tetraborate with and without additional boric acid has been previ- ously reported.‘!?? It was found that the presence of borate increases the solubi1lity of uranium 1n sodium hydroxide but that the concentration of borate required to give a useful solubility is so high that the melting point of the resulting mixture 1is above 500°C. Additional studies have been made utilizing other hydroxides and borates in an attempt to find a mixture which will dissolve sufficient uranium and still possess a sufficiently low melting point. Nelther the addition of borates nor of any of several other materials appears likely to yield a hydroxide mixture melting below 600°C and containing 2 to 4 wt % dissolved uranium. All experiments (113, Glasstone, Textbook of Physical Chem- istry, p- 375, Van Nostrand, New York, 1940. (12)). p. Bedman and L. G. Overholser, ORNL~1154, op. cit., p. 16 were performed at 650°C with 5 wt % uranium added as uranium trioxide. Uranium Solubility in Hydroxide- Borate Mixtures. The results of some of the experiments in which borate- hydroxide mixtures were used as solvents are summarized in Table 10.5. These data 1ndicate that the solubility of uranium at 650°C is too low to be of any interest. It does not appear likely that any combination of hydroxide and borate will be found having a melting point below 500°C and dissolving 2 to 4% uranium at 600°C. Table 18.5 Soelubility of Uranium in Hydroxide-Borate Mixtures CONSTITUENT IN MIXTURE SOLUBILITY (we %) OF URANTUM NaOH KOH .1 OH B203 (we %) 85 15 0.3 70 30 1.3 95 5 0.3 90 10 0,7 85 15 1.6 80 20 1.5 75 25 (a) 99 8 0.3 84 16 0.9 80 20 0.8 48 A7 5(b) 1.0 43 49 15 1.7 (“)Me]ted above T00°C. (b)yhen 10 wt %cesium fluoride was added to this mixture the solubility was found to be 1.1%. Uranium Solubility in Hydroxides with Various Additives. Some addi- tional experiments have been performed to determine the effect of other materials on the solubility of uranium in sodium and lithium hydroxides. The 83 ARP PROJECT QUARTERLY PROGRESS REPORT results of some of the runs made at 650°C are given in Table 10.6. None of the materials listed shows an appreciable beneficial effect on the solubilivy. Table 10.6 Effect of Various Additives om the Solubility of Uranium in Hydroxides CONSTITUENT IN MIXTURE SOLUBLLITY (we %) OF URANIIM LiCH | NaOH LiF Na28203 RLOH (wt %) 95 5 0.6 48 47 5 0.3 48 47 5 1.0 45 45 10 1.2 40 40 20 0.8 The addition of BRbOH to mixtures of sodium and lithium hydroxide appears to have a slight detrimental effect on the solubility of uranium. Uranium has been shown to dissclve in RbOH to the extent of about 0.1 wt % at 650°C. Addition of uranium trioxide to RbOH does, however, produce a slurry very similar to that obtained with NaOQH®3:14) It should be noted that the RbOH used contained about 5% Rb,CO;. MODERATOR-COOLANT DEVELOPMENT Development of moderator-coolants is still concerned primarily with preparation of pure alkali and alkaline earth hydroxides and with phase equilibrium studies of various mixtures (1335, p. Redman, D. E. Nicholson, and L. G. Overholser, “Suspensions of Uranium Compounds in Sodium Hydroxide,” Aircraft Nuclear Propulsion Project Quarterly Progress Report {or Period figg{?g March 10, 1951, ANP-60, p. 135 (June 19, (14)y,. G. Overholser, D. E. Nicholson, and J. D. Bedman, “Suspensions or Solutions of Uranium in Molten Hydroxides,” ANP-63, op. cit., p. 96. 84 of these materials withother compounds. Since 1t 1s apparent that the most difficult problem 1n use of these materials 1s control of the corrosion, major emphasis in the program 1s in that field (see Sec. 11, Corrosion Rescarch). Relatively pure samples of the bydroxides of sodium, potassium, barium, strontium, and lithium have been prepared. Study of one bimnary hydroxide system and several hydroxide fluoride systems were completed during the past quarter. Pending encourage- ment from corrosion studies no further studies of this kind seem justified at present. Preparation of Pure Hydroxides (L. G. Overholser, D. E. Nicholson, E. E. Ketchen, and D. R. Cuneo, Materials Chemistry Division). Sodium hydroxide has been purified in pound batches by removing sodium carbonate from a 50% aqueous solution of NaOH followed by dehydration. The residual sodium carbonate concentration 1is about 0.5 wt %. Attempted methods for purifying potassium hydroxide have included recrystallization of this material from isopropyl alcohol and precipitation of carbonate as barium carbonate in aqueous medium. While not completely satisfactory, the former method has proved the bhetter, yielding potassium hydroxide with a total alkalinity greater than 99% and with less than 0.5 wt % potassium hydroxide. Approximately 10 1b of barium hydroxide has been purified by removing the carbonate 1n aqueous solution, recrystallizing barium hydroxide octahydrate, and dehydrating under carefully controlled conditions. The resulting samples have contained less than 0.4 wt % contaminant 1n the form of either the oxide or carbonate. Strontium hydroxide is being purified in an analogous manner. Commercial lithium hydroxide monohydrate has been found to be quite pure on dehydration. FOR PERIOD ENDING DECEMBER 198, Kou-LioH (J. P. Blakely, L. M. Bratcher, and C. J. Barton, Materials Chemistry Division). This system . closely resembles the NaOH-LiOH system previously described.('3) The eutectic point at 70 mole % KOH is at 225°C while the incongruently melting com- pound appears unstable above 315°C. The series of halts at 245°C on the KOH side of the eutectic, as shown 1in Fig. 10,1, is probably due to the solid transformation of KOH reported by von Hevesy(!®) at 248°C, 600 F T | T T - 500 - D \ e w400 - . 399 q & 300 7 © o | a ~o. : § M . 200 @ o _] 100 L < L L 0 10 20 30 40 50 &0 70 'O 30 100 i KOH LicH KOH (mole %) Fig. 18.1i. The System KOH-LiOH. Hydroxide-Fluoride Systems (J, P. Blakely, L. M., Bratcher, and C. J. Barton, Materials Chemistry Division). The relatively few determinations made in the system RbOH-RbF indicate that these materials, like the analogous sodium and potassium compounds, form solid solutions with the melting point of the mixture rising sharply up to about 50 mole % RBRbF and then somewhat less steeply in the range 50 to 100 mole % RbF. The pseudo-binary systems NaOH- (LiF-NaF-KF) and KOH-(LiF-NaF-KF) in which the alkali fluoride composition represents the ternary ecutectic have been studied to ascertaln whether (15)g A. Allen and W. C. Davis, “Binary Hydroxide Systems,” ORNL-1154, op. cit., p. 1 (163G, . Hevesy, “Uber Alkalihydroxyde. I,” Z. physik. Chem, 73, 667 (1910). 1551 low-melting liquids of low hydroxide content can be prepared. It is apparent that up to 15 mole % NaOH or up to 30 mole % KOH can be added to the alkali fluoride eutectic without elevation of the melting point. CGOLANT DEVELOPMENT J. P. Blakely L. M. Bratcher C. J. Barton Materials Chewmistry Division Investigation of low-melting fluoride mixtures of possible appli- cation as coolants has been continued during the past quarter. While a few binary mixtures have been examined, largely to fill gaps in the available literature, most of these studies have been concerned with termary systems of the various alkali fluorides and beryllium fluoride. In addition to the value of such data in coolant development, any low-melting composi- tions which are found are of potential value as solvents for UF, 1in the development of fuels of low uranium content. BbF-LiF. This binary system has been examined i1n order to complete the ternary systems containing these com- pounds. As indicated in Fig. 10.2, this system forms a simple eutectic melting at 470 % 10°C at 42 mole % LiF. RESTRICTED D'AG 12528 sco T i T T T T _ 8OO g W 700 - @ D = W 600 = w = 500 / T € g @ 400 ) 1 b 1 I i I L 0 0 20 30 40 S0 &0 70 A0 30 W00 RbF LiF {mole %) LiF Fig. 10.2. The System RbF-LiF, 85 ANP PROJECT QUARTERLY PROGRESS REPORT NaF-BeF,. The phase diagram from the open literature(®) has been used in construction of termary diagrams involving these materials. Various di fficulties with some of these diagrams has, however, prompted re- examination of this system. Figure 10.3 shows the results obtained. Below 50 mole % BeF, the published diagram was essentially confirmed although some minor differences were noted. In the region above 50 mole % BeF, thermal data are apparently un- reliable. The dotted curve in this region shows the quenching data of Roy, Roy, and Osborne. (17) RISTRICTEG LWG.13589 L © © - Ll & 2 < g TN e x o > K : = X o © i -0 ¢ e o % = : o e e o o a 60 vo 8O S0 100 BeF Naf BeF, (mole %) ¢ Fig. 10.3. The System NaF-BeFQ. NaF-KF-RbF. The lowest melting point found in this system 1s at 621 + 10°C at 74 mole % RbF and 21 mole % NaF. The failure to find a low-melting region on this diagram (Fig. 10.4) is probably due to the high melting points of the solid solutions of RbF and KF. KF-LiF-RbF. The equilibrium diagram shown 1n Fig. 10.5 indicates that LiF is considerably more effective than NaF in lowering melting points 1in the KF-RBbF system. The data on this system are still far from complete and (17)8. Roy, D. M. Roy, and E. F. Osborne, J. Am. Cerem. Soc. 33, 85 ¥1950). 86 RbF RESTR-CTED 795°( CwG 13530 KF 850°C The System NaF-KF-RbF, RbF RESTRICTED 5 795°G OWG 13531 S o o\oolo 1 3 ° §5g/8 [ g $ o @ KF / l LiF 850°C 285 G 845°C Fig. 10.5. The System KF-LiF-RbF, further study may change the contours considerably. The lowest melting point so far discovered 1s 440 = 10°C at 27.5 mole % RbF and 40 mole % LiF. NaF-LiF-BbF, The incomplete data on this system have been used to plot the diagram shown in Fig. 10.6. Since all three binary systems form simple eutectics, the melting point available with this system is considerably lower than that with the two 1mmediately preceding. The ternary eutectic appears to melt at about 425°C and " FOR PERIOD ENDING DECEMBER 10, 1951 to be near 42 mole % LiF and 6 mole % NaF. RAbF RESTRILTED 795°¢ .9 750 100 875°C 650 ey ) § B 4T0°G % % ® &£ %, &3 &9 @ S$ B 5L ®, 7 % & NaF \ ” 1 LiF 995G 650°C 845°C Fig. 10.6. The System NaF-LiF-RbF. Ternary Systems Containing BeF,. Three ternary systems containing BeF, are under investigation at present, Study of these systems is difficult both because of the precautions necessary in handling the Bel, and because of the supercooling and generally poor thermal behavior of the melts. The lowest melting points so far observed are at 345°C in the NaF-KF-BeF,, 325°C in the LiF-KF-BeF,, and 220°C (possibly considerably supercooled) in the LiF-NaF-BeF, systems. Study of these materials is continuing because of their possible use in very-low-melting fuel systems. SERVICE FUNCTIONS G. J. Nessle V. C. Love H. S. Powers C. J. Barton Materials Chemistry Division The Fuel Preparation and Service Group was called upon to fill a wide variety of containers during the past quarter. The amount of fuel required ranged from 0.5 to 3900 g. The 3900-g batch of material required the con- struction of special equipment for melting and filtering the fused salt. This equipment has been used several times successfully. The filtering material 1s sintered stainless steel, Table 10.7 shows the number of con- tainers filled for different types of experiments, Table 10.7 Container Filling Services NO. OF CONTAINERS TYPE OF EXPERIMENT FILLED Static corrosion tests 112 Cyclotron bombardment 31 Radiation damage 23 Loop tests 2 Circulating system 1 Density measurement 1 In addition to the container- filling services, about fifty batches of fused fluoride mixtures have been prepared for various uses. This group has also started tests designed to elucidate the mechanism of fluoride fuel stabilization and to indicate the best method of carrying out this stabilization. 87 ARP PROJECT QUARTERLY PRGGRESS>REPBRT 11. CORRCSION RESEARCH H. W. Savage ANP Division W. D. Manly Metallurgy Division W. R. Grimes Materials Chemistry Division During the past quarter a consider- able change in emphasis occurred in the corrosion testing program. While fluoride fuels and coolants continued to receive some attention, a major effort was placed on attempts to contain the alkali hydroxides in nickel and the structural metals. Corrosion tests have been run at 800 to 815°C in various nonmetallic media, including the molten hydroxides of barium, stromium, sodium, lithium, and potassium and molten sodium cyanide. Some tests were run on the liquid metals sodium and lead to check previous results. In the corrosion tests with the hydroxides 1t was found that only nickel, copper, and the more noble metals silver, gold, and platioum will withstand the corrosive action of these media. Therefore coatings of copper and nickel on structural mate- rials, stainless steels and inconel, were tried with variable results. Tests are underway to determine the minimum coating thickness needed. The hydroxides exhibit considerable metal transport {mass transfer) in tests when there 1s a temperature gradient across the tube and the hydroxides are flowing. Sodium, potassium, and lithium hydrdxides, circulated in nickel thermal-convection loops at 1400°F, each provoked suffi- cient mass transfer to plug their respective loops. This mass transfer phenomenon 1s being studied with thermal-convection loops of two different designs and with two addi- tional pieces of equipment, a thermal gradient standpipe furnace and a 88 "seesaw" furnace in which the hydrox- ides are sloshed back and forth from the hot to the cold portion of a tube. In addition, work has been started on a fundamental approach to the problem of the reaction of the hydrox- ides with metals at high temperature. Some of the first tests tobhe run in the NaOH-nickel system demonstrate the role oxygen plays 1in the metal trans- port phenomenon; in thermal-convection loops the presence of oxygen will increase the amount of mass transfer. Results from thermal gradient stand- pipe tests have shown the effect of hydrogen atmosphere onmetal transport. Considerable research will be needed, however, before the transport phenomenon will be understandable. There i1s no evidence of unusual problems concerning the behavior of fluoride mixtures 1n static tests with structural metals. Such tests have demonstrated that the pretreated fluoride mixtures can be contained in capsules of inconel or 316 stainless steel. Dynamic-corrosion tests of fluorides in inconel convection loops have been completely satisfactory although similar tests 1w stainless steel loops have shown appreciable mass transfer. There is evidence that the iron and nickel content of the pretreated fuel decreases after 100 hr of exposure 1n the structural metal while the chromium content of the fuel increases. This phenomenon, however, is not completely certain and 1s being investigated further because of the light i1t may shed on the pretreatment problem. FOR PERIOD ENDING DECEMBER 10, STATIC CORROSION BY FLUOGRIDES H. J. Buttram C. R. Croft N. V. Smith J. M. Didlake Materials Chemistry Division No information was obtained which would change the previously reported(!’ conclusions that properly treated fluoride mixtures present no serious corrosion problems with structural metals, such as inconel and stainless steel, 1n static tests for periods up to several hundred hours. This applies also to the low-uranium-content fuels, as well as to the uranium-free alkali fluoride mixtures. Exposure at temper- atures up to 900°C for 100 hr also yielded satisfactory results. Although the fluorides were pre- treated in the majority of cases, untreated fluorides did not seriously attack inconel test specimens. Pre- treatment with a single metal rather than by alloys appears to be feasible. The mechanism of the pretreatment process 1s not completely clarified yet, but it 1s receiving further attention. Mixtures containing lead fluoride (60 mole % NaF, 23 mole % PbF,, 17 mole % UF,) were subjected to the standard corrosion test. Materials containing PbF, are unsatisfactory. The mixture was found to be unstable and the metallic lead which formed alloyed with the container metal. No further tests are planned with mixtures containing lead fluoride. Stabilization of capsules of stain- less steel and inconel by pretreatment with elemental fluorine was attempted wlthout significant effect. No difference was observed in corrosion (s gtatic Corrosion by Fluoride Fuels,” Air- craft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 138514, OB§L91154. 1951 resistance to fluorides of these materials as compared with untreated specimens. STATIC CORROSION BY HYDROXIDES W. D. Manly, Metallurgy Division F. Kertesz, Materials Chemistry Division A number of corrosion tests have been made at 800 and 816°C for 100 hr in nonmetallic media which might find possible application as reactor moderators or coolants. These include the molten hydroxides of barium, sodium, strontium, lithium, and potas- sium. JThe media were dehydrated commercial products and especially pure sodium, potassium, and barium hydroxides. Purified sodium hydroxide appeared somewhat less corrosive than did the dehydrated commerical material; similar benefits do not appear from purification of potassium and barium hydroxides. Without reference to change in mechanical properties it would appear that potassium hydroxide 1s less corrosive on inconel than is sodium hydroxide. Changes 1in mechanical properties of the test specimens have not been examined. In these tests, however, nickel and copper resisted chemical reactionm with most of the hydroxides. Stainless steels coated with copper and nickel were therefore a logical development to combine strength with corrosion resistance. It has been found, however, that thin coatings are not effective 1n suppres- sing corrosion whereas heavier coatings do show promise. Tests are underway to determine the minimum coating thickness required and the factors which influence the adherence of these coatings. The temperature dependence of the attack was also investigated; at 500°C potassium hydroxide showed only a very 89 ANP PROJECT QUARTERLY PROGRESS REPORT slight penetration which increased somewhat at 600°C and became much stronger when temperatures of 700 and 800°C were reached. As far as time was concerned, 1t appeared that most of the attack occurred during the first few hours of the exposure. Corrosion of Uncoated Metals (A. D. Brasunas, D. C. Vreeland, E. E. Hof fman, and R. B. Day, Metallurgy Division). Several metals were tested inmolten hydroxides of sodium, lithium, potassium, barium, and strontium for 100 hr and at temperatures up to 816°C. These tests were conducted in evacuated containers of the same composition as the metal specimen., The results with barium, strontium, sodium, and lithium hydroxides are summarized in Tables 11.1, 11.2, 11.3, and 11.4, respec- tively. Most metals and alloys are susceptible to hydroxide corrosion 1in varying degrees, the notable exceptions being nickel, copper, and monel. : These vresults were obtained with especially purified sodium hydroxide, while the potassium hydroxide was a dehydrated commercially pure product. There was evidence of decreased corrosion i1n sodium hydroxide which Table 11.1 Summary of Corrosion by Ba(OH), at 816°C for 100 hr All specimens tested in evacuated capsules of like materials DEPTH OF METAL WT. CHANGE AFFECTED MATERIAL (g/in.z) {(mils) METALLOGRAPHIC NOTES 304 stainless steel 5.5 0.010-in. oxide layer 310 stainless steel +0, 046 3.5 0.005-in. oxide layer 316 stainless steel 9.0 0.010-in. oxide layer 318 stainless steel -0.061 4.5 0.006-in. oxide layer 321 stainless steel -0.740 4.5 0.005-in. oxide layer 347 stainless steel -0.083 4.7 0.006-in. oxide layer 446 stainless steel 5.0 0.008-in. oxide layer (Fig. 11.5) Copper (O,F.H.C.) -0.003 0.5 No evidence of attack Iron (Globe) -0.994 8.5 Heavy oxide formed Zirconium -0.221 9.0 0.001-in. gray oxide layer; voids observed to depth of 0.004 1in. Hastelloy B -1.109 9.5 Heavy oxide formed Inconel X 6.5 0.020-in. oxide layer (Fig. 11.6) Nickel Z 6.0 Erratic attack,0.002 in. average, (Fig. 11.7) 0.006 in. maximum Monel -0.005 0.0 No evidence of attack 90 FOR PERIOD ENDING DECEMBER 10, 1951 Table 11.2 Summary of Corrosion by Sr(0H), at 816°C for 100 hr All specimens tested in evacuated capsules of like material except where noted DEPTH OF METAL AFFECTED MATERIAL (mils) METALLOGRAPHIC NOTES 304 stainless steel 12.0 Heavy oxide layer; specimens cracked on bending 310 stainless steel 4.7 Very brittle oxide layer 316 stainless steel 3.0 0.003-in. oxide layer 318 stainless steel 2.9 0.006-in. oxide layer 321 stainless steel 4.5 0.006-in. oxide layer 347 stainless steel 7.0 0.020-in. oxide layer 446 stainless steel 4.0 0.011-in. oxide layer Iron (Globe) 2.0 Heavy oxide layer, 0,017 in., on speci- men (Fig. 11.8) Inconel X 3.0 0.007-1in. oxide layer Hastelloy B Specimen heavily oxidized and embrittled Copper (O.F.H.C.) 0.0 No evidence of attack Nickel A No evidence of attack Nickel Z 0.002-in. layer of attacked zone Zirconium™ 7. 0.001-in. oxide layer *Tested in nickel capsule. could be ascribed to purification. Potassium hydroxide did not show any improvement when especially purified. The depth of attack on the different stainless steels tested varied from 2.9 mils [Sr(OH), on 318 stainless steel] to 24.5 mils (LiOH on 304 stain- less steel). Typical oxide layers are shown in Figs. 11.1 and 11.2 which 1llustrate the attack by barium hydrox- ide on 304 and 446 stainless steel, respectively. Inconel 1s attacked to a depth of 6.5 mils by barium hydroxide with the formation of a 20-mil oxide layer (Fig., 11.3). The attack by barium hydroxide on nickel Z was erratic, with a maxipum depth of attack of about 6 mils (Fig. 11.4). Chromium, while suffering a uniform surface solution to a depth of 3 mils in sodium hydroxide, showed no evidence of this attack (Fig. 11.5). The attack by lithium hydroxide on inconel seems to be of a different nature, as can be seen in Fig. 11.6. The penetra- tion does not follow the grain bound- aries but shows a wavelike pattern. On the basis of the tests performed to date, dehydrated RbOH attacks inconel 91 ANP PROJECT QUARTERLY PROGRESS REPORT Table 11.3 Summary of Corrosion by Sodium Hydroxide at 816°C for 100 hr Specimens tested in evacuated capsules of like material except where noted DEPTH OF SPECIMEN WT. CHANGE AFFECTED MATERIAL (g/in.?) (mils) METALLOGRAPHIC NOTES 304 stainless steel 24.5 0.025-in. oxide layer Chromium -0.1 3.0 No evidence of attack; uni- form solution only (Fig. 11.9) Zirconium® -2.0 82.0 Very severe attack; specimen almost completely consumed Nickel A +0,001 0.0 Slightly attacked Nickel Z 0.6 20.0 Severely attacked Al,0, (sapphire)* 2.3 Uniform solution; attack not severe *Tested in a nickel capsule. Table 11.4 Summary of Corrosion by Lithium Hydroxide at 816°C for 188 hr All specimens tested in evacuated capsules of like material DEPTH OF SPECIMEN WT. CHANGE AFFECTED MATERIAL (g/in. 3 (mils) METALLOGRAPHIC NOTES 304 stainless steel -0.185 4.5 0.014-1in. oxide layer Inconel +0,043 4.0 0.006-1n. oxide layer Nickel A -0.001 0.0 No evidence of attack more severely (Fig. 11.7) than KOH under the standard test conditions.¢?’ (H. J. C. R, Croft, Effect of Exposure Tinme Buttram, N. V. Smith, (2)F; gure 10.11, ORNL-1154, op. cit., p. 115. 92 and J. M. Griffin, Materials Chenistry Division). In order to have a standard to which subsequent results might be compared, two series of experiments were initiated. They consisted of a study of the attack of sodium and potassium hydroxides on inconel as a function of exposure time, using 800°C FOR PERIOD ENDING DECEMBER 10, 1951 UNCLASSIFIED Y-4696 Fig. 11.1. Surface of 304 Stainless Steel after 100 hr of Exposure to Barium Hydroxide at 816°C. UNCLASSIFIED Y-4701 Fig. 11.2. Surface of 446 Stainless Steel Specimen after 100 hr of Exposure to Barium Hydroxide at 816°C. 93 ANP PROJECT QUARTERLY PROGRESS REPORT r’ / i. Wmmwmfix‘.uw 11. 3. Fig. posure to Barium Hydroxide at 816°C. as the reference temperature. The periods of exposure were 1, 25, 50, 100, 250, 500, and 1000 hr. These tests are complete except for the 1000-hr runs. Figure 11.8 demonstrates the corrosion of inconel by sodium hydrox- 94 s S o M ol UNCLASSIFIED Y-4700 Surface of Inconel Specimen after 100 hr of Ex- ide at 800°C as a function of time. After 1 hr the metal is penetrated to a depth of 3 to 4 mils; this increased after 50 hr to 3 to 10 mils. After 100 hr a nonmetallic layer is visible under the corroded metallic layer. Little additional attack is evident after 250 hr. FOR PERIOD ENDING DECEMBER 10, 1951 UNCLASSIFIED Y-4732 Fig. 11.4. Surface of Nickel Z Specimen after 100 hr of Exposure to Barium Hydroxide at 816°C. Fig. 11.5. Surface of Chromium Specimen after 100 hr of Exposure to Sodium Hydroxide at 816°C. 95 ANP PROJECT QUARTERLY PROGRESS REPORT UNCLASSIFIED T-497 Fig. 11.6. Surface of Inconel Specimen after 100 hr of Exposure to Lithium Hydroxide at 800°C. UNCLASSIFIED ) T.347 . 2 . o . . \ N % .’ T o 4 . o % . 4 P C - = : . . ¥ . - . A - * * < * .1 'L - . ’ i : . < . . L8 0 b : & , /o : » s . - e T . - - " , s . -I .‘ - \ . )... . "k‘« A\ 2 b - R ’ g i - A T A e & f\ * 3 r e e e, ML B ¢ T IES Fig. 11.7. Surface of Inconel Specimen after 100 hr of Exposure to Rubidium Hydroxide at 800°C. 96 FOR PERIOD ENDING DECEMBER 10, 1951 UNCLASS | F IED { hr 50 hr Fig. 11.8. Hydroxide at 800°C. ORIGINAL SURFACE £ :a . - ". . . . & : ; 3 . -.' " . . = -,y ¢ . P " - )W el ™o : i .I_ & e - * - A Tal N 8 - . L 1 {:" ‘ & Vet [ .j. i X~Hk? i{a‘n v 2y i - ‘\ 2 » o » s g 5 * 4 L ."‘ ," -j! . { hr 50 hr Fig. 11.9. Hydroxide at 800°C. Figure 11.9 shows the analogous attack on inconel at 800°C by potassium hydroxide. After 1 hr the penetration is only 1 to 2 mils, and relatively slight additional attack occurs during succeeding test periods. Measurements indicate 3% weight loss of the specimen after 500 hr. Apparently KOH is less 100 hr 250 hr Effect of Exposure Time on the Corrosion of Inconel by Sodium UNCLASS I FIED 100 hr 250 hr Effect of Exposure Time on the Corrosion of Inconel! by Potassium corrosive than NaOH under these conditions. Effect of Temperature (H. J. Buttram, N. V. Smith, C. R. Croft, and J. M. Griffin, Materials Chemistry Division). Of equal importance is the effect of the test temperature. It was desirable 97 ANP PROJECT QUARTERLY PROGRESS REPORT to determine the temperature at which only negligible attack occurs during the usual period of 100 hr. Figure 11.10 shows the behavior of potassium hydroxide in inconel at 500, 600, 700, and 800°C. At 500°C there is no visible attack. At 600°C there is localized attack. In the sample shown a penetration of 5 to 6 mils can be observed. This is not found to be the case 1n other portions of the test specimen and 1s pictured here to illustrate occasional erratic be- havior of the hydroxide-metal systems. At 700°C there is some pitting and penetration which does not always follow the grain boundaries. The attack at 800°C is considerably worse. Weight loss data (plotted in Fig. 11.11) indicate a strong increase with increasing temperature. A series of thermal standpipe tests 1s 1n progress using potassium hydrox- ide at 538 to 815°C in an attempt to determine the rate of reaction at these temperatures and the effect of the types of atmosphere above the hydroxide on the rate. Vacuum, air, and hydrogen are being used. Preliminary results in air and vacuum indicate that both 600°C 500°C Fig. 11.10. Effect of Temperature Hydroxide in 100 hr. 98 316 stainless steel and inconel are attacked. The depth of attack at 538°C is of the same order of magnitude as that noted i1n a static corrosion test at 816°C, i.e., around 0.006 in. In hydrogen, however, no corrosion product was noted on either 316 stain- less steel or inconel at 538°C. At 704°C shallow corrosion had occurred. Verification of the above preliminary results is being sought. Corrosion of Coated Metals. Nickel and copper have been found to be among the few metals which are resistant to molten hydroxides. Therefore these materials have been applied as protec- tive coatings on stainless steels for use 1n contact with the hydroxides. Preliminary results indicate that corrosion protection can be achieved in this manner. Type 304 stainless steel was found to be severely attacked by sodium hydroxide at 815°C in a 100-hr test. The depth of metal showing oxidation by the molten caustic was 25 mils. Consequently, it was decided to evaluate the protective coatings of nickel and copper by depositing them on this steel. UNCLASSIFIED o l; i i ® G . < o o8 . . . VaRieE . S e . = s R e s O A ,\1._-.": \'c -(_' ;" -) .l : C‘ o . o g % .," e \ » - - 2 B © g - te.. ' Jj wee EEL O * .., 23 ? _"_'fl '.“.‘ in 3 g Ve " : \ ;;' { n . By L et SRy ‘ o . .~::‘" : l':'. o t . .’ ", : b .Y - . o : o o 0O C 800C on the Corrosion of Inconel by Potassium FOR PERIOD ENDING DECEMBER 10, 1951 UNCLASSIFIED DWG. 43629 l 1 T ? ! { — i /- / 100 | ) 1.0 / i 90_? // —403 / 80 t— 0.8 n}}— 0.7 mq /dm>/day 60 50 0.5 40 0.4 mg /dmz/duy 30 Q.3 20 — 1 500 600 700 TEMPERATURE ({°C) 800 Fig. 11.11, Weight Loss of Inconel Specimen in Potassium Hydroxide for 100 hr as a Function of Temperature. Electroplated Specimens. A 3-nmil copper-plated coating on 304 stainless steel reduced attack by sodium hydrox- ide to 5 mils in 100-hr tests at 816°C. A 3-mil nickel plate also resulted in a reduced attack of 9 mils on the underlying stainless steel. Copper-plated specimens were tested in copper capsules, and nickel-plated specimens were tested in nickel capsules. Electroplated specimens tested in sodium hydroxide appeared to have poor plate adherence while those tested in barium hydroxide were very adherent. The data are summarized in Table 11.5. 2 In tests with barium hydroxide for 100 hr at 816°C the 3-mil nickel plate on 304 stainless steel offered good protection. The 10-mil oxide layer which was obtained with the uncoated specimen (Fig. 11.1) is almost entirely eliminated although a small amount of oxidation, as shown in Fig. 11.12, was observed. With the 3-mil copper plate the specimens were severely attacked. with the exception of the corners, which were relatively unattacked. This may possibly be attributed to the heavier electroplate which occurs Table 11.5 Summary of Hydroxide Corrosion of Clad Metal Specimens at 816°C for 100 hr STAINLESS DEPTH OF METAL WT. CHANGE STEEL COATING BATH ATTACKED (in.) (g/in. D) REMARKS 304 None NaOH 0.025 (uniform) Uniform attack 304 3 mils of Cu NaOH 0.005 (uniform) +0.094 [Cu plate not adherent 304 3 mils of Ni NaOH 0.009 (uniform) -0.094 {Ni plate not adherent 304 None Ba (OH) , 0.029 (uniform) -1.0 ‘Uniform attack 304 3 mils of Cu Ba (OH) , 0.094 (maximum) -1.92 Adherent Cu plate 0.027 (average) 0.001 (minimum) 304 3 mils of Ni Ba (OH) , 0.001 (uniform) +0.002 Adherent Ni plate 99 ANP PROJECT QUARTERLY PROGRESS REPORT UNCLASSIFIED Y-4815 Fig. 11.12. fication 500X; reduced to 96%. at the corners. Heavier copper coatings will be tried in order to test this possibility. Nickel-clad 316 stainless steel and nickel-clad inconel Clad Specimens. specimens were prepared using the "picture frame" technique and tested in sodium hydroxide for 100 hr at 816°C. The nickel coating was approx- imately 5 mils thick at the sides and appreciably heavier at the ends. The thin sections of nickel were attacked but the thicker end sections were not (Fig. 11.13). There is no satisfactory explanation for this behavior at this time. On inconel and 316 stainless steel a heavier nickel cladding (10 100 Surface of 304 Stainless Steel Coated with 3-mil Nickel Electro- plate after 100 hr of Exposure to Barium Hydroxide at 816 °C. Original magni- mils) produced from nickel powder gave excellent protection from corrosion 1in NaOH at 816°C for 100 hr (Fig. 11.14). No evidence of oxidation 1s apparent. Dissolution of Metals in Sodium Hydroxide (C. R. Croft and N. V. Smith). Attempts have been made to determine the amounts of various metals dissolved in sodium hydroxide under 1inert atmospheres as a functionof temperature and exposure time. The sodium hydrox- ide is melted under an argon atmosphere in a container fabricated from the test metal, or of nickel in the case of hard-to-fabricate metals such as chromium. Filings of the test metal FOR PERIOD ENDING DECEMBER 10, 1951 11.13. Fig. Sodium Hydroxide at 816°C. 50X were present in all cases to increase the rate at which equilibrium could be established. The sample was main- tained at temperature for 4 hr before a sample was drawn from the container through a filter of graphite to avoid the dispersed but undissolved metal. UNCLASSIFIED Y-4948 ATTACKED Ni COATING ON OUTSIDE 316 ON INSIDE Nickel-Clad Inconel (Nickel Sheet) after 100 hr of Exposure to The metal content of the hydroxide was then determined by analysis. Table 11.6 lists the results obtained to date. The oxidation state of the dissolved metal and the mechanism of process are as yet unknown. 101 ANP PROJECT QUARTERLY PROGRESS REPORT i /A - \ | L w Fig. 11.14. Nickel-Clad Inconel (Nickel Powder) after 100 hr of Exposure to Sodium Hydroxide at 816°C. 100X Table 11.6 Metal Content of Sodium Hydroxide as Function of Temperature METAL CONTENT (wt %) 4 HR AT 4 HR AT 4 HR AT 24 HR AT 24 HR AT METAL 400°C 600°C 700°C 400°C 600°C Copper 0.15 0.19 0.6 0.16 0.54 Nickel 0.14 0.15 0.26 Chromium 0.40 1.94 0.60 Iron 0.17 1.07 1.17 STATIC CORROSION BY with moderating properties. Results FLUORIDE-HYDROXIDE SYSTEMS H. J. Buttram C. R. Croft N. V. Smith J. M. Griffin Materials Chemistry Division Mixtures of sodium hydroxide—sodium fluoride and potassium hydroxide— potassium fluoride were studied because of their potentialities as coolants 102 obtained show some reduction of corro- sion on inconel and stainless steel, but it appears that the improvement is mostly due to the dilution of the hydroxide by the fluoride. This fluoride addition cannot be great for otherwise the melting point will be above the desired range. The systems are mutually soluble in each other without evidence of eutectic formation. " FOR PERIOD ENDING DECEMBER 10, 1951 STATIC CORROSION BY SODIUM CYANIDE A. D. Brasunas R. B. Day E. E. Hoffman D. C. Vreeland Metallurgy Division Numerous molten salts have been considered for use 1in a reactor, but cyanides do not appear to have been mentioned. This nonoxidizing salt is quite stable at high temperatures and has been used extensively in the heat treating and carburization of metals. Pure sodium cyanide melts at 564°C (1047°F); however, this tempera- ture can be lowered by suitable addi- tions. The specific heat of sodium cyanide is reported as 0.25 Btu/lb (solid) and 0.40 Btu/lb (liquid); the heat of fusion is 135 Btu/lb. Uranium is appreciably soluble in sodium cyanide at 816°C. A number of static corrosion tests on a variety of metals and alloys were run at 816°C for 100 hr using evacuated capsules. The results are given in Table 11.7. The cyanide reaction consists in carbon and/or nitrogen absorption by the metal. This frequently results in embrittlement as determined by a simple bend test at room temperature. There is, of course, the possibilaity that embrittlement does not occur at high temperatures but is merely a room- temperature phenomenon. The superiority of nickel and high nickel alloys (inconel and nichrome V) in resistance to carburization is well known. Iron also appeared to be quite inert to the molten cyanide, while stainless steels showed varying tendencies for reaction. The results obtained with iron were somewhat surprising. The lack of carburization and subsequent embrittlement could be caused by the absence of oxygen which is essential to this reaction. (3)4.5.M. Handbook, p. 694, 1948. Silicon additions are known to be very potent in suppressing carburi- zation of commerical alloys that are ordinarily susceptible to carbon adsorption. STATIC CORROSION BY LIQUID METALS A. D. Brasunas R. B. Day E. E. Hoffman D. C. Vreeland Metallurgy Division Corrosion tests of up to 1000 hr of sodium on 315 stainless steel and lithium on coated 304 stainless steel have been completed. The attack by sodium was negligible even at tempera- tures up to 1000°C and did not appear to increase after the first 100 hr. Molybdenum and chromium coatings on stainless steel have somewhat, though not yet sufficiently, reduced the attack by lithium. Sodium on Stainless Steel. Long- time tests in liquid metals initiated some time ago have recently been evaluated. Tests have been run with 316 stainless steel in contact with sodium up to a thousand hours at 816 and 1000°C. The attack was negligible at both temperatures. Static corrosion tests have also been made on 347 and 316 stainless steel and inconel in sodium at 650, 704, and 816°C for 100 hr. Attack in these tests was also negligible, although some embrittlement was noted. The results are summarized in Table 11.8. Lithium on Coated Stainless Steel. Type 304 stainless steel specimens were thinly coated with molybdenum and chromium by the Linde Air Products Company. These specimens were tested in lithium at 1000°C for 100 hr in the usual manner. The attack of the under- lying metal, although somewhat mini- mized, was not eliminated as much as had been hoped. It is believed that 103 ANP PROJECT QUARTERLY PROGRESS REPORT Table 11.7 Corrosion Data Obtained at 816°C Using Molten Sodium Cyanide for 100 hr DEPTH OF METAL AFFECTED METAL (in.) EMBRITTLED REMARKS Iron (Globe) 0.000 No No evidence of attack Nickel A 0.000 No No evidence of attack Nickel Z 0.005 Yes 0.002-in. outer layer; gray constituent formed to depth of 0.005 in. Inconel 0.003 No Integranular penetration of gray constituent to 0,003 in. Hastelloy B 0.020 Yes 0.002-in. outer layer, nitride needles throughout Hastelloy C 0.003 0.003-in. outer layer only (nitride?) Nichrome V 0.005 No 0.00l-in. outer layer; carbu- rized to 0,005 in, Uranium 0.01 Appreciable solution Beryllium 0.002 Gray layer formed to depth of 0.002 1in. 405 stainless steel 0.002 Yes 0.002-1in. layer of scattered voids 430 stainless steel 0.020 Yes Completely carburized; voids to depth of 0.002 1in. 446 stainless steel 0.003 Yes 0.001-in. outer layer; 0.002-in. carburized layer beneath 304 stainless steel 0.011 Yes 0.001-in. outer layer; 0.0ll-1in. carburized layer beneath 316 stainless steel 0.007 Yes 0.002-in. outer layer; 0.007-in. ’ carburized layer beneath 347 stainless steel 0.007 Yes 0.002-in. outer layer; 0.007-in. carburized layer beneath 310 stainless steel 0.012 Partly 0.002-in. outer layer; 0,012-in, carburized layer beneath thicker coatings would probably be more planned in order to check this possi- effective and additional tests are Dbility, 104 Ly Static Corrosion FOR PERIOD ENDING DECEMBER 106, 1951 Table 11.8 Tests in Scdium TEMP., TIME OF TEST ¥T. CHANGE MATERIAL (°C) (hs) (g/in.2) REMARKS 316 stainless steel | 1000 1000 +0, 001 No attack; no subsuarface phase 316 stainless steel 815 1000 +0.013 Thin film; no cracks on bending 90°; subsurface phase to depth of 1 mil 316 stainless steel 815 100 +0.014 Thin film cracked on bending 90°; specimen contains a subsurface phase to a depth of 1 mil 316 stainless steel 704 100 +0, 009 No attack; no evidence of cracking on 135° bend 316 stainless steel 650 100 +0, 006 No attack; cracked slightly on 90° bend 347 stainless steel 815 100 +0,016 No attack; no cracking on 135° bend 347 stainless steel 704 100 +0.394 Thin surface deposit; no attack but trace of sub- surface phase present; cracked on 90° bend 347 stainless steel | 650 100 +0,010 No attack; no cracking; trace of subsurface phase Inconel 815 100 +0, 004 No attack; cracked om 135° hend Inconel 704 100 +0.003 No attack: broke om 45° bend Inconel 650 100 +0, 002 No attack; cracked om 135° bend STATIC CORROSION OF FUEL period. These tests included static CAPSULES IN SODIUM H. W. Savage and W. C. Tunnell ANP Division Static-corrosion tests on fuel- containing capsules to determine the combined corrosion effects of the fluoride fuel NaF-BeF,-UF, and sodium at 1500°F were completed during the runs on inconel fuel~contzining cap- sules in sodium of 600, 800, 900, and 1000 hr duration. In all, omne complete series of tests has been rum on inconel and on 316 stainless steel fuel cap- sules. Fach series specified that seven capsules be tested 1n sodium at 1500°F for periods of 100, 200, 400, 600, 800, 900, and 1000 hr. Each capsule contained a slug of the capsule 105 ANP PROJECT QUARTERLY PROGRESS REPORT material in addition to the fuel. The analysis of the capsule and fuel are reported in Tables 11.9 and 11.10 for the inconel and the 316 stainless steel capsule tests, respectively. In general, the effect of either corrosive fluid was no greater than that obtained in tests 1n which the other fluid 1is DYNAMIC CORROSION IN THERMAL CONVECTION LOOPS G. M, Adamson, Metallurgy Division During this period the major effort with thermal-convection loops has been divided into corrosion studies with not present. fluorides and hydroxides. Table 11.9 Analysis of Inconel Fuel Capsules in Sodium at 880°C Yarious - WT. CHANGES'¢) WT. CHANGES'?®’ RUN TIME OF CAPSULE OF SLUG ANALYSIS OF FUEL (ppm) (hr) (mg/dmzfday} (ug/dm?/day) Fe Ni Cr 0 (start) 0 0 125 1065 20 100 -14.3 +1.0 160 110 4180 200 -7.5 -6.9 120 66 2880 400 -5.8 ~-1.6 215 180 1880 800 -2.6 +0,7 400 870 2550 (°)Weight changes (b)Weight changes of slug are due to fuel corrosion. Table 11.10 of capsule are due to sodium corrosion frem the outside. Anzlysis of 316 Stainless Steel Fuel Capsules in Sodium at 800°C WT. CHANGE(#’ WT. CHANGE'?®’ RUN TIME OF CAPSULE OF SLUG ANALYSIS OF FUEL (ppm) (hr) (mg/dm?/day) (mg/dn?/day) Fe Ni Cr 0 (start) 0 0 125 1065 20 200 -4,32 -11.15 85 <15 3650 400 -8.78 -5,88 1750 300 6640 600 -2.05 -4,44 90 <20 630 626 -1.5 -5.97 110 <20 1820 900 -2.21 -0,848 140 <20 2310 1000 -4.97 -1.58 150 35 1110 (a)Wéight changes of capsule are due to sodium corrosion from outside. (6)Wéight changes of slug are due to fuel corrosion. 106 FOR PERIOD ENDING DECEMBER 10, 1951 ‘hydroxides have been run in nickel and in every case excessive ‘mass transfer has taken place. Fluoride :salt mixtures have been successfully ‘circulated in . inconel but have plugged loops, in stainless steel. Fluoride Corrosion. to be the most desirable material for circulating fluorides. loop has now been operating at 1500°F for 240 hr with the fluoride fuel and 29.2 wt % LiF to which enough UF, has ) Since no falling off is being found in the. -cold-leg temperatures, mass transfer. ~does not seem to be taking place. ‘Another inconel loop has been operating ‘with the above fluoride mixture with UF, for 570 hr, again without any sign The 316 stainless steel loop containing the same mixture The loop has been X-rayed and shows both light "and dark patches in the hot leg. Which ‘constituent caused the plugging and ‘the exact location of the plug could A 316 stainless steel loop was terminated after 173 hr' of operation with the uraniumless This loop developed ~a leak about the center of the vertical ‘portion of the hot leg. While the leak was responsible for the termination, ‘the cold-leg temperature had dropped 55°F, which indicates that some mass ‘transfer was also taking place. X rays of this loop also showed several dark areas in the hot leg. Whether these are segregation or gas pockets formed on ~cooling can be determined only when the | However, since the they are Metal- lographic examinations have not yet 11,7 we % NaF, 59.1 wt % KF, ‘been added to make 1.1 mole %. of mass transfer. plugged completely in 82 hr. ‘not be determined. “fluoride mixture. loop is sectioned. . cold leg does not show any, most likely to be segregation. "been made on these loops. Hydroxide Corrosion. ‘potassium, and lithium hydroxides were ‘circulated in nickel with a hot-leg On the basis of preliminary tests, inconel appears An inconel ‘been partially dissolved. Sodium, temperature of 1400°F. All these loops failed by plugging in the hot leg as a result of mass transfer. The operating times were: sodium hydrox- ide, 54 hr; potassiumw hydroxide, 51 hr; and lithium hydroxide, 317 hr. The lithium hydroxide loop has not yet been completely examined but appears to be like the other two which are almost identical, Photomicrographs of both the hot and cold leg of the KOH loop are shown in Fig. 11.15. Although the surface of the hot leg appears to be well polished, examination under the micro- scope shows that this is a false surface which has been pulled away from the original material. Under the micro- scope, the cold leg surface appears guite rough with a layer of equi-axis crystals on the surface. Held in the rough spots on this surface is a con- siderable quantity of a nonmetallic constituent. Dendritic crystals were found lightly attached to the wall in all sections of the loop except the hottest part of the hot leg. The walls in all sections of the loop have ‘ In the sodium hydroxide loop about 0,002 in. had been dissolved from the cold leg and 0,007 in. from the hot leg. The figures are even higher with the potassium hydroxide loop. From X rays and dissolving out the hydroxides, which leaves the dendritic masses in place, the plugs in all loops have been found to be located in the horizontal section and lower part of the vertical section of the hot leg. Although dendrites were found attached to the wall in the cold leg, no large crystal masses were found in this section., Figure 11.16 is two views of the plug in the horizontal section of the sodium hydroxide loop. The top view is as cut while in the bottom the hydroxide has been washed out. The metallic masses are high-purity nickel. From X rays of the plug the 107 ANP PROJECT QUARTERLY PROGRESS REPORT B UNCLASSIFIED T.522 (b) Fig. 11.15. Nickel Thermal-Convection Loop Walls with Potassium Hydroxide. (a) Hot Leg. (b) Cold Leg. 250X 108 FOR PERIOD ENDING DECEMBER 10, 1951 UNCLASSIFIED Y-4884 UNCLASSIFIED Y-4883 (b) Fig. 11.16. Plugged Section Nickel Thermal Convection Loop with Sodium Hydroxide Coolant. (a) As Cut. (b) Metallic Constituent. 109 ANP PROJECT QUARTERLY PROGRESS REPORT crystals appear to grow from the bottom of the pipe and slant into the coolant stream. No intergranular type of attack was found i1n any portion of these loops. Inconel loops were terminated after 1000 hr with sodium at 1600°F and with NaK at 1500°F. Neither loop developed serious amounts of mass transfer or corrosion. A small amount of pitting was found in the hot area, but no intergranular type of attack was found. FUNDAMENTAL CORROSION RESEARCH W. D. Manly Metallurgy Division W. R. Grimes Materials Chemistry Division Reported failure of the empirical approach to furnish a satisfactory container material for molten hydrox- ides has emphasized the need for fundamental understanding of the mechanism of corrosion and mass trans- fer of metals by these liquids. At present 1t 1s not possible to state with certainty the nature of the chemi- cal reactions taking place; attempts to minimize or control the corrosive action are difficult to plan system- atically. Efforts to understand the nature of the reacting species are, at present, directed along two parallel lines. Electrochemical studies of two different types are attempting to define the ionic species involved in corrosion and mass transfer and are intended as useful tools for study of these com- plex phenomena. In addition, a careful study of mass transfer, as well as other aspects of corrosion such as diffusion of various atoms from liquid phase into the solid phase, and the effects of various atmospheres and other environmental factors on corro- sion has been undertaken. One of the 110 primary aims of this work will be concerned with the development of a suitable inhibitor for corrosion in the chemical sense and for inhibition of metal crystal (mass transfer) formation. These programs have been initiated only recently, and results to date are tentative and incomplete. Sufficient progress has been made, however, to indicate that useful data will result from these approaches. EMF Measurements in Hydroxides (A, R. Nichols, Jr., Materials Chemistry Division). Apparatus and materials are being assembled for the measurement of electrode potentials i1n fused hydroxides. The apparatus will permit measurements involving a range of temperatures (325 to 1000°C) of solvents (fused hydroxides), of solutes (NiO, Cr,0,, FeO, etc.), of electrode materials (Ni, Fe, Cr, Ag, Pt, C, alloys), and of atmospheres (H,, He, vacuum). A nonmetallic cell container and diaphragm will be used over as much of the temperature range as possible. An attempt will be made to develop reversible electrodes which correspond to reactions of the type involved 1in the corrosion and mass transport phenomena. If such electrodes can be devised, it will be possible to deter- mine free energies and temperature dependencies. This may provide an understanding of the mechanisms of these processes and hence point the direction for their prevention. Measurements now in progress involve the concentration cell Ni, NiO (dissolved in NaOH at c,) || NiO (in NaOH at ¢,), Ni to determine whether the Ni-dissolved- NiO half-cell is reversible. Polarographic Study of Sodium Hydroxide (R. A. Bolomey, Materials . FOR PERIOD ENDING DECEMBER 10, 1951 Chemistry Division). Techniques for the study of polarographic curves to _indicate the presence of contaminants ~in fused caustic are under investi- gation. Because of the high conductivity "of molten NaOH, a rather elaborate _compensating circuit 1s required in ~order to distinguish the polarographic waves. As presently operated, using ~a Brown recorder to trace the curves, - the position of the waves is indicated - by sharp peaks 1in an otherWISeascendlng " current-voltage curve. Sufficiefit reproducibility has not not yet been achieved, but the results ~are encouraging. JIrials with a - stationary platinum electrode showed - two peaks with NaOH contained in = platinum crucible under a nitrogen ~atmosphere. These peaks, at -1 and -1.4 volts, may have been attributable "to platinite and platinate ionmns. . The instrument was modified and it was ~ found advisable to operate the polaro- - graphic cell in a vacuum te avoid ~erratic behavior apparently due teo convection currents caused by surface cooling in the presence of a gas. Hydrogen was an ountstanding offender in this connection. It will be of ~interest to modify the apparatus so that gaseous products from the mixture ~can be analyzed. So far it seems ~improbable that the peaks can be assoclated thh Na* or NaH. Experiments with NaOH contained in ~silver.give peaks occurring at different ~voltages than those encountered in " platinum; however, the experiment is ~ being repeated under improved condi- tions. The solidified NaOH contained a mat of fine metallic needles which appeared to have been freely suspended in the upper part of the melt. These are pictured in Fig. 11.17 but have not yet been identified. Snrveytlfthe Mzss-Transfer Pheneme; non {W. D. Manly, Metallurgy Division; F. F. Blankenship, and R. P. Metecalf, Materials Chemistry Division). The phenomenon of mass transfer of nickel in systems comntaining caustic has been reported from a number of 1nstalla- tions which have experimented with these materials under a variety of conditions. The experiments which have definitely shown this phenomenon are descrlbed briefly helow. : Summary of Experimental Observe- tions. Experimental Engineering thermal-convection loops have shown adherent wall deposits in colder regions and Spongy dendritic plugs in hotter regions in three experlments using loops of Y%-in. nickel pipe approximately 90 in. in circumference. Velocities of about 25 ft/min and a temperature differential of 760 to 680°C prevailed in these experiments. NACA. Temperature gradient loops with centrifugal circulation utilizigng a temperature differential of 815 to 800°C with velocities of 15 ft/sec gave massive wall deposits in cooler regioas. Mass transfer was accelerated by the addition of sodium metal. ' BMI. Corrosion tests 1in a nick§1 pot having a thermocouple well, gas inlet tube, and specimen support immersed in the melt showed heavy deposition of nickel at the liguid level around metal walls and at culder portions. NACA. Corrosion tests 1in sealed crucibles with a temperature differ- ential of 815 to 730°C from top to bottom gave rise to a crystalllne deposit in the colder portion. X-10. Asingle low-velocity thermal-~ convection loop locaded under hydrogen utilizing a temperature differentizal of 700 to 425°C developed a scattered deposit on the cold wall with small evidence of dendritic plugging in in 300 hr of operat1on. , 111 ANP PROJECT QUARTERLY PROGRESS REPORT Fig. 11.17. Containing Sodium Hydroxide. Y-12. Corrosion tests in sealed tubes with temperature cycling showed small amounts of deposited metal after 12 cycles of 8 hr each from 800 to 200°C using helium gas inside sealed tubes. Instances in which directly observ- able deposits of metal were not apparent were as follows: Y-12. Isothermal corrosion tests (25 experiments) in capsules sealed under helium and heated in a vacuum at 800°C for 100 hr showed erratic weight losses but no significant mass transfer. X-10. Temperature-gradient corro- sion tests with hydrogen covering the melt and surrounding a vertical tube 17 in. long, 400°C at top, 800°C at middle, and 600°C at bottom, gave no 112 Deposit of Silver Crystals by Mass Transfer UNCLASSIFIED T«511 _ in Silver Capsule evidence of mass transfer and very little oxidation product in 100 hr. Chemical Examination of Plugged Convection Loops. In view of the failure of mass transfer to develop in certain cases 1n which 1t was expected, particular attention has been paid to the contents of plugged loops obtained from the Experimental Engineering experiments mentioned above and described in greater detail elsewhere in this report. A mapping of the extent and general appearance of deposited metal was carried out. No correlations were achieved, largely because of the lack of uniformity of the deposit on colder walls. There rough sparse deposits were blotched with dull surfaces and with regions containing fine dense crystals. Occasionally the line of FOR PERIOD ENDING DECEMBER 10, demarcation between sharp, bright needlelike crystals and a dull surface was surprisingly sharp. Such lines occurred in both the upper and lower arms, running longitudinally along the tube as though tracing a liquid level., Uniformly, the hotter sections beyond the dendritic plugs were highly polished. Surfaces in the immediate vicinity of plugs were visibly pitted but bright. Hydroxide from various portions of the loops was submitted for analysis with the results shown in Table 11.11. A careful study of the oxidation products from sections of the KOH loop is still in progress. On being leached with water, the hotter portions gave rise to a gelatinous black precipitate which dried to a brownish powder. Colder regions yielded brown 1951 powder, and the trap contained a mixed green and black powder. Tiny metallic crystals were also found in each case. X-ray-diffraction methods showed only NiO and nickel in varying proportions in the residues obtained from nine different regions in the KOH loop. An extensive black deposit from the cold leg of the LiOH appeared to be NiO from the standpoint of X-ray diffraction, but the chemical analysis showed only 49% Ni. There was much less discoloration of the surface of the hydroxide as found in the filling chamber in the case of LiOH, Spectrographic examination of the pipe of which the nickel loop (con- taining NaOH) was constructed showed Table 11.11 Analysis of Hydroxide From Nickel Thermal Convection Loops After Plugeging LiOH NaQH KOH Ni NiO Ni NiO Ni10O2 Ni NiO REGION (%) (%) (%) (%) (%) (%) (%) Hot leg 1.02 1.35 3,79(a) | 0,0 0.01 } 23.3%) | 0,01¢(%) Cold leg 1.31 15,402 | 0,06 0.01 | 0.06 Top arm 0.08 1.58 400 ppm | 0,01 Bottom arm 9.45 0.01 24.4 0.01 Darkest crust from fill chamber 0.3 1.2 6,07 0.83 0.20 1,92 20.2 Clean hydroxide from fi1ll chamber | 1 ppm 100 ppm |Total Ni (“)Sample contained some metallic dendritic sponge. (b)x-ray-diffraction analysis of this sample showed 40 to 70% NaOH, 20 to 40% Ni, and (probably present) 5 to 15% NiO. (C)A lar case of LiOH was unique. amount of oxide adhered to the wall near the bottom of the cold leg. In this respect the 113 ANP PROJECT QUARTERLY PROGRESS REPORT 1.3% Fe, 0.6% Co, 0.8% Cu, and 0.04% Mn. The dendritic sponge contained 0.04% Fe, 0.3% Co, 0.8% Cu, and 0.01% Mn, indicating that cobalt and copper tend to be transported to a greater extent than iron and manganese. Conclusions. Regardless of differ- ences in proposed mechanisms, the oxidation of nickel 1s an essential part of all plausible explanations of mass transport., There 1s reason to believe that, even with large tempera- ture gradients, mass transfer would 114 not be a problem in nickel—-molten hydroxide systems operated under conditions such that the oxidation of nickel does not occur. A small step toward this end has been the attempted development of nickel vessels suitable for carrying out the final stages of hydroxide purification, and of such design that the molten hydroxide could be trans- ferred to experimental apparatus without exposure to air or other sources of contamination. FOR PERIOD ENDING DECEMBER 10, 1951 12. HEAT-TRANSFER RESEARCH AND PHYSICAL PROPERTIES H. F. Poppendiek, Reactor Experimental Engineering Division Experimental temperature measure- ments in a simulated liquid- fuel element, using a brine solution as the hcat-generation medium, indicated significant reduction of fuel-element center temperatures at high heat fluxes as a result of free convection, These results are in general agreement with the theoretical work which has been carried on. The fused-salt heat-transfer apparatus has been completed, and test-section heat-loss calibrations are currently being made. Upon com- pletion of these calibrations heat- transfer coefficient measurements for the NaF-KF-LiF eutectic are to be obtained. Progress has been made in ligquid-metal heat-transfer work on lithium and sodium systems. Analysis of data on entrance-region heat transfer in sodium is presented, along with mathematical solutionms. Mathe- matical solutions have also been derived for forced-convection heat transfer (laminar and turbulent flow) in long smooth pipes containing fluids with uniform-volume heat sources. Data are being obtained on heat capacity, density, and vapor pressure of materials., Some data have been obtained on viscosity and thermal conductivity, but further measurements awailt the completion and testing of additional equipment, Heat-capacity data have been determined for uranium fluoride, a lead-bismuth alloy, a fluoride-fuel mixture, nickel, and sodium hydroxide. Measurements on thermal conductivity of diatomaceocus earth have been obtained by use of a radial-flow apparatus, while other equipment for measuring this property in ligquids and in solids 1s nearaing completion, Data on density and vapor pressure of fluoride salt mixtures have been obtained, from which coefficients of expansion and heat of vaporization have been calcu- lated. Some values for the viscosity of fluoride salt wixtures have been obtained, while other viscosity apparatus remains under construction. NATURAL CONVECTION 1IN LIGQUIB-FUEL ELEMENYS F. E. Lynch, ANP Division D. C. Hamilton, R. F. Redmond, and M. Tobias, Reactor Experimental Engineering Division Preliminary data have been obtained on a new simulated fuel-element system, described in the previous guarterly report.{!) A 25% solution of NaCl in water was used as the heat-generation medium, and cooling was effected by natural convection from the wall of a 3-mm quartz tube to a stirred bath. The results plotted in Fig. 12.1 are for six series of tests at coolant temperatures of 8, 12, 17, 29, 32, and 66.5°F, respectively. The ordinate 9/9c 1s the ratio of the measured temperature difference between the wall and the axis of the tube to the temperature difference computed from the conduction equation for the systen. The data are in general agreement with analytical (laminar flow) so- lutions for values of the scource term that are less than 110 watis/cm?; at this point the ordinate decreases sharply. This decrease is attributed (D E. Lynch and M. Tobias, ‘““Measurement of the Fuel-Element Temperature Distribution,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending Septesmber 10, 1951, ORNL-1154, p. 133 (Dec. 17, 1951). 115 ANP PROJECT QUARTERLY PROGRESS REPORT UNCLASSIFIED DWG. 13593 Sttt T 17 T 17 1 1 1.0 — © o —— — — o 0.8 — ] o 0.6 | — ~ | § =0 CENTER ~ @ WALL ] ® 8.=8 FOR CONDUCTION ONLY 0.4 |— — 0.2 }— ~ o ey e 0 20 40 80 80 100 120 140 160 180 q'"' (WATTS PER CUBIC CENTIMETER) Fig. 12,1, Temperature Ratio inp a 3-mm Tube Filled with Brine in which Heat is Geperated Uniformly. to a change in the flow from the laminar to the turbulent regime. _ A coolant jacket is being added to the quartz-tube system so that the cooling will be accomplished by forced flow parallel to the quartz tube. The flow rate and the axial temperature gradient of the coolant will be measured. Four analytical solutions which relate radial temperature distributions to several dimensionless moduli, including such variables as pipe diameter, volume heat source, physical properties of the fuel and coolant, and coolant fluid flow rates for free-convection systems, have been obtained for the laminar flow region.(® (2)Three memorandums on this are to be issued around January, 1952. 116 Similar solutions for the turbulent- flow regime are being sought. A new parallel-plate apparatus in which the velocity distribution will be measured is being constructed which will employ an electrolyte as the heat-generating mediuvm. HEAT-TRANSFER COEFFICIENTS Heat Transfer in Fused Hydroxides and Salts (H. W. Hoffman and J. Lones, Reactor Experimental Engineering Division). The apparatus for the determination of the heat-transfer coefficients of molten salts and hydroxides has been completed. Heat loss calibrations of the test section are being carried out. Heat-transfer measurements using a fluoride mixture will be made upon completion of these calibrations, ™ FOR PERIOD ENDING DECEMBER 18, 1951 Since the last report the experi- mental system has been fully instru- mented, and all tanks and flow lines have been cleaned with 2 nitric acid solution and filled with approximately 200 1b of the NaF-EF-LiF cutectic. During the filling operation several cold spots were noted i1n the system, and the auxiliary heating circuit was altered to eliminate these regions. The system has been further modified to include mixing pots of a helical- coil design immediately preceding and following the test sectiom in order to determine more accurately the entering and leaving mixed-mean temperature of the test fluad. Heat Transfer in Moltem Lithiun (H. C. Claiborne and G. M. Winn, Heactor FExperimental Engineering Division)., It was previously re- ported¢?’ that the modified figure- e1ght system for the determination of lithium heat-transfer coefficients had been successfully operated, and that preliminary heat-transfer data had been obtained. To facilitate accurate separation of the individual thermal resistances from the experi- mental overall heat transfer coef- ficients, a large pump (approximately 40 gpm at 100 psi) 1is required. Since a pump of the required capacity appeared to be unavailable 1n the next few months, 1t was decided to hold the figure-eight system in stand- by status until the regquired capacity pump is obtained and to build a smaller system utilizing a heat-transfer section that is resistance-heated by an electrical current. The latter method obviates fluid thermal-resistance separation techniques and has the additional advantage that it allows the measurement of the actual heat flux distribution along the tube length, (3)0. P. Coughlen 2nd [[. C. Claiborne, ““Heat Tra?§fer in Molten Lithium,’ OBRNL-1154, op. cit., p- 9. Construction of the new system is about 95% complete. Data are expected to be obtained by Janunary, 1952, for Reynolds meoduli up to 40, 000. The design gqualifications for the test section are: 1. About 95% of the heatis generated in the tube wall, 2. The average temperature differ- ence betwsen the wall and the fluid mean temperature is at least 10°F. 3. The axial mean temperature rise is at least 10°F, 4., At least 75% of the thermal resistance 1s 1n the lithium. 5. There is a maximum of 3000 amp to the test section. 6. The length-to-diameter ratio of the test section is 100. Qualifications 1 and 2 require a metal having high thermal and electrical conductivities. Copper was selected as the material for the test section, A 10-hr static corrosion test(*) at 450°F indicated that a polished copper specimen lost only 0.02% in weight. Therefore it appears that copper will resist attack by lithium long encugh (about 20r 3 hr) to take the necessary data if the maximum operating tempera- ture does not exceed 450°F. A flow diagram of the system 1is shown in Fig. 12.2. The dimensions of the test section are: length, 11¥% in.; 1.d.,, 0.1175 1in.; and o.d., 0.1715 in. Flow is produced by two electromagnetic pumps 1in series (previous operational data indicate that this arrangement will give 1.7 gpm at 40 psi). Flow measurements will be made with an (4)1,. A. Abrams, personal communication to W. D. Manly, Oct. 5, 1951. 117 ANP PROJECT QUARTERLY PROGRESS REPORT —a— GAS INLET UNCLASSIFIED DWG. 13594 MICROMETULLIC a-MELT AND FILTER TANK COPPER COOLING COILS FILTER Ei 1 COOLING WATER E{ E{ GAS !NLET7 = Y E( l BUS BAR "L" —[ T -= BY PASS MIXING CUP—" il SURGE TANK-» T TEST SECTION —= EXPANSION JOINT~\H‘1::i;:i ELECTROMAGNETIC EZ }i FLOW METER T - ELECTROMAGNETIC PUMPS + I g [:] ~~—GAS INLET DRAIN AND CALIBRATION TANK —= Fig. 12.2. Flow Diagram for Lithium Heat-Transfer Experiment. electromagnetic flowmeter that will be calibrated in place. The bypass is used for testing the circuit and for obtaining approximate equilibrium. This prevents unnecessary exposure of the copper test section to lithium. Before data on the test section are obtained, the bypass will be frozen and disconnected to prevent short- circuiting of the heating current to the test section. The tube can be easily replaced in case of failure. It 1is felt that the described system will produce satisfactory 118 lithium heat-transfer data and, with modification of the test section, will allow the heat-transfer charac- teristics of other liquid metals to be determined in the same apparatus. Entrance-Region Heat Transfer in a Sodium System (W. B. Harrison, Reactor Experimental Engineering Division). The system built for obtaining entrance- region heat-transfer coefficients to molten sodium has been described in previous guarterly reports (see p. 1v for list of previous reports). Briefly, the test section consists ofa circular FOR PERIOD ENDING DECEMBER 10, 1951 copper plate having a hole in the center anda tube around the periphery. Sodium flows through the center hole, and a cooling or heating medium is circulated through the tube at the periphery., The plate is insulated to promote radial conduction, and the temperatures at different vadial positions were measured in order to determine the copper surface tempera- ture at the copper-sodium interface and the rate of heat flow through the plate. A total of 48 experimental runs have been made to date. Plates of 1/16 and 1/8 in. thickness have been used 1n conjunction with hole diameters of 1/16, 5/64, and 3/32 in. Values of P, D/L ranged from 140 to 585, where P_is the Peclet modulus, D is hole diameter, and L 1s the plate thick- ness, All data were taken with the sodium in turbulent flow at tempera- tures between 250 and 310°F. The data have been extremely erratic and very low when compared with the predictions based on the postulates of slug flow (uniform velocity distri- bution} with heat transfer by molecular conduction only. In a number of cases the data were low even when compared with predictions based on the postulate of laminar flow (parabolic velocity distribution). Following are several possible explanations for the low data: 1. Peculiar velocity distribution, This might be caused by creep of the gasket upstream from the plate. It was found that some creep actually does take place. However, the effect was practi- cally eliminated by giving the gaskets a preliminary compression set at the operating temperature. 2. Nonwetting of the copper by the sodium. There are few data available on wetting of surfaces 4. by sodium, but it appears that a clean surface is wet by sodium without difficulty, even at the low operating temperature, The surface in the test section was degreased with a detergent and electrochemically polished. No significant changes resulted in the data. Precipitation of oxide on the heat-exchange surface. This was considered to be a possibility during the runs in which the sodium was being cooled. The apparatus was equipped for heat- ing the sodium without signifi- cant improvement of the data. Contamination of the sodium such that its properties have been altered; particularly, the thermal conductivity has been reduced, The chief contaminant is the oxide, with some iron particles also in suspension, This appears to be the wmost likely cause for the erratic and low data. The system is presently being cleaned and modified so as to minimize contamination. There has been a fairly consistent decrease in the heat-transfer coefficients with time, implying that the condition is being aggravated by a buildup of contaminant., It is believed that particles of oxide and 1iron have been suspended in the stream, such that the extent of contamination exceeds the equi- librium solubility of sodium oxide in sodium. The oxygen comes from three sources: (a) residual oxygen in the system prior to loading the sodium; (b) residual oxygen in the form of oxide on the surfaces of sodium bricks which were loaded into the system; and (c) oxygen present in the argoun used in the system., A filter is being in- ‘stalled in the sodium line so as 119 ANP PROJECT QUARTERLY PROGRESS REPORT to remove oxide 1in excess of the equilibrium solubility. In order to eliminate the oxide introduced during loading of bricks directly into the system, a separate melt tank will be used. From this tank the sodium will be trans- ferred into the system through a filter,. In order to reduce contamination from the gas, helium will be passed through a NaK scrubber befere it reaches the system. These changes in the system are being incorporated at the present time. Jt 1s hoped that the system may be started up again by the middle of December, Heat Transfer in a Circualating rFuel System (H. F. Poppendiek and .. Palmer, Reactor Experimental Engineering Division)., Mathematical solutions have been derived for temperature structure in the case of forced- convection heat transfer (laminar and turbulent flow) in long smooth pipes containing fluids with uniform-volume heat sources; heat i1s transferred to or from the fluids at the pipe wall. Some specific evaluations of the radial temperature distributions for the case of no heat transfer to or from the fluid at the pipe wall have been made for laminar and turbulent flow. Diwmensionless temperature profiles for the case of turbulent flow are presented in terms of Reynolds modulus, Prandtl modulus, the volume heat source, and the boundary hesat transfer. HEAT CAPACITY W. D. Powers G. C. Blalock R. M. Burnett, Reactor Experimental Fngineering Division The heat capacities of nickel, sodium hydroxide, uranium fluoride, a lead-bismuth alloy, and a fuel mix 120 have been determined by means of Bunsen ice calorimeters.(3* %) The equations in Table 12.1 give the heat capacity as a function of temperature; H, - Hyo, is in cal/g, ¢, is 1in cal/g*°C, and T 1is in degrees centi- grade. THERMAL CONDUCTIVITY A Deem type apparatus for the measurement of thermal conductivity of liguids has been completed and 1is currently being tested with lead. A longitudinal flow apparatus for measuring thermal conductivity of solids 1s mearing completion. Some measurements of thermal conductivity and sintering have been made on diatomaceous earth, a material antici- pated for use as insulation in the circulating-fuel —~water-moderated reactor. Thermal Conductivity of Ligquids (S. J. Claiborne and M. Tobias,Reactor Experimental Engineering Division),. A Deem type apparatus for the measure- ment of the thermal conductivity of liquids has been modified as indicated in an earlier report.{?’ The apparatus is now complete and 1s being tested using lead. Some difficulty has been encountered in keeping thermocouple wells leak-free, but this problem has been solved, at least for lead as the test liguid. Replacement of the stainless steel bellows with one of nickel-plated copper or brass is now contemplated, because the present bellows is very stiff and causes a certain amount of distortion in the apparatus. An additional apparatus 1s (5)A. R. Frithsen, “Physical Properties,” OFNL.-1154, op. cit., p. 134, (6)a. R. Frithsen, “Physical Properties,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending June 10, 1951, ANP-65, p. 159 (Sept. 13, 1951). (7). Tobias, A. R. Frithsen, and L. Basel, “Thermal Conductivity of Liquids,” ORNL-1154, op. cit., p. 135, 121 Nickel NaQk UF Pb-Bi alloy (55.3 mole % Be) Fuel mixture {46.5 mole % NaF, 26.0 mole % KF, 27.5 mole % UF4} Table 12.1 Heat Capacities of vVarious Substances 250 - 1000°C H, (solid) - Hyo, (solid) = 0.26 + 0.119T +1.53 x 10”572 340 - 1000°C H, (Jiquid) - Hyo, (solid) = 69 + 0.497T 300 - 930°C Hp (solid) - H o (solid) = &.0 + 0.073T + 3.1 x 107572 200 - 950°C HT (liquid) - Hyo. (solid) = 5.8 + 0.0345T 240 - 535°C H, (solid) - Hjo. (solid) = -0.3 + 0.15T 535 - 1000°C HT (liquid) - Hjo. (solid) = -13.4 + 0.237 H 1t 0.12 + 3.1 x 10°°T 0.49 ¢ 0.02 0.073 + 6.1 x 10°°T 0.034 £ 0.0015 0.15 + 0.01 0.23 £ 0.01 IS6T ‘0T HIGWADIA ONIGNT GOIUAS NMOJ ANP PROJECT QUARTERLY PROGRESS REPORT being build so that delays caused by failure of equipment parts may be reduced. Thermal Conductivity of Solids (W. D. Powers, Reactor Experimental Engineering Division). Construction of an additional logitudinal flow apparatus for the thermal conductivity of solids 1s 90% complete and will be checked with Armco iron shortly. This apparatus can be used in a vacuum or in an 1nert atmosphere and will permit more accurate measurements than the original(®) apparatus. Thermal Conductivity of Diatomaceous Silica Powder (D. F. Salmon, ANP Division). Research on the circu- lating-fuel—water-moderated reactor has led to a search for an insulating material for the fuel containers. Thais insulation would allow the water to be maintained at a considerably lower temperature than the fuel mixture, thereby simplifying many of the problems associated with high-tempera- ture high-pressure systems, Diato- maceous silica powder seemed feasible; hence tests were made to determine thermal conductivity and sintering effect at reactor temperatures under atmospheric and reduced pressures. The material used was a Johns-Manville product, Celite. The apparatus consisted of two annuli formed by concentric tubes around a standard 750-watt 115-v tubular heating element. The annulus next to the heater element contained diatomaceous silica while the outer annulus was a water passage, Experi- mental parameters included heater power, sheath temperature, density to which insulation was packed, water flow rate, and water inlet and outlet temperatures. (8)M. Tobias, “Thermal Conductivity of Solids,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending March 10, 1951, ANP-60, p. 243 (June 19, 1951). 122 The first series of tests was made at atmospherzc pressure. The heater element sheath temperature was main- tained at 1500°F for 75 hr at 1675°F for 100 hr, and at 1800°F for 100 hr. During this entire series the tempera- ture rise of the water never exceeded 10°F., Figure 12.3 shows the variation in thermal conductivity with tempera- ture both from the literature values(®) and from the experimental ones. In- spection at the conclusion of these tests showed that a thin layer of the powder next to the heater element had darkened considerably, probably as a result of oxidation of the inconel sheath., Apparently no significant change in physical properties had occurred, since the thermal con- ductivity remained consistent through- out the tests. A second series of tests was begun in which the powder annulus was evacu- ated to determine the effect of the presence of air on thermal conductivity and sintering. The insulation annulus was loaded to a packing densaity of 15.7 1b/ft? and evacuated to approxi- mately 0.17 wm Hg. Other experimental conditions were unchanged. At a heater sheath temperature of 1800°F, an apparent conductivity of 0.025 Btu/hrft*°F was calculated, representing a reduction of some 60% over that at atmospheric pressure. After 24 hr of operation, however, the conductivity began to increase as evidenced by a decrease in sheath temperature, while the power on the heater element remained essentially constant; however, insufficient time had elapsed to establish the extent of this increase. Examination of the powder after testing indicated that a layer approximately 1/16 in. thick around the heater had sintered and darkened appreciably in color. ()G, B. Wilkes, Heat Insulation, Wiley, New York, 1950. 3y vh FOR PERIOD ENDING DECEMBER 190, 1951 UNGLASSIFIED D¥6G. 13595 " [ I ! T B | ! I | ! 0.140 |— _ CALCULATED DENSITY i o OF POWDER TESTED, 17.2 ib/ ft = AVERAGE WATER FLOW RATE Jx oazof— 135 1b/ hr — ~le° EGEN oo A- L 0- } VALUES FROM LITERATURE N © -- 0.100 - — > o~ EXPERIMENTAL VALUES AT ATMOSPHERIC PRESSURE . s - EXPERIMENTAL VALUES AT REDUCED PRESSURE 2o b/ }_ |5 ,..-v"‘/ L ] § 0080 flw&”““ 187Ib/fl 5 O THEORETICAL ,4r"’”’wfl ’flflflM{}mflfwf~;w_flflaw~‘ ., o % | m,:::::ji::::::d,,,flv~w~*”” 12.5 Ib/ ft3 i o I l_ 0.040 ONE POINT _ AFTER 12hr * e B S 0.020 L - - 0.000 | 1 I [ [ bt [ ‘ , | 0 j00 200 300 400 500 60O 700 800 900 1000 MEAN TEMPERATURE (°F) Fig. 1Z.3. No reason has been advanced to explain why diatomaceous silica sintered at reduced pressure but failed to sinter at atmospheric pressure. Further tests at reduced pressure are underway. DENSITY OF LIQUIDS J. M. Cisar, ANP Division Density measurements have been made on three more fused salt mixtures. Compositions, melting points, and temperature ranges over which data were taken are given in Table 12.2. The equations representing the data within 5% are indicated, where p = density and T = temperature (°C), The data on the NaF-KF-LiF-UF, mixture, although consistent in them- Thermal Conductivity of Diatomaceous Earth. as the salt was accidentally allowed toremain exposed to air for a 3-hr period before the runs were made. A sample 1s being analyzed to determine the selves, may not bhe reliable, effect of hydrolysis during thais period, The apparatus used for these determinations 1s based on ligquid buoyancy principles; it was described in a previous quarterly report, (1% In addition, a second density apparatus of this type has been built and is now in operation, VISCOSITY The falling-ball viscometer has been modified and tested with satis- factory results. Approximate results (1035 J. Kaplan, “Density of Liquids,” ANP-60, op. cit,,p. 246. 123 ANP PROJECT QUARTERLY PROGRESS REPORT Table 12.2 Data on Several Fused Salt Mixtures COMPOSITION MELTING POINT TEMPERATURE RANGE COMPOUND | (mole %) °c) (°c) EQUATION NaF 48.92 558 625 - 890 p = 4.54 - 0.0011T KF 26.8 UF, 25.0 NaF 10.9 430 - 440 525 - 850 p = 2.647 - 0.00090T KF 43.5 LiF 44.5 UF, 1.1 NaF 11.5 455 550 - 850 p = 2.385 - 0.00059T. KF 42.0 LiF 46.5 are expected with the use of a Zahn type viscometer now hearing completion, Some values for fluoride salt mixtures have been obtained using the modified Brookfield Synchro-Lectric viscometer (Fig., 12.4). Falling-Ball Viscometer (S. I. Kaplan and T. N. Jones, Reactor Experimental Engineering Division), The falling-ball viscometer has been tested at room temperatures, using SAE 10 lubricating oil as the test fluid. The viscosity of the o0il was checked with a Brookfield viscometer and found to agree with the result of the falling-ball instrument to within about 10%. After high-temperature testing 1is complete, the instrument will be used for investigations of molten salt mixtures, Recent modifications of the falling- ball viscometer include removal of the solenoid valve at the bottom and substitution of an externally heated straight pipe extension and instal- 124 lation of a removal cap to permit cleaning. Zahn Type Viscometer (M. Tobias, Reactor Experimental Engineering Division). A Zahn type viscometer has been designed to give rapid, approxi- mate values for fuel and coolant-salt mixtures. The viscometer will consist of a stainless steel cup, at the base of which 1s a hole 1 in. long and 1 mm in diameter. The cup will be dipped into the melt whose viscosity 1s desired. The melt container will then be dropped away from the cup and the time of efflux of 5 cm?® of molten material from the cup will be measured. Since the apparatus will be inside a furnace, the discharge time will be obtained by weighing the cup erther continuously with a dial-indicating balance or by noting when the weight of the cup plus its contents falls below a predetermined amount., To minimize corrosion, the apparatus will be operated inan inert gas atmosphere. The device will be calibrated using c K FOR PERIOD ENDING DECEMBER 10, 1951 RESTRICTED PHOTO 1048 VISCOME TER MANIPULATION PORT oo ] -INERT GAS QUTLET THERMOCOUPLE ORLE+ e FILLING FUNNEL ; GUARD PLATE FURNACE AND | CONTAINER — 1 - L ' FURNACE LEADS INERT GASFNLETSV?’ Fig. 12.4. Viscosity Apparatus. 125 ANP PROJECT QUARTERLY PROGRESS REPORT materials of known viscosity, both at room temperature and at high tempera- tures, The equation which an apparatus of this type obeys 1s of the form v = At - B/t where v is the kinetic viscosity of the liquid, A and B are constants characteristic of the apparatus, and t is the time of efflux, Viscosity of Fluoride Mixtures (F. A. Knox and F. Kertesz, Materials Chemistry Division). Preliminary viscosity determinations of the various fluoride mixtures were con- tinued. Although trials were made with a Saybolt type apparatus, 1t was found desirable to continue the use of a modified Brookfield Synchro-Lectric viscometer. In order to eliminate at least some of the disadvantages of the earlier techniques, the furnace was enclosed in a metallic shell, covered with a gas-tight cover which contained the measuring instrument, a glass- covered viewing hole, and an opening for a rubber glove to permit certain manipulations while the experiment was in progress. The enclosed unit was under slight argon pressure. In this way oxidation of the material during measurements should be reduced very considerably, Details of the apparatus can be seen in Fig. 12.4. The need for such an essentially gas-tight apparatus was apparent from the results obtained with beryllium- bearing fluoride mixtures. In a poorly sealed system the mixture be- came cloudy while the measurements were 1n progress, indicating the formation of insoluble beryllium oxide. The viscosity values at a given temper- ature continued to increase as repeated measurements were made, resembling the behavior of a suspension with rheopectic 126 properties rather than a Newtonian liguid. In order to establish the effect of the uranium concentration on the viscosity, the viscosity was determined for fluoride salt mixtures contalning various amounts of uranium fluoride. The latest values for the viscosity of the NaF-KF-LiF ternary eutectic range around 3 centipoises at 800°C. Results with 2 and 15 wt % UF, in the eutectic show that these mixtures have es- sentially the same viscosity as the uranium-free composition., VAPOR PRESSURE OF LIQUID FUELS R. E. Moore, Materials Chemistry Division The apparatus and procedure employed in this work have been described in a previous report,(!?) The cylindrical vessel of 316 stainless steel con- taining the NaF-KF-UF, eutectic mixture was heated in a monel block 1na pot furnace wound with Kanthal A wire, Temperatures were measured by chromel- alumel thermocouples attached to the outer wall of the vessel. A hand- controlled Powerstat was used to regulate the temperature of the furnace. A preliminary test of the apparatus and procedure was made in which vapor pressures of mercury were determined at several temperatures using a ma- nometer containing Hyvac oil feor pressure measurement, The results showed this method was satisfactory; hence measurements with fused salts were 1nitiated, The results of the vapor pressure determination of the NaF-KF-UF, (11)p E. Moore and C. J. Barton, *'Vapor Pressure,;’ ORANL-1154, op. cit., p. 136. @y e FOR PERIOD ENDING DECEMBER 10, 1951 eutectic mixture are given in Table 12.3. The values at 1199 and 1267°C were obtained with a meréury manocmeter and are not so accurate as the others, which were obtained with a wanometer containing Hyvac oil. These data are represented by the equation logln P__ He - -9500/T + 7.234. The heat of vaporization as calcu- lated from the equation is 43.5 kecal/mole. Table 12.3 VaporPressure(fl?theNaF°KF~vE¢Eutectic TEMPERATURE PRESSURE, OBSERVED, (°c) {mm Hg) 1073 1.43 1131 2. 82 1138 3.34 1199 6.5 1242 9.02 1267 12.0 127 ANP PROJECT QUARTERLY PROGRESS REPORT 13. W. D. METALLURGY AND CERAMICS Manly, Metallurgy Division T. N. McVay, Consultant Solid-fuel-element fabrication has been studied in order that the effect of several variables on the metal- lurgical bond between the fuel-bearing core and the protective cladding and the distribution of the UO, in the metallic carrier in the core may be understood. Previous reports have outlined this effect by such variables as percentage hot reduction, per- centage UO, present, composition of cladding, particle size of the metallic carrier 1n the core, and degrees of cold-working of the flat plate after rolling. In this report the following variables have beemn investigated: effect of UO, particle size range, effect of rolling tempesraturs, and elimination of the capsule used during hot-rolling. Preliminary experiments have shown that fuel-plate lawminates may be spot- welded together. Fuel-tube bundles can be held together by spot-welding bundle straps to the individual tubes. A study of high-temperature brazing indicates Nicrobraz to be guite useful in joining two fuel-plate laminates without floating the U0, to the sur- face and for joining a fuel plate to a stainless steel sheet in a T joint. A new brazing alloy containing 60% Pd and 40% Ni has been found which has a favorable nuclear cross-sectionaswell as good high-temperature brazing properties. The creep and stress-rupture labo- ratory for the testing of materials 1in an inert gas atmosphere has been completed. Results are presented which show the effect of grain size on the creep strength of inconel and of heat treatment on the creep strength of Duranickel, 128 Installation of equipment for the ceramics laboratoryis still in process although the laboratory has been in partial operatiocn since early fall, The work to date has included ceramac coating of radiators and the hot pressing of alumina, SOLID-FUEL-ELEMENT FABRXCATION E. S. Bomar and J. H. Coobs Metallurgy Division Of the three basic approaches to preparing a solid-fuel element out- lined in earlier veports (see p. 1v for list), the e¢mz upen which primary emphasis has been placed to date makes use of picture frames and cladding plates, The objectives of these in- vestigations are twaofold: (1) to abtain a metallurgical bond between a fuel-bzaring core and protective cladding, and (2) to create a distri- bution of the solid fuel 1n 2 metallic carrier which will give acceptable heat-transfer conditions. The results of completed experiments indicate that flat or curved fuel plates can be fabricated with 300 series stainless steel cladding on a UD, ceramel core, Earlier reports listed the effects of the following variables: 1. Percent hot reduction. 2, Percent UQ, present. 3. Composition of cladding. 4 . Particle size of metallice carrier 1n core. 5. Coarse U0, particles. 6§, Cold-working of flat plates following hot-rolling. FOR PERIOD ENDING DECEMBER 19, 1951 To these we may now add: ' Effect of U0, Particle Size. Vari- 3 ~ ationindistribution of U0, with three 1. FEffect of 1o, particle size. ; different sizes of U0, particles is j § : shown 1n Fig. 13.1. The mesh grades . 2. Effect of rolling temperature. | employed supplied particles measuring . _ ‘ : from 250 to less than 10 4, the distri- 3. Elimination of capsule {origi- bution improving inversely as the size nally used duvring hot-rolling). of the parvicles. This range of particle _ ; sizes is of interest since it is Preparation of tubular sections is believed to cantain the minimum size more difficult and a satisfactery end which will suppress radiation damage product is not yet assured, f to a tolerable level. SECRET e AR ; T GO o TR CORE G 0% UQ, 44200 MESH ?Qo?.:- 3 e { B ‘,S’u:s:_ ( STAINLESS | [ofn STERL j Vi 8Y VOL. CORE | CORE 200 . 0T U, deums f MESH ?QG& -:5\. ,_‘ 7 - STaniEss {0228 Greny o |MESH BY VOL, CORE : i MESH - PO 302 1 L. STAINLESS <« 250 STEEL g BY VOL. fe) Fig., i3.1 a, b, c. Effect of U, Particle Size on UG, Distribution. 175X, 12¢ ANP PROJECT QUARTERLY PROGRESS REPORT ¥Ry [ ?'._' { A oy Q%fi w 4 W g ROLLING TEMP 1335 °C (b) Fig, 13.2 a, b. Eifect of Rolling Temperature ca Uo, Distribution. 175X, Effect of Rolling Temperature. capsule during hot-rolling.¢!*2) This Dependence of oxide distribution on step, however, proved to be one of the rolling temperature is shown in Fig. more time-consuming 1tems 1in the 13.2. The obvious benefits of the fabrication schedule, Production of higher temperature are somewhat offset fuel plates in large numbers would be by the deeper penetration of surface facilitated by working out a schedule ox1dation of the samples during rolling. omitting the capsulating step. . . . (])flm T Fi 1 PI 2 tal L. ype Fue ate,”” Metallurgy Division Effect of Eliminatiom of Capsule. Quarterly Progress Report for Period Ending The first sample fuel plates were January 31, 1951, OPFNL-987, p. 57 (June 7, 1951). prepared by enclosing a stainless steel (2)Hot-Rolled Clad Fuel Plate,” Hetallurgy "picture frame,”" fuel-bearing ceramel, and cladding plates i1n an evacuated 130 Division Quarterly Progress Report for Period €3gi?g April 30, 1951, ORNL-1033, p. 56 (Oct. 9, w“l FOR PERIOD ENDING DECEMBER 10, 1951 Two samples were processed without protective capsules with results whlch - were quite encourag1ng. The first laminate was heliarc- ~welded around its outer edges in a dry box containing a purified helium atmosphere. The second laminate was prepared by a combination sintering and hot-forging operation. A thin PICTURE / FRAME f PICTURE FRAME (b) Effect of Elimination of Capsule on Uflé Bistrihutién. Fig. 13{3 a, b. 150X, : band of stainless stéel powder was painted around the ocuter periphery of the picture frame using Nicrobra:z cement as a binder; cladding plates were spot welded on and the assembly sintered overnight at 1250°C. The sintered laminate was next hot-forged at 900 to 1000°C by pressing at 5 tsi. Photomicrographs of specimens taken from the samples after hot-rolling are shown in Fig. 13.3. 302 ’ : STAINLESS i SfiflfiET R -mmz 302 STA*NLESS STEEL - | U0, GCORE i3l ANP PROJECT QUARTERLY PROGRESS REPORT Preparation of Tubular Fuel Elements. Short segments of tubular geometry have (1) cold- into semi- been prepared in two ways: forming of flat plates circles and joining with heliarc welds, and (2) "rubberstatic" pressing of fuel bearing mixtures 1inside a seamless cladding tube. A series of 12 specimens was pre- pared, 11 by the first method and one by the second, cold-drawn at Superior Tube Company, and examined at QOak Ridge National laboratory. Three metal powders were used as carriers for the UO, in the core — 302 stainless steel, nickel, and iron, however, the clad- type 316 stainless steel. Metallographic examination revealed in every specimen a tendency toward a In every instance, ding was stringer type of UO, distribution, varying from moderate in the 1iron cores to severe 1n the 302 cores. Cladding-to-core bonding was 1in most instances satisfactory after the cold- drawing. Further work will have to be done to determine conditions of draw- ing under which more acceptable uo, distribution results. Twc of these seamless-tube fuel elements formed by "rubberstatic™ pressing at 40 tsi with 302 powder are shown in Fig. 13.4. WELDING TECHNIQUES P. Patriarca and G. M. Slaughter Metallurgy Division Cone Arc(3) welding. Cone-arc apparatus which incorporates a fixed magnet nozzle on an inert-arc welding torch has been used with moderate success to fabricate typical inconel tube-to-header assemblies. The inconel (D, R. Mann, Means for Meking Uniform Circular Heliarc Welds by Deflecting the Ion Beam Continucusly, ANP-63 (Apr. 9, 1951). 132 tubes were 0.188 1in. o.d. with a 0.025-in. wall thickness. Headers were stamped from 0.0625-in.-thick inconel sheet, Although the feasibility of tube- to-header welded construction was demonstrated by these experiments, results were somewhat inconsistent. Apparatus has been completed which incorporates a Selsyn stator as part of the magnetic inert-arc welding torch nozzle as described in a previous report.¢?? Jt is felt that the proper control of the additional welding variables introduced by this method will improve the consistency of results. An investigation of the effect of these parameters on the type and quality of welded tube-to-header joints will be conducted concurrently with fabrication of full-sized tube- to-header assemblies for testing by the Experimental Engineering Group. Resistance Welding. Experiments have been conducted using available equipment which has demonstrated the feasibilaty of joining tube bundle straps to fuel tubes by spot welding. A typical spot weld 1s presented in Fig. 13.5 and shows a 0.016-in.-thick inconel strap spot-welded toan 1inconel tube 0.188-1n. o.d., 0.025 in. wall thickness, It may be seen that penetration and soundness of the spot weld 1s excellent, A preliminary investigation has been conducted to determine the feasibility of spot-welding clad fuel elements together. The fuel elements consisted of a mixture of UD, and 302 stainless steel powder. A sound spot weld could be made which apparently bonded the two sheets together without any gross macroscopic movements of the powder layers. Fig. 13.6. This 1s 1llustrated in FOR PERIOD ENDING DECEMBER 10, 1951 ‘, ~160 30% U0, +200 i - MESH STAINLESS 4 7220 0% 302 { STEEL PRESSED AT 40781 SINTERED 12509¢ 316 TUBE - =100 30% U0, 3§ +200 . | MESH 70% 302 :_f 135 STAINLESS { ~22 STEEL | MESH PRESSED AT A0TS! | SINTERED 1310°C 316 TUSE {B) Fig. 13.4. Seamless-Tube Fuel Eléments Formed by "Rubberst:atic'" Pressing. {a) 2530x. (b) 175X. ~ 133 ANP PROJECT QUARTERLY PROGRESS REPORT Fig. 13.35. Transverse Section of a 9.815-in.«~Thick Inconel Sheetl Spot Welded to an 0. 188-in.-0.d4. Inconel Tube of $.025-in. Wall Thickness. FEtchant, aqua regia and glycerin. 175X, SECRET 8 Fig. 13.6. Secticnef a Spot Weld Joining Two Stainless Steel—=—Clad Fuel Plates. Etchant, agqua regia and glycerin. 60X, 134 FOR PERIOD ENDING DECEMBER 10, 1951 A 20-kva combination spot and pro- ~jection welder with suitable electronic controls has been ordered for further ~experiments. It is expected that the production of amultitude of resistance " welds with consistent results will ' require precise mechanical and elec- - tronic control, : BRAZING TECHNIQUES P. Patriarca and G. M. Slaughter Metallunrgy Division An introductory study has been directed to the subject of furnmace brazing under a controlled hydrogen atmosphere. The use of a commercial high-temperature brazing alloy (Nicrobraz) has been investigated, and the effects of some of the brazing From the it appears that Nicro- brazing 1s an entirely feasible technique of fabricating reactor components. However, since Nicrobraz contains on the order of 4% boron, the use of this brazing alloy for fabri- cation of reactor components 1is some- ~what limited. Tests performed using a modified Nicrobraz in which beryllium - was substithted for boron have shown - little promise. The wetting properties of this modified alloy were poor. Further research i1s necessary to completely evaluate thls approach to the problem. ~ variables have been studied. work to date, Effect of Brazimg Time. The in- fluence of brazing time on Nicrobraz " joints made in type 316 stainless steel can be seen in Fig. 13.7. The - joint held for 5 min at the brazing temperature shows somewhat less diffusion of the brazing alloy 2nto the base metal than that held at the brazing temperature for 30 min, indi- cating that time at temperature is not critical within these limits. A similar Nicrobrazed joint on 304 stainless steel was struck by a sharp hammer blow at 1500°F. The high- temperature ductlllty of the joint was excellent. Brazing of Ctad Fuel Elements. A preliminary investigation has been conducted on the brazing of clad fuel elements., From the rather limited amount of work that has been done on this subject, 1t seems. entirely feasible that these elements can be brazed either to a stainless steel sheet or to anether clad fuel element, Figure 13.8 shows a clad fuel element Nlcrobrazed toa stainless steel sheet. As can be seen, the brazing alloy wets the stainless steel very well and apparently has little tendency to float any of the fuel powders to the surface. A Nicrobrazed edge-to-edge joint of two clad fuel elements was formed, and, although stainless steel in this case 1s also wet very well by the brazing alloy, the powder appeared to be recessed in each sheet, This probably resulted from the preliminary grinding operation, since no brazing alloy was observed in the recessed area and the powder did not appear to be contaminated by the brazing alloy. Nickel-Palladium Brazing Alloy. In another attempt to find a brazing alloy with a more favorable nuclear ecross- section, consideration has been given to the nickel-palladium system with minor additions of silicon for melting point control. Figure 13.9 is a photomicrograph of an inconel tube-to- header joint using 40% nickel —60% palladium alloy brazed at 1270°C in dry hydrogen for 20 min. As can be seen, the flow characteristics and wetting properties are apparently good. Room temperature tube-to-header tensile strengths were of the order of 80,000 psi. Further consideration will be given this alloy system pending results of corrosion tests being con- ducted in molten coolants and fuels, 135 ANP PROJECT QUARTERLY PROGRESS REPORT 136 E SECRET | Y-4779 .- s =, T % N ‘ - /j?-:::?;‘,‘ ,<;,$:';;‘_:_,? (b) Fig. 13.7. Longitudinal Sections of a Type 316 Stainless Steel Tube-to-Header Joint Nicrobrazed at 1120°C in a Dry Hydrogen Atmosphere. Etchant, aqua regia and glycerin. 75X. (a) Nicrobrazed for 5 min. (&) Nicrobrazed for 30 min. FOR PERIGD ENDING DECEMBER 10, 1951 ‘ Fig. 13.8. ‘Steel=-Clad Fuel-Element Butt Nicrobrazed to a Stain- less Steel Sheet. CREEP AND;STRESS RUPTURE OF METALS R. B. Oliver G. M. Adamson C. W. Weaver Metallurgy Division ‘ The laboratories for creep and stress-rupture tests and for the Etchant, none. Transverse Section of a Stainless 40X. testing of metals immersed in liquid media are essentially complete. Extension measurements in the forwer have been modified for greater accuracy and recalibrated. A stress-rupture curve has been obtained for both "coarse' and "fine" grained inconel at 137 ANP PROJECT QUARTERLY PROGRESS REPORT 13. 9, Fig. with a 60% Pd - 40% Ni Alloy in Pry Hydrogen at 1270°C for 20 min. aqua regia and glycerin. 75X, 815°C. Several elongation curves were obtained for a variety of heat-treated nickel Z (Duranickel) specimens, Operation of Creep and Stress- Rupture Equipment. The creep and stress-rupture testing laboratory was placed in operation without having been properly instrumented in order to train technicians and to avoid the delay associated with the development of an extension-measuring system, During the past quarter it was neces- sary to revise the existing measuring system as 1t showed the creep of the entire specimen, the threaded con- 138 Transverse Section of an Inconel Tube-to-Header Joint Brazed Etchant, nections, and pull rods, and the seating of the knife edges of the lever arm, as well as the differential expansion of the entire assembly and an effect from the sealing bellows. The extension is now measured by attaching a scribed platinum strip extensometer across the gage length of the speeimen. Marks are selected on the center strip, and either of the side strips and the distances between the pairs of marks are measured with a micrometer microscope having a least scale division of 0.00005 in. The Instrument Development Group has built FOR PERIOD ENDING DECEMBER 10, 1951 and 1s testinga recording extensometer designed around a motor-driven microme- ter screw. This system was selected because 1t has an optimum combination of range, sensitivity, and stabality. The lcading of all machines was checked with a 65-mil sheet specimen on which were mounted two SR-4 strain gages connected in series. This standard specimen im turn was cali- brated on the Baldwin tensile-testing machine using a Baldwin type L strain indicator. The effective area of the bellows brated specimen in the furnace chamber: then at several pull-rod locads the change in specimen load was observed for vacuum and several positive pressures, The area so measured averaged 3.47 in.?; for example, under vacuum the bellows delivers a com- pressive load of 51 1b, Creep-Rupture Tests of Inconel. Seventeen creep-rupture tests on inconel sheet have been run or are in process. These tests were made at 815°C in an argon atmosphere. Figure 13.10 18 a plot of the logarithm of stress vs. the logarithm of rupture was evaluated by enclosing the cali- time. The percentage figures placed UNCLASSIFIED DWG. Y-5304 e (\ 20,000 \\\\\\\ "~ N ~ FINE GRAIN ~ Nt ~ e \"'--.. ~ 23%“;\\\\ 10,000 : N x\\\ A I ‘:‘__" \\_ P e L e e & ~. | K] 5% , @ R e seanl LTI YT 7 £ L I s P S S COARSE GRAIN — 15% ® 5,000 \/’ .- < 3% g ::xm\ I g \\\\\,\ I N X\ 8% \ \..\ ~ I~ \'\ e 2,000 1,000 10 100 1,000 Fig. 13. 10. RUPTURE TIME (hr) Stress-Rupture Time for Inconel Sheet. 139 ANP PROJECT QUARTERLY PROGRESS REPORT on the chart are the elongations determined from measurement of the specimens before and after testing, One set of points relates the stresses and rupture times for specimens from the annealed sheet and is indicated as fine grained. The other set of points refers to specimens that were heated to 2050°F for 1 hr and air cooled prior to testing; 1s designated as this curve coarse grained. These grains were about 16 times larger than the grains in the annealed sheet, Creep ©f Nickel Z. Five tests were conducted on Nickel Z (Duramnickel) sheet: are presented the elongation-vs.-time curves in Fig. 13.11. The elongations indicated were measured with a dial gage on the pull rod and include from several These were conducted at 815°C in argon atmosphere, and, as indicated, several of the specimens hence sources. erTrors tests were in the cold-rolled temper and others were quenched and aged. It is interesting to note that the effect of heat treatment is retained at the testing temperature, Lack of data for materials at elevated temperatures under nonoxidizing conditions has Stress-Rupture Tests. adequate stress-rupture required conducting tests to provide this information, Values obtained in this laboratory 1n molten sodium, fluorides, and other materials will be compared with those obtained elsewhere in vacuum and inert atmospheres. Because many of the stresses found in reactor auxiliary systems are hoop stresses, these were the first type to be investigated. Test specimens were prepared by machining tube samples to desired over a Z-1n. section, welding one end closed, applying helium pressure inside the tube, and then iwmersing the tube 1n a thickresses UNCLASSIFIED Y-5303 R, — E— E | | | | 030 _ SPECIMEN DIMENSIONS: 0.024-in. THICK - Y -in. X 3-1n, TEST SECTION — . 815°C IN ARGON ] o X - — AS RECEIVED - 2750 psi I 024 f--- DEAD LOADED - LDAD ADJUSTED FOR CHANGES — 3 N BELLOWS SPRING LOAD z ] = — < s £ 018 . £ ~ — ] £ z x - S / AS RECEIVED 2 o2 — J 1500 psi ] o «—AS RECEIVED ~ 1950 psi QUENCHED AND AGED %ifl_f/,—"E'DBCONTWUED 5 - * / / 2750 psi e | o A /___.A——“ / . .@/O e ] A}f::::::lk,,,—~fflé'“ QUENCHED AND AGED rf‘__,a,@.—i:::f;afi 1950 psi . ——/‘___;_/___,,_J_er o e =t P DISCONTINUED | ° -e @ - J 400 600 800 1,000 1,200 TIME IN TEST (hr) Fig. 13.11. Time-Elongation Curve for Nickel Z. 140 FOR PERIOD ENDING DECEMBER 10, 1951 bath of molten material for 1000 hr or unti]l 1t ruptured. Two tubes of type 316 stainless stee]l lasted 1000 hr in 1500°F sodium with hoop stresses of 1900 and 2600 psi, respectively. Examination revealed little or ne stress cor- rosion or attack on either specimen. On the other hand, two inconel samples stressed at approximately 1500 psi failed 1n less than 1000 hr, these failures apparently resulting from intergranular attack inside the tube rather than from exterior sodium attack. An inconel sample was later prepared from a new lot of seamless tubing and stressed to 1200 psi in sodium; it failed in 465 hr through a small area which seemed to be high in impurities. Again, the rupture was due to interior attack although exterior attack up to 0.003 in., deep was noted. As a supplement to the above tests a program for determining stress- rupture values by applying a dead load to flat thin (0.020-1in.) plates of test materials immersed in sodium is being carried out, Approximate values obtained for inconel stressed at 13,000 and 10,000 psi and for type 316 stainless steel stressed at 7000 pSi indicate that sodium at 1500°F has no marked effect. Both the tube-burst and stress- rupture machines are undergoing modifi- cation to permit use ofmolten fluorides as bath materials., | CERAMIC LABORATORY T. N. McVay, Consultant Metallurgy Division Active work in placing the laboratory in operation was started in August and Pre- liminary work on protective coatings for the radiator of the ARE has been initiated at Ohio State University. lLaboratory facilities for carrying on this work at OBNL are now virtually complete, 1s now practically complete. Since Cr-UQ, cermets are being considered for fuel elements, ment for processing and cermets is being designed and should be in operation within four months. Some work has been done on the hot pressing of alumina, and it is planned to proceed with a study of a Al,0,-U0, material for fuel elements. eguip- testing Fquipsent. A vacuum dilatometer and a Smith vacuum furnace for phase studies has been completed, and the dilatometer i1s in operation. Work has started on a vacuum induction furnace, and apparatus for tensile strength tests at high temperatures has been ordered. Egquipment for the hydro- static pressing of ceramic bodies; a high temperature X-ray camera, a molybdenum wound tube furnace, and specific heat and conductivity eguip- ment are 1in the design stages. Subcontract Work. A working agree- ment has been negotiated between Carbide and Carbon Chemicals Company and the Electrotechnical Laboratory of the U. S. Bureau of Mines, and work on beryllia crucibles and other special refractory shapes has started. Thomas Shevlin, of the Ohio State Engineering Experiment Station, is working on Ni-BeO cermets for valve seats to hold liquid metals. Also, hot-pressed beryllia 1s available at the Experiment Station for various refractory shapes including valve parts. | 141 ANP PROJECT QUARTERLY PROGRESS REPORT 14. D. S. Billington, A. J. Miller, Studies have continued on the effects of radiation on the physical and chemical stability of the con- stituents of the airplane reactor. Considerable data have been obtained from experiments in the X-10 graphite pile, and emphasis has shifted to preliminary experiments in the higher flux LITR facility. 1In all phases of the work preparations are being made for further experimentation in the LITR and tests in the MTR. Additional radiation damage experiments have been carried out with the Y-12 and Berkeley cyclotrons. The work has been concerned to a large extent with the stability of the fused fluoride salt mixture (NaF-KF-UF,, 46.5-26-27.5 mole %) proposed as fuel for the sodium-cooled reactor. In the X pile a series of runs was made in which approximately 65 watts was dissipated for 450 hr in each cubic centimeter of U?3%_enriched fuel, With the Y-12 cyclotron energy dissipations up to 415 watts/cm® for 1 hr were achieved with 20-Mev protons. In both cases no evidence of radiation damage was observed. Some radiation damage appeared to take place in the single preliminary experiment wade in the LITR at about 1000 watts for 115 hr, but a more exacting control run and additional experiments are necessary before a conclusion can be reached 1in the matter. The low uranium content of the fused fluorides proposed for use 1n the circulating-fuel reactor would reduce the power production during pile irradiation by a factor of 15 or 20. In this case only in the MTR and cyclotron can power dissipations of aircraft reactor intensity, about 142 RADTATION DAMAGE Physics of Solids Institute searc 1rector’s Jivision He rch Director’s 1) 1 3000 watts/cm®, be achieved. Prepa- rations for experiments with circu- lating type fuels and materials with related compositions are underway. Additional information on the fuel stability studies and i1nformation from experiments on fused KOH stability, thermal conductivity of metals, and creep are contained in the following sections of this report, Complete details onall radiation-damage studies are contained 1n the Physics of Solids Institute quarterly report for period ending October 31, 1951. IRRADIATION OF FUSED MATERIALS G. W. Keilholtz, Materials Chemistry Division The effects of radiation on the stability of the fused fluorides and on fused KOH have been under investigation using the X pile, the Y-12 cyclotron, and the LITR. In experiments with fused fluorides in inconel at 1470 to 1500°F no evidence of radiation damage was found in the X-pile or cyclotron run. Some evidence of what appears to be radiation damage was found in the single preliminary LITR experiment, which still requires control and check runs, Data from the single experiment on escape of xenon from the melt was inconclusive, Pressure build-up measurements on capsules of fused KOH in the X pile and LITR indicated no instability to radiation, Pile Irradiation of Fuel and KO capsules (J. G. Morgan, H. E. Robertson, C. C. Webster, P. R. Klein, and B. Kinyon, Physics of Solids Tnstitute). For reasons of safety, capsules con- taining the fused materials were first FOR PERIOD ENDING DECEMBER 10, 1951 checked in the pile for pressure evolution at low fluxes, and then at higher fluxes. The capsules were pressurized with helium and the pressures were continuously measured by means of a strain gage to an accuracy of #0.5 psi. The standard NaF-KF-UF, fuel and KOH were tested as shown in Table 14.1 with no evidence of pressure increases due to irradi- ation. When pressure tests on the fused fluoride mixture were completed in the X pile, 0.5 g samples of the material enriched in U??® were subjected to a thermal flux of 10'? at 1472°F wall temperature in sealed inconel capsules. After several preliminary tests, five capsules of 0.223 1.d. were irradiated for 300 to 450 hr at 1472°F and one at pile ambient temperature., Chemical and metallographic analyses, which are mostly completed, have showed no in- crease of iron, chromium, or nickel in the fuel, decomposition of the fuel, or observable effect on the inconel as compared to bench tests. The capsule in Table 14.1, which was used in the LITR pressure test, with a power dissipation of about 1000 watts/cm?, showed increased damage in all three Table Tests on Standard NaF-KF-UF, and KOH respects. However, a control sample prepared simultanecusly is yet to be bench-tested and examined. In order to determine 1f xenon is evolved from the fluoride melt under irradiation, the following experiment was conducted. A sample of enriched urenium metal was placed in a flux of 10'? neutrons/cm® and irradiated for the same length of time as a sealed microcapsule of melted salt. Both were removed and the microcapsule was opened., When they were counted, no significant difference between the twe was noted, which seemed to indi- cate that the xenon formed did not escape from the melt. This test is not considered conclusive and a more significant experiment is planned. Cyclotrom Irradiation of Fuel and KOH Capsules (W. J. Sturm and M. J. Feldman, Physics of Solids Institute; R, J. Jones, J. S. Luce, and C. L. Viar, Electromagnetic Research Di- vision). Sixteen inconel pins, 0.052 and 0.100 in., i.d., containing the standard fuel were bombarded with 20- Mev protens. Thermocouples on the irradiated surface of the pins indi- cated average run temperatures between 14.1 MATERIAL FREE SPACE APPROXIMATE | WEIGHT | CAPSULE [CAPSULE | TOTAL | MATERTAL | THERMAL FLUX | TIME MATERI AL (g) | MATERIAL | (em®) | (ew®) | TEMP. (°F) | PILE | (n/cm® - sec) | (hr) Fael Normal uranium 10 Inconel 6 17 200-1472 (X pile| 8.5 x 10! 12 93. 4% U335 10 Inconel 6 17 1472-1560 | X pile| 8.5 x 10 {170 0.5 Inconel 2 13 1370-1500 |LITR 1.6 x 10*3 | 116 KOH 5 316 Stain- 6 17 805 X pile| 8.5 x 10! 16 less steel : 3 Inconel 3 13 805 LITR 1.6 x 103 68 143 ANP PROJECT QUARTERLY PROGRESS REPORT 1200 and 1475°F. Pins irradiated with low beam currents were usually radi- ation-cooled, while those irradiated with high currents were attached to a cold-water coil. In the case of the water-cooled pins there was probably a layer of solid fuel at the water- cooled face., Five of the 1nconel cases melted owing to local over- heating, which probably occurred in all pins to some extent. Energy dissipations between 30 and 415 watts per cubic centimeter of fuel were achieved in various runs, usually of 1 hr duration. Analyses of the fuel and metallographic examinations of the capsules indicated no radiation- induced damage when comparison was made with the from eight bench-tested Several irradiations by deuterons of fuel 1in inconel containers have been made by the North American Aviation group with the Berkeley 60-in. cyclotron. The results are being analyzed. resnlts controls. Stainless steel pins contalning fused KOH have been irradiated with protons, and analyses of the results are in progress. Capsules containing dilute circulating type fuel are being prepared, IN-PILE CIRCULATING LOOPS W. E. Brundage R. M, Carroll C. D. Baumaan C. Ellis Physics of Solids Institute 0. Sisman W. W. Parkinson A. S, Olson The general results of circulating lithium at 1000°F through hole 58N of the X pile were reported last quar- ter.¢!) A detailed metallographic (D¢, b Baumann, R. M. Carroll, Q. Sisman, W. W. Parkinson, and . Fllis, “*Liquid Metals In-Pile Loop,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending gs%gsnbcr 10, 1951, OBNL-1154, p. 174 (Dec. 17, 144 examination now has beena made of various sections of the 316 stainless steel loop. No evidence was found of increased intergranular penetration or other damage due to radiation, An inconel loop for circulating sodium in the X pile at 1500°F has been designed and partially con- structed., Design of equipment 1is nearing cowpletion for stress cor- rosion and creep tests on inconel 1in circulating sodium at 1500°F in the LITR. CREEP UNDER IRRABIATION J. C. Wilson J. C. Zukas W. W. Davis Physics of Solids Institute Work during last quarter(?) on cantilever creep showed that X-pile radiation at a flux of 4 x 10!° fast neutrons/cm? caused an increase 1in total creep strain of about 20% 1in 347 stainless steel after about 250 hr of exposure, the duration of the tests., Extrapolation of the bench and in- pile curves to longer times indicated that the difference between them in- The first of the current tests should supplement the above observations. The time has been extended to 500 hr under the same conditions (1500°F at 1500 psi), but, because the expected beam deflection would exceed the range of the micro- former, the transducer has been omitted and fiducial marks on the loading beam and baseplate will permit creased with time. measurement of the total extension after the experiment has been with- drawn from the reactor. The deflection will then be compared with that observed in a bench test at the same (2)J. C. Wilson, J. C. Zukas, and W. W. Davis, ‘Creep Under Irradiation,”’ ORNL-1154, op. cit., p. 170. FOR PERIOD ENDING DECEMBER 190, temperature and stress levels. The irradiation has been completed, and the activity of the apparatus is being permitted to decay to a safe level for handlaing. Metallographic and X-ray comparison of bench and irradiated samples will be made. A second test is still in the pile. It was stressed to B8.000 psi and for its first 500 hr has been cperating at 1200°F. The temperature will be subsequently raised in steps to observe the temperature dependence of strain rate under irradiation for comparison with a companion bench test. Prepa- rations for tensile type tests and for experiments in the LITR and MTR are in progress. (A, F. Cobhen, *“Radiation Effects on Thermal Copductivity,” ORNL 1154, op. cit., p. 171. 1951 RADIATION EFFECTS ON THERMAL CONDUCTIVITY A. Foner Cohen, Physics of - Solids Institute In a preliminary relative thermal conductivity experiment on inconel, reported last quarter,(3) a large decrease 1in thermal conductivity was observed upon three davysofirradiation at approximately B825°C in the X pile. To check this result, a carefully heat-treated inconel specimen has been irradiated at 575 and 250°C with ne apparent effects due to the radiation, Testing at 825°C is 1n progress. In addition to the above experiment, an absolute measurement of the thermal conductivity of a 316 stainless steel specimen at 100 to 200°(C is being carried out in the LITR, 145 SUMMARY AND The survey of the supercritical- water reactor by the Oak Ridge National Laboratory’s subcontractor Nuclear Develepments Associates, Inc., is essentially complete and has confirmed the feasibility of aircraft nuclear propulsion using that reactor. Although ORNL will not further pursue this reactor cycle because of the prevailing belief in the greater potentialities of low-pressure liguid-coolant re- actors, NDA will continue its work, together with Pratt and Whitney, under direct AEC contract. The new results in the NDA survey of the cycle for ORNL are outlined in Sec. 15, together with a brief cycle analysis that was undertaken at ORNL. Alarge analytical chemistry program is required to support the materials research program now being undertaken by the ANP project. This includes not only routine service analysis ~~ 688 individual samples were analyzed during the past quarter = but 1n many cases the development of new analytical procedures. This development, as reported in Sec. 16, is largely con- cerned with the determination of oxygen in gases and fused mixtures and similar studies of hydroxides and fluorides. ' INTRODUCTION The "List of Reports Issued,” Sec. 17, includes seven laboratory reports and 39 informal documents on ‘all phases of the ANP project. A directory of the research preojects of the Aircraft Nuclear Propulsion Project of the Oak Ridge National Laboratory is given in Sec. 18. The research projects of the Laboratory’s subcontractors to its ANP project are iisted, as well as the research now in progress at the lLaboratory. In additzon, such research as is being performed by OBNL for the ANP programs of other organizations 1is included and marked as such. A chart of the technical organi- zation of the Aircraft Nuclear Propul- sion Project at OBNL is included (Sec. 19) to identify the personnel and emphasis associated with the various phases of the project. There are now about 2ZB5 scientific and technical people employed on the project and about 26 active consultants, Although contractural changes at the start of this fiscal year shifted several former Laboratory subcontrac- tors to the AEC, there are, in all, seven allied laboratories performing research for the ANP project, where ORNL has either the direct contract or technical supervision. 149 ANP PROJECT QUARTERLY PROGRESS REPORT 15. For the past year (see p, 1v for list of previous reports) Nuclear Development Associates have been studying the supercritical-water reactor first proposed in Wash-24.,(1) A first survey analysis by NDA 1s essentially complete and a final report is being written. This report, when 1ssued, will complete the sub- contract work by NDA for OBNL on the supercritical-water reactor. NDA will, however, continue i1ts work in this field under subcontract with Pratt and Whitney Aircraft Division, In addition to the above-mentioned survey of the supercritical-water reactor, an analysis of the cycle was undertaken at OBRNL. This analysis, which will also soon be 1ssued as a separate report, is summarized below. ANALYSIS OF WATER SUPERCRITICAL REACTOR BY NDA The only new results since the last quarterly report pertaining to the supercritical-water reactor system are in the control aspects. The reactor has been shown to be stable for small oscillations, and a start-up technique has been outlined in which the control mechanism is derived from change of water density in the reactor. Stability. A calculation has been carried out¢?) on the stability of the design point reactor (400,000 kw, 2.5-ft square cylinder core, flux flat radially) connected between constant- pressure reservoirs, The system was studied with equilibrium xenon con- centration, Under the assumptions made, the stabilizing density effects (1)Aircraft Reactor Branch of USAEC, Appli- cation of a Water Cooled and Moderated Reactor to Aircraft Propulsion, Wash-24 (Aug. 18, 1950). (2)NpaA Quarterly Report on ANP Activities, June 1 to August 31, 1951, Y-¥5-55. 150 SUPERCRITICAL-WATER REACTOR on reactivity were found to compensate the unstabilizing effect of the xenon; the reactor i1s stable for swmall oscillations., The longest two decay periods were calculated as 5.9 min and 22 sec. The response to a sudden step- function i1ncrease in reactivity 1is such as to cause the power level to overshoot 1ts equilibrium increase by a factor of 10. When the reactivity is raised linearly during an interval of 2 sec, the overshoot is by a factor of 2; when 10 sec is taken to accomplish the linear reactivity increase, there is no overshoot. Start-up. The start-up problem has been considered in a preliminary way with the reactor connected to a simple heat-exchanger external systemand with no control mechanism other than that afforded by changes of water density in the machine. The envisaged start-up procedure 1is as follows: (1) Fill the system with high-temperature high-pressure water whose density i1s too low to make the machine reactive; (2) keeping the system at constant pressure, clrculate and slowly cool the water; the cooling rate is adjusted(3®) so as to bring the reactor power safely up to some fraction (say, 3%) of rated output, the circulation rate being large enough so that the water density is still substantially uniform throughout the reactor; (3) adjust the cooling rate so as to hold the reactor power con- stant at this 3% level, and slowly reduce the circulation rates through the moderator chamber and fuel tubes, while the density variation along the fuel tubes builds up to the normal operating pattern, (3)After the nmeutron flux has become suf- ficiently strong, the flux can be employed for automatic control of this adjustment. FOR PERIOD ENDING DECEMBER 10, 1951 Questions concerning the external equipment called for and how and when the reactor can be switched to connect to the power plant system have not been 1nvestigated. ANALYSIS OF SUPERCRITICAL-WATER REACTOR BY ORNL One of the most widely advocated methods ofutilizing the supercritical- water reactor for aircraft nuclear propulsion is to use the compressor- jet propulsion cycle. A brief study of this fundamental cycle to discover the relations existing among the various parameters involved 1s de- scribed, Compressor-Jet Cycle. In this cycle energy is added to the airstream in a low compression ratio blower and then in a radiator. The supercritical steam from the reactor expands through a turbine which operates the air blower and the water-return pump. The turbine exhaust steam is then con- densed, and 1t transfers its heat of vaporization to the airstream flowing through the radiator. The particular parameters that have been considered - condenser air- inlet face area, condenser weight, and were evaluated per terms of the temperature and pressure:of the steam leaving the reactor and the condenser airflow rate -— pound of thrust 1in pressure. This was accomplished for flight at 45,000 ft altitude at Mach numbers of 0.9 and 1.5. Reasonable values were assumed for component efficiencies and actual test data used for the condenser perfor- mance. The results should give attainable performance of the funda- mental cycle at the two design points investigated. Somewhat better per- formance can probably be obtained through a program of optimization of the cycle and equipment., However, the data presented here should give per- formance not far from the optimum, and hence should be useful for preliminary design studies of the system. Results. The results of this in- vestigation indicate that for reactor outlet steam conditions ranging from 1000 to 1500°F and pressures from 5000 to 10,000 psi, the following statements can be made: 1. The specific impulse is low 1in all cases (from 15 to 20 1b per pound of air per second at a flight Mach No. of 0.9 and from 9 to 14 1b per pound of air per second at a flight Mach No. of 1.5). 2. Increasing the reactor ocutlet steam temperature or pressure or condenser pressure effects some improvement in all cases and in all parameters considered, 1.e., specific impulse, specific heat consumption, and specific radiator weight and frontal area. 3. Increasing steam condenser pressures above 400 psi1 gives relatively littlé improvement in performance. reactor steam from 5000 to relatively 4. Increasing the outlet pressures 10,000 ps1 gives litctle improvement in perfor- mance. 5. Cycle performance is insensitive to the relative amounts of energy put into the air by the turbine- compressor and by the condenser. In general, the obvious advantages of this cycle are the use of water, a familiar and fairly noncorrosive substance, for both coolant and moderator., The disadvantages include an inherently low specific impulse and the necessity of developing an entirely new type of aircraft engine. 151 ANP PROJECT QUARTERLY PROGRESS REPORT 16. ANALYTICAL CHEMISTRY C. D. Susano, Analytical Chemistry Division Since the use of diatomaceous earth as an insulating agent in the aircraft reactor is being contemplated, studies are presently being made of the compo- sition of these materials with partic- ular attention to elements of high neutron-capture cross-section, of which boron and the rare earths are the most important. Preliminary results on two samples of diatomaceous earth indicate that these materials may contain several hundred parts per million of boron, which i1s higher than can be tolerated. It has been mecessary to develop or adapt methods for the determination of metallic nickel, nickel oxide (Ni0), and available oxygen in alkali hydrox- ide melts in connection with tests in which these hydroxides are circulated through nickel loops at 1400°F. These methods appear to be satisfactory with the exception that, if the nickel occurs 1n a massive crystalline form, difficulty is encountered in separating it completely from the oxide. As a part of the corrosion test program for the evaluation of fluoride eutectic coolants, methods are under study for the determination of iron, nickel, chromium, molybdenum, and copper in these materials. Methods for the precise determination of the major constituents are also under consider- ation. A study of the pH values of aqueous solutions of one of the alkali fluoride cutectics and its components is reported. Tests were made to determine whether or not alkali and alkaline earth hydroxide melts could be removed from metal tubes by dissolution in water without appreciable corrosion of the container material during the disso- lution step. Monel, inconel, nickel, 152 and stainless steels 316 and 347 were attacked to only a negligible degree and copper was corroded only slightly more. The development of two methods for the determination of uranium trifluo- ride has been completed. It appears that the method involving measurement of the hydrogen evolved on reaction of of the UF, with acid is more precise than the method which depends upon the total oxidizing power of the sample. Although some additional work remains to be done, 1t appears that oxygen can be determined in sodium-potassium alloy (NaK)} by a modification of the n-butyl bromide method for the determination of oxygen in sodium. A survey of the results obtained 1in the determination of oxygen in tanks of Burcanof Mines helium 1s presented. Approximately 85% of the cylinders tested contained less than 25 ppm of oxygen and were, therefore, acceptable for use by the ANP Experimental Engineering Group. Tests are being made for the purpose of producaing a borated water solution (1% boron) at minimum cost which will retain 1ts clarity 1n contact with concrete, 1nconel, and 1iron. Summaries of the service analysis work, which indicate the distribution with respect to sample type and groups originating the samples, are presented. STUDIES OF DIATOMACEOUS EARTH J. C. White and W. J. Ross Analytical Chemistry Division The use of diatomaceocus earth as an insulating agent in the aircraft FOR PERIOD ENDING DECEMBER 10, 1951 reactor 1s contemplated, and studies are being made of the composition of these materials with particular atten- tion to elements of high cross-section for thermal-neutron capture, chiefly boron and the rare earths. Diatomaceous earth, which consists almost entirely of the siliceous skeletons of minute marine animals, is composed essentially of 5i0, but may vary significantly in minor and trace constituents. An effort is being made to find an infusorial earth of suitably low boron content (less than 10 ppm). ' Spectrographic analysis of two types has been completed todate: one, a type used primarily as a paint pigment, showed 600 ppm of boron and no detectable rare earths; the second, a "fresh-water™ earth with the trade name "Sil-o-cel,” showed 300 ppm of boron and ne detectable rare earths. Since these concentrations are higher than can be tolerated, the possibility of reducing the boron content of these materials by washing with hydrochloric acid and other solutions is under investigation. A new method for the determination of boron is under study.(!’ This method depends upon the color complex between boron and 1,1°-dianthrimide which is formed in sulfuric acid 1n aqueous medium. Preliminary tests indicate that the method is extremely sensitive (0.0l ug of boron per milli- liter). DETERMINATION OF Ni, NiO, AND O IN ALKALI HYDROXIDES J. C. White Analytical Chemistry Division A series of tests has been conducted in which lithium, sodium, and potassium (1)G, M. Ellis, E. G. Zook, and O. Baudisch, “"Colorimetric Determination of Boron Using 1, 1’-Dianthrimide,” Anal. Chem. 21, 1345 (1949). hydroxides were circulated through nickel loops at 1400°F. The loops were operated for various periods, ranging from 51 hr for KOH to roughly 300 hr for LiOH, before the runs were terminated because of the formation of plugs. The loops were X-rayed in order to determine the locationof the plugs, cut in sections, and submitted for both chemical and metallographic exami-~ nation. Some development effort was necessary to arrive at suitable methods for the determinationof nickel, nickel oxide, and available oxygen in these plugs. Available Oxygen. A method similar to that employed for the determination of available oxygen in lead peroxide is used in this case. The amount of iodine liberated from an acidic iodide solution is titrated with standard sodium thiosulfate. 1In only one case was an appreciable amount of available oxygen found in the samples submitted. Metallic Nickel. Since the plugs were formed from large, lustrous nickel dendrites, metallic nickel was deter- mined directly. Samples from other sections of the loops were characterized by the presence of much more finely divided nickel, and i1n these cases the displacement technique was applicable. This method depends upon the displace- ment from solution of a metal lower than nickel in the electromotive series. The excess of the added 5% copper sulfate reagent can be deter- mined and is a measure of the amount of metallic nickel originally present. Nickel Oxide. Work is underway to separate nickel and nickel oxide when the metal exists largely in a massive crystalline form. The use of i1odine to convert the metal to Nil, will be investigated. In samples 1in which the nickel 1s more finely divided, the oxide is separated out by filtering the inscluble nickel oxide and copper metal from the solution used for the 153 ANP PROJECT QUARTERLY PROGRESS REPORT determination of nickel by the dis- pl acement method. The residue is digested with acid, and the nickel 1is determined by the dimethylglyoxime precipitation. Studies of the higher oxides of nickel are being made in an attempt to gain an understanding of the reactions which are taking place 1n the molten hydroxides under the conditions of these tests. STUDIES OF TERNARY ALKALI FLUORIDE EUTECTIC J. C. White Analytical Chemistry Division A proposed coolant for the ARE reactor is the ternary alkali fluoride eutectic composed of 11.7% sodium fluoride, 59.1% potassium fluoride, and 29.2% lithium fluoride by weight. This eutectic melts at 450°C. VWork on this eutectic is confined at present to study of the pH of aqueous solutions of the eutectic and its components, of the composition of the eutectic, and analysis of the eutectic for metallic impurities. pll of Agueous Solutions of Alkali Fluorides. The alkali fluorides will hydrolize completely at temperatures Table 16.1 pH Values of Aquecus Soclutions of Certain Alkali Fluorides at 25°C MOLAR SALT CONCENTRATION pH LiF 0.05 5.3-5.4 LiF (fused) 0.05 8.3-8 .4 NaF 1.0 7.2-7.3 NaF (fused) 1.0 8.9 KF 7.95 9.2 KF (fused) 7.95 9.9 154 around 1000°C but not significantly at room temperature. Hence, a deter- mination of the pH of an agueous solution of the cooled fused salt at room temperature will indicate the degree of hydrolysis which has taken place at the high temperature required for fusion (pyrohydrolysis) and is a me asure of the moisture present at these temperatures. Table 16.1 shows typical values for the higher pH of the fused salts as compared to com-~ parable concentrations of unheated c.p. reagent grade salts. The following comments may be made: 1. The increase inbasicity on fusion is very likely a consequence of pyrohydrolysis, the water being already present in the salt. Potassium hydroxide, despite its hygroscopic natuvre, appeared more stable to pyrohydrolysais than the other alkali metal fluorides. Agueous solutions of this fluoride are more basic, however, than solutions of either sodium or lithium fluoride. 2. The alkali fluorides may be classified as salts of strong bases and fairly weak acids (the ionization constant of hydroflu- ric acid is 7.7 x 10°*), so that aqueous solutions of these salts should be slightly basic. How- ever, aqueous lithium fluoride tested acidic and sodium fluo- ride nearly neutral. This 1s believed due to the presence of some free hydrofluoric acid or acid fluoride salt in the reagent. 3. pH is relatively independent of concentration except possibly in the case of unfused anhydrous potassium fluoride, the reason for this 1s under current investi- gation. The pH of a saturated solution of the eutectic, determined from three FOR PERIOD ENDING DECEMBER 10, 1951 lots prepared in various container materials, was 9.8, roughly that of a solution of fused potassium fluoride. A fourth lot, prepared using hydrated potassium salt dried overnight under vacuum, had a pH of 6.9, which is nearly neutral. Evidently some basicity may be removed by this drying process. ' Composition of the Eutectic. Pre- liminary studies have been made to compare gravimetric and flame-photo- metric methods for the determination of i1ndividual alkali metals in the eutectic. It appears that flame- photometric methods will not be suffi- ciently accurate but that total alkali and total fluoride determinations will provide all the information required. Total alkali can be determined gravi- metrically and total flunoride will be determined by pyrohydrolysis. Metallic Impurities. As a part of the corrosion test program, the trace impurities iron, nickel, chromium, molybdenum, and copper will be deter- mined in the eutectic. Colorimetrac methods are being developed for this purpose. CORROSION OF METAL CONTAINERS BRY HYDROXIDE SOLUTIONS J. C. White Analytical Chemistry Division The present method for removing alkali and alkaline earth metal hydroxides from metal containers in preparation for the determination of metal constituents involves dissolution of the hydroxide melt with water. This procedure exposes the container to a warm, saturated hydroxide solution for as long as 2 to 3 hr in some cases; hence, a study was made to determine the extent of corrosion of the container metal from this source. Results indicated that this procedure has little effect, and that the present method 1s suitable for all containers with the possible exception of copper. DETERMINATION OF URANIUM TRIFLUORIDE W. K. Miller and D. L. Manning Analytical Chemistry Division In a previous report{?’ two methods for the determination of uranium tri- fluoride were described. It was reported that the ceric sulfatemethod appeared more promising than the hydrogen evolution method, although it was believed that the precision of the latter could be improved by reducing the volume of the apparatus. This modification has resulted in a marked improvement in the results by the hydrogen evolution method, which 1is now considered more precise. Although no standard sample 1s available for a direct test of the two methods, the agreement between the methods indicated that sither can be satisfactorily used to determine trivalent uranium. DETERMINATION OF OXYGEN IN Nak J. C. White and W. J. Hoss Analytical Chemistry Division Efforts are currently being directed toward the application of the n-butyl bromide method(?? for the determination ¢f oxvygen in sodium to the determi- nation of oxygen in sodium-potassium alloy (NaK), a eutectic composed of 22% sodium and 78% potassium. The principal physical and chemical properties of the alloy which are of (D%, K. Miller and D. L. Manning, “Uranium Trifluoride in Uranium Tetrafluoride,” Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1951, ORRL- 1154 p. 202 (Dee. 17° 1981) (3)5. C. White and W. J. Ross, “Determination of Oxygen imn Sodium,” Aircraft Nuclear Propulsion Project Quarterly Progress Report forPeriod Ending March 10, 1951, ANP-60, p. 336 (June 19, 1951) 155 ANP PROJECT QUARTERLY PROGRESS REPORT significance in this particular study are: (1) It is a free-flowing liquid above 15°C, (2) it has a rather high coefficient of expansion upon freezing, and (3) it is extremely reactive toward oxygen. Owing to the high reactivity of NaK with n-butyl bromide, the reagent must be added in small increments rather thanmn in one single portion. The reaction 1s complete within 1 hr for 1- to 2-g samples of NaK. The results obtained on test samples have shown somewhat higher oxygen content (about 0.25%) than has usually been the case for sodium samples. DETERMINATION OF OXYGEN IN HELIUHX J. C. White and W. J. BRoss Analytical Chemistry Division Data for the determination of oxygen 1in helium by the Brady method have shown that about 85% of the United States Bureau of Mines cylinders tested have oxygen contents below 25 ppm, and hence, according to the present standards, are acceptable for use. No attempt was made to determine the average oxygen content, as this method 1s unreliable for concentrations greater than 50 ppm. Because of 1ts wider working range, a modification of the Winkler method, as modified by Pepkowitz,¢*’ has been considered as a substitute for the Brady method. It has been i1mpossible to determine the precision of this method since acceptable low- or zero- oxygen standards have not been obtained. PREPARATION OF OXYGEN-FREE SODIUM SAMPLES H. R. Bronstein, ANP Division Widely varying results on sodium samples taken from operating systems 156 pointed out the unreliability of analytical methods for determining quantitatively the oxygen in sodium. Unreliability was felt to be largely due to lack of suitable standards by which analytical methods could be evaluated. Consequently, a search was begun to find suitable methods for producing samples containing a known amount of oxygen. A literature search revealed a method believed suitable for producing standard samples. DBriefly, the method consists in 1mmersing a highly evacuated glass bulb containing a filament in a bath of molten sodium nitrate. An electrode (anode) is also immersed in the bath external to the evacuated bulb. The filament i1s heated by a 220-v alternating current; at the same time, a 220-v direct current is placed on the filament and anode. Sodium ions are pulled out of the glass, neutralized by electrons from the filament, and replaced in the glass by sodium 1ons from the bath. This method allows pure sodium to be plated out on the inner side of the bulb. deposited may be determined by either weighing the bulb before and after electrolysis or calculated by Faraday’s law of electrolysis. Standard samples may be produced by adding known amounts of oxygen to known amounts of sodium by high-vacuum techniques. Amounts During the guarter equipment was assembled, and preliminary experiments were conducted to evaluate the equip- ment and method. Both appear to be adequate, and sufficient experience has been gained to proceed with preparing actual samples.(®’ L. p. Pepkowitz, private communication. )T, E. Willmarth, “Microscopic Study of a Submitted Sample of Diatomaceous Earth,”” ORNL CF-51-11-12 (Nov. 1, 1951). - FOR PERIOD ENDING DECEMBER 10, 1951 CLARITY OF BORATED WATER IN CONCRETE TANKS(6) H. P. House Analytical Chemistry Division It has been proposed that the con- taining concrete tank of the ARE be filled with borated water (at least 1% boron by weight) after shutdown, while subsequent disassembly operations are performed remotely. Hence, several solutions of boron salts have been tested for prolonged clarity while in contact with concrete. Saturated borax solution met the requirement for clarity, but contained only 0.8% boron by weight. Potassium tetraborate met both requirements, but its higher cost discourages its use. Mixtures cf the two salts in agueous solution form a precipitate on contact with concrete. Final studies are being made te determine the optimum borated solution. (S)H. P, House and €. D. Susano, Clarity of Borated Water in Concrete Tanks, CRCCC Y-12 Memo ¥B-31-273 (July 24, 1951). ANALYTICAL SERVICES H. P. House J. W. Robinson L. J. Brady The bulk of the analytical work carried out during the quarter for the ANP Project consisted of determinations of (1) corrosion products in reactor fuels (fluoride salt mixtures), (2) purity of components used in compounding reactor fuels and for coolants, {(3) corrosion products in sodium, potas- sium, and barium hydroxides, and (4) oxygen and corrosion products in sodium and NakK. A summary of service analyses per- formed this quarter is shown in Table 16.2. Table 16.2 Summary of Service Apaiyses Samples on hand 8/10/51 151 No. of samples received 686 Total no. of samples 837 No. of samples reported 688 Backlog as of 11/2/51 149 157 ANP PROJECT QUARTERLY PROGRESS REPORT 17. LIST OF REPORTS ISSUED REPORT NO. TITLE OF REPORT AUTHOR(S) DATE ISSUED Reactor Physics CF 51-10-83 Enlargement of Cross-section Program C. E. Larson 10-12-51 CF 51-11-92 Pile Simulator Study of Flux, etc. S. Hanauer 11-12-51 Y-F10-73 Suggested Correction to Age-Diffusion R. R. Coveyou 11-6-51 Equation as Used by the ANP Physics Group B. T. Macauley Y-F10-69 Numerical Integration of Differential R. R. Coveyou 8-20-51 Equations; Multi-Point Boundary Problems Y-F10-71 Physics Calculations on the ARE Control Rods R. J. Beeley 8-29-51 OBNL- 1099 The Elements of Nuclear Reactor Theory, S. Glasstone No date Part [ M. C. Edlund Y- B4- 39 Nuclear Properties of U?3*. A Literature E. P. Carter 9-12-51 Search Y-F10-64 Heating in the B,C Curtain Due to Neutron C. B. Mills B8-16-51 Absorption and the B!%(n,a)Li” Reaction Y-F10-67 Effect on Reactivity of ABE of Flooding J. W, Webster §-14-51 Coolant Channels with Borated Water ANP-68 Solution of Kinetic Equations of Cylindrical M. J. Nielsen 9-18-51 Reactor J. W. Webster Y-F10-74 Note on the Doppler Effect R. R. Coveyou 11-26-51 Y-F10-56 Some Results of Criticality Calculations on J. W. Webster 10-15-51 BeQ and Be Moderated Reactors 0. A. Schulze Shielding Research ORNL-1130 Analysis of Bulk Shielding Facility Neutron S. Podgor 11-26-51 Dosimeter Data CF 51-9-112 Power Calculations of the Unit Shield Reactor E. B. Johnson 9-18-51 CF 51-10-70 Introduction to Shield Design E. P. Blizard 10-12-51 CF 51-10-94 Calculations of Leakage from the Bulk Shield- J. L. Meem 10-5-51 ing Facility Reesctor CF 51-10-203 Tentative Comparison of Ionization Chambers R. H. Ritchie 10-31-51 CF 51-10-212 Application of a Scintillation Counter to F. K. McGowan 10-16-51 Gamma Ray Dosimeter C. E. Clifford CF 51-11-95 Experiment V-A at Bulk Shielding Facility; H. E. Hungerford 11-15-51 158 Shadow Shield Measurements with a Na24 Source REPORT NO. CF 51-11-96 CF 51-11-139 Y-F5-57 CF 51-11-168 Y-F12-6 Y-F17-9 ANP-72 ANP-71 CF 51-11-23 CF 51-11-67 CF 51-11-72 Y-B4-38 Y-B4- 41 Y-811 CF 51-11-78 CF 51-9-64 CF 51-10-178 CF 51-11-47 CF 51-10-178 FOR PERIOD ENDING DECEMBER 10, 1951 " TITLE OF REPORT Fast Neutron Measurements at Bulk Shielding Facility Preliminary Estimate of Circulating-Fuel Reactor Shielding The Divided Shield Proposal for Divided Shield Experiments Component of Reactor Systems Status of Reactor Coolant Pump Program Performance Characteristics for a General Electric G-3 Electromagnetic. Pump Containment of Helium in Stainless Steel and Incone] at the 1500°F+ Range Thermal Conductivity of Steel Wool and Some Granular Solids Metallurgy Hydroxide Corrosion Fission Product Analysis of ANP Material High Temperature Mechanical Properties Bismuth ~ Selected Physical Properties in the Temperature Range 100 to 1000°C Selected Physical Properties of Stainless Steel in the Temperature Range 100 to 1000°C Y-12 Alkali and Liquid Metal Safety Committee AUTHOR(S) R. G. Cochran H. E. Hungerford E. P. Blizard L. A, Wills E. P. Blizard C. E. Clifford A. Simon H. L. F. Enlund J. L. Meen J. F. Haines A. G. Grindell = S. G. Wischhusen F. Salmon F. Bailey . D. Manly A. Reynolds H. Boss Martha Wilson Frances Sachs p. L. Hill Physical Properties and Heat Transfer Research Density of One Mixture of NaF-KF-UF, Heat Capacity of Fuel Mix Temperature Distribution in Thin Walled Reactor Passages Review of Air Cycle Heat Transfer Analysis Temperature Distribution in Thin Walled Reactors Having Noncircular Flow Passages . M. Cisar D. Powers S. Farmer . B. Harrison S. Farmer DATE ISSUED 11-20-51 11-26-51 9-17-51 No Date 10-26-51 10-24-51 10-16-51 11-93-51 11-5-51 11-14-51 11-15-51 9-14-51 10-1-51 8-13-51 11-14-51 9-13-51 10-23-51 11-9-51 10-23-51 159 ANP PROJECT QUARTERLY PROGRESS REPORT REPORT NO. ORNL- 1087 Y-F30-3 Y-B31-305 CF 51-10-194 CF 51-11-129 Y-F26-25 CF 51-11-159 Y-F26-23 160 TITLE OF REPORT Heat Cepacity of Molybdenum Forced Convection Heat Transfer in a Pipe System with Volume Heat Sources Within the Fluids Miscellaneous Analytical Chemistry — ANP Program Quarterly Progress Report for Period Ending November 20, 1951 The H,0 Moderated, Salt Cooled Heterogeneous Aircraft Reactor NDA Quarterly Report on Supercritical Water Reactor Work, September 1 to November 23, 1951 Directory of Active ANP Research Projects at ORNL Review of the Feasibility of the Air Cycle Nucl}ear Reactor ANP Information Meeting of Nov. 14, 1951 AUTHOR(S) &= A. Redfield . H. Hil} F. Poppendiek Palmer D. Susano M. Weinberg Gale Young W. J. W. B. Cottrell F. Lane B. Cottrell DATE ISSUED 9-24-51 11-20-51 11-20-51 10-8-51 11-21-51 12-1-51 12-29-51 11-21-51 FOR PERIOD ENDING DECEMBER 10, 1951 18. DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS AT OBNL(1) December 1, 1951 I. REACTOR AND COMPONENT DESIGN A. Aircraft Reactoer Design 1. Core and Pressure Shell 9704-1 Waislicenus 9201-3 Schroeder 2. Heat Exchanger and Radiator 9704-1 Fraas 9204-1 Hamilton 3. Pumps and Plumbing 9204-1 Wyld Haines 4, Control 9704-1 FErgen 9201-3 Bettis 5. Shielding 9704-1 Ergen 3022 Blizard 6. Nuclear Physics 9704-1 Ergen Mills B. ARE Reactor Design 1. Core and Pressure Shell 9201-3 Hemphill ' Wesson 2, Fluid Circuit Design 9201-3 Cristy . Lawrence Jackson Eckerd 3. Pressure and Flow Instrumentation 9201-3 Reese 4. Structural Analysis 9201-3 Maxwell 5. Thermodynamic and Hydrodynamic Analysas 9201-3 Lubarsky 6. Remote Handling Equipment 9201-3 Hutto Alexander 7. Hazards Analysis 9704-1 Ergen 8. Monitoring Equipment for Na Leaks K-1005 Cameron McKown 9. Electrical Power Circuits 3500 Owens Beler C. ARE Control Studies 1. High Temperature Fusion Chamber 2005 Hanauer 2. Control System Design 2005 Epler Kitchen Ruble D. ARE Building Facility 1. Construction 7501 Nicholson Co. 2. Internal Design 1000 Browning (Dmngs directory has been printed separately in a document by W. B. Cottrell Directory of Active ANP Research Projects ¢t ORNL, Y-F26-25 (Dec. 1,(1951). 161 ANP PROJECT QUARTERLY PROGRESS REPORT F. G. 162 Reactor Statics 1. U - . -3 11. 12' 13. 14. Statics of the Circulating Fuel Reactor Parametric Studies of H,0 Moderated Circu- lating Fluoride-Fuel Reactor Statics of the Critical Experiments Statics of the NaOH Moderated Reactor Final Report on Sodium Cooled Stationary Liquid Fuel Reactors Investigation of Simplified Calculations Preparation of Reactor Calculations and Cylindrical Coordinates for the IBM Investigation of Errors in Multigroup Procedures Problem of Minimum Critical Mass Energy Distribution of Thermal and Epi- Thermal Neutrons IBM Calculations for the ORNL ARE Proposals IBM Calculations for the GE-ANP Proposals Age Calculations of Hydrogen-Moderated Reactor (GE) Simplified Reactor Theory Reactor Dynamics 1. 2@ 3. Kinetics of the Circulating-Fuel Reactor Perturbation Calculation of Kinetics of Circulating-Fuel Reactor Kinetics of the Stationary lLiquid Fuel Reactor Critical Experiments 1. Graphite Critical Test Assembly 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9704-1 9213 Mills Macauley Smith Mills Macauley Smith Mills Smi th Holmes Mills Mills Prohammer Edmonson Coveyou Coveyou Coveyou Macauley Uffelman Johnson Macauley lLeeth (GE) Johnsaon Macauley Leeth (GE) Thompson Mills Smith Macauley Ergen Smith Mills Smith Macauley Callihan Zimmerman Williams Haake Scott 2. 3. FOR PERIOD ENDING DECEMBER 10, 1951 Air-Water Critical Assembly (GE) ABE Critical Experiments H. Pump Development 1. 2. Centrifugal Pump-Gas Seal Centrifugal Pump-0il-Graphite Seal Centrifugal Pump for ARE Flectromagnetic Pump Canned Rotor Pump Frozen Sodium Seal Pump Frozen Fluoride Seals Meghanical Seals Rocking Channel Sealless Pump Seals for NaOH I. Valve Development 1. 2. Self-Welding Tests Bellows Tests at High Temperatures J. Heat Exchanger and Radiator Development 1. 2. 4! Ligquid to laiguid Liquid to Aar Boeing Turbojet with Na Radiator Fuel to Liquid 9213 6213 9201-3 9201-3 9201-3 5201-3 9201-3 9201-3 9201-3 9201-3 BMI BMI 9201-3 9201-3 9201-3 9201-3 9201-3 9201-3 Callihan Zimmerman Williams Haake Scott Callihan Zimmerman Williams Scott Haake McDonald Cobb McDonald Grindell McDonald Cobb Haines McDonald Southern Wyld Richardson Blalock McDonald Huntley McDonald Smith McDonald Huontley Dayton Simons Allen Adamson Petersen Reber Reber Fraas Wyld LaVerne Fraas Whitman LaVerne Fraas LaVerne ¥raas Whatman 163 ARNP PROJECT QUARTERLY PROGRESS REPORT 164 5. Htadiator Design Finid Dynamics 1. Full Scale ARE Test Facilities 2, Fuel Manifold Mockups IT¥. SHIELDING RESEARCH Cross-Section Measurements 1. Neutron Velocity Selector 2. Be (n,2n) Cross-Section 3. Analysis for He in Jrradiated Be 4. Total Cross-Sections of Fe 5. Total Cross-Sections of N (GE) 6. Cross-Sections for BeQ and C Shiclding Measurements 1. Divided Shield Mockup Tests (GE) 2. Gamma Shadow Shield Experiment (GE) 3. Bulk Shielding Reactor Power Calibration 4. Bulk Shielding Reactor Cperation 5. Heat Release per Fission 6. Metal Duct Tests 7. Li? Bremsstrahlung Measurement 8. Air Duct Tests {(GE) Shielding Theory and Calculations 1. Survey Report on Shielding 2. Shielding Section for Reactor Technology 3. Calculations of Removal Cross Sections 4, Theory of Neutron Transmission in Water 9201-3 9201-3 9201-3 2005 2005 3026 9201-2 9201-2 3001 3010 3010 3010 3010 3010 3001 3025 3001 3022 9704-1 3022 3022 3022 fFraas Kackenmester Wischhusen Ward Pawlicki Smith Klema Arfken Parker Ergen Willard Bair Kington Willard Bair Kington Clifford Flynn Blosser Meem, etc. Hungerford Johnson McCammon Holland Leske Roseberry Meem Hullings Mockenthaler Sisman Clifford Flynn Blosser Blizard Welton Blizard Blizard Blizard Enlund FOR PERIOD ENDING DECEMBER 10, 1951 5. Interpretation of Pb-H O Lid Tank Data 2005 Simon 6. Divided Shield Calculations 3022 Murray 7. Divided Shield Theory and Design NDA Goldstein Feshbach B. Design of Liquid Ammonia Unit Shield 3022 Blizard . Enlund Wyld 9. Air Duct Theory (GE) 3001 Clifford Simon Pp. Shielding Instrusments 1. Gamma Scintillation Spectrometer 30140 Maienschein 2. Neutron Dosimeter Improvement 3010 Hurst Glass Cochran 3. Proton BRecoil Spectrometer for Neutrons 3010 Cochran Henry 4. He® Counter for Neutroas 3010 Cochran 9. L1 I Crystals for Neutrons 3010 Maitenschein | Schenck 6. Neutron Spectroscopy with Photographic Plates 3010 Johnson ‘ 1 Haydon yHoneycutt E. Shielding Materials 1. Preparation of High Hydrogen Rubber Goodrich Davidson ' Co. 2. Development of Hydrides for Shields MH1 Banus IXI. MATERIALS RESEARCH A. Corrosion by Liquideetals 1. Static Cerrosion Tests in Ligquid Metals 2000 Vreeland ' Day Hoffman 2. Static Corrosion Tests in Liguid Metal Alloys 2000 Vreeland Day Hof fman 3. Static Corrosion Tests in l.ow Melting Point 2000 Vreeland Alloys Day Hof fman 4. Dynamic Corrosion in Na-Isothermal 2000 Vreeland | Trotter 5. Dynamic Corrosion Research in Harps 2000 Adamson 6. Effect of Crystal Orientaticn on Corrosicn 2000 Smith ' Cathcart Bridges 165 ANP PROJECT QUARTERLY PROGRESS REPORT 166 7‘ 8. 9 10. 11. Corrosion by 1. 2. Effect of Carbides on liguid Metsl Corrosion Mass Transfer in Mclten Metals Diffusion of Molten Media into Solid Metals Structure of Ligquid Pb and Ba Alloys, Mixtures and Combustion of Ligquid Sodium Flvorides Static Corrosion of Metals in Fluoride Fuels Static Corrosion of Special Alloys in Fluoride Fuels Static Corrosion Tests in Fluoride Salts Mechanism of Fluoride Corrosion Dynamic Corrosion Tests in Fluoride Corrosion Effect of Flow Velocity on Fluoride Corrosion Effect of Contaminants on Fluoride Corrosion Corrosion by Hydroxides i. -3 . Static Corrosion of Metals Static Corrosion of Special Alloys in Hydroxides Mass Transfer in Molten Hydroxides Physical Chemistry of the Hydroxide Corrosion Phenomenon Static Corrosion by Hydroxides Static Corrosion by Hydroxides Mechanism of Hydroxide Corrosion Anodic Protection of Metals in Hydroxides 2000 2000 2000 2000 2000 2000 2000 9766 9766 9201-3 9201-3 9201-3 2000 2000 2000 2000 9766 BMI BMI 9766 Brasunas Richardson Brasunas Richardson Richardson Smith Bridges Smith Vreeland Day Hof finan Vreeland Day Hoffman Kertesz Buttram Smith Meadows Kertesz Buttram Smath Adamson Coughlen Adamson Coughlen Adamson Coughlen Vreeland Day Hoffman Vreeland Day Brasunas Richardson Cathcart Smith Kertesz Buttram Croft Smith Meadows Jaffee Craighead Pray Kertesz Buttram Smith 9. 10. FOR PERIOD ENDING Dynamic Corrosion Tests in Hydroxides Corrosion. by Homogeneous Fuels Physical Properties of Materials 10 2' 10. 11. Density of Liquids Viscosity of Liquids Thermal Conductivity of Solids Thermal Conductivity of Liqqids Specific Heat of Selids and Liquids Thermal Diffusivity Wetting Electrical Resistance of Fluoride Salts Viscosity of Fluoride Salts Vapor Pressure of Fluorides Vapor Pressure of BeF, Strength of Materials 1. 2. Creep Tests in Fluoride and Hydroxides Creep and Stress Rupture Tests of Metals in Controlled Atmospheres Creep and Stress Rupture Tests of Materials an Liquid Media High Temperature Cyclic Tensile Tests Tube Burst Tests Tube Burst Tests Relaxation Tests of Beactor Materials DECEMBER 190, 9201-3 9766 9204- 1 9204-1 9204-1 9204-1 9204-1 9204-1 9201-3 9201-3 9766 9733-3 BMI1 9201-3 2000 2000 2000 9201-3 2000 2000 1951 Adamson Kertesz Buttram Croft Kaplan Cisar Tobias Kaplan Tobias Powers Chandler Claiborne Powers Jones Tobias Cisar Wischhusen Ward Affel Kertesz Knox Barton Moore Patterson Clegg Adamson Coughlen Oliver Woods Weaver Oliver Woods Weaver Oliver Woods Weaver Adamson Oliver Woods Weaver Olzver Woods Weaver 167 ANP PROJECT QUARTERLY PROGRESS REPORT 168 8. 9. Creep Tests in Thermal Convection Loops Creep Tests of Materials (GE) Hetals Fabrication Methods 1. N » Oy oo W . . ].O. 11. 12. New 1. 2. Welding Techniques for ARE Parts Brazing Techniques for ARF Parts Molybdenum Welding Research Molybdenum Welding Research Resistance Welding for Mc and Clad Metals Welds in the Presence of Various Corrosion Media Nondestructive Testing of Tube to Header Welds Basic Evaluation of Welds Metal Deposits in Thick Plates Evaluation of the Cone Arc Welding Technique Evaluation of the High Temperature Brazing Alloys Development of High Temperature Brazing Alloys Welding in Presence of Hydroxides Wetals Development Mo and Cb Alloy Studies Heat Treatment of Metals Ceramics and Metals Ceramics 1. 2. BeO Fabrication Research Metal Cladding for BeQ B,C Control Rod Development Hot Pressing of Tungsten Carbide Bearings Hot Temperature Firing of Uranium Oxide to Produce Selective Power Sizes Devel opment of Cr-UO2 Cermets for Fuel Elements Ceramic Coatings for Stainless Steel Valve Parts for Liquid Metals and Fluorides 9201-3 2000 2000 2000 BMI MIT RP1 2000 2000 2000 2000 2000 Wall Colmonoy 9766 2000 2000 Gerity Mich. Gerity Mich. 2000 2000 2000 9766 0766 9766 Adamson Oliver Woods Weaver Patriarca Slaughter Patriarca Slaughter Parke Wulff Nippes Vreeland Patriarca Slaughter Patriarca Slaughter Patriarca Slaughter Patriarca Slaughter Patriarca Slaughter Peaslee Kertesz Buttram Croft Bomar Bomar Coobs Graaf Graaf Bomar Coobs Bomar Coobs Bomar Coobs Johnson Shevlin White Shevlin 9. 10. FOR PERIOD ENDING Application of Ceramic Materials to Reactors Crucible Development for High Temperatures I. Solid Fuel Element Fabrication 1. 2. -~ N Sclid Fuel Element Fabrication Diffusion-Corrosion in Solid Fuel Elements Determination of the Engineering Properties of Solid Fuel Elements Electroforming Tube to Header Configurations Electroplating Mo and Cb Carbonyl Plating of Me and Cb Rolling of Fuel Plate Laminates (GE) J. Liguid Fuel Chenistry 10 S O O~ *r & & = Phase Equilibrium Studies Preparation of Standard Fuel Samples Mechanism of Fuel Pretreatment Reaction of Fluoride Fuels with O, and H,0 Ionic Species in Fluoride Fuels ‘evelopment of Homogeneous Fuels Stability of Slurries of U03 in NaOH Phase Equilibria Among Silicates, Borates, etc. Fuel Mixtures Containing Hydrides Chemical Literature Searches K. Liquid Moderator Chemistry 1. Preparation of Pure Hydroxides DECEMBER 10, 1951 9766 Norris Elect. iab. 2000 2000 2000 Gerit Mich) Gerity Mich, 2000 3012 2000 9733-2 9733-2 9733-2 9733-2 9733-3 9733-3 BMI BMI MH1 9704-1 9733-3 Johnson Wilson Doney Bomar Coobs Bomar Coobs Bomar Coobs Graaf Graaf Bomar Cunningham Bomar [l.eonard Barton Blakely Bratcher Barton Nessle Powers Love Barton Nessle Powers lLove Barton Nessle Powers Love Barton Robinson Overholser Redman Patterson Crooks Banus Lee Overholser Nicholson Cuneo 169 ANP PROJECT QUARTERLY PROGRESS REPORT 4’. 3. Thermal Stability of Hydroxide Mixtures Solubility of Metals in Hydroxides Moderator Systems Containing Hydrides Hydroxide-Metal Systems L. Liguid Coeclant Chemistry 1. 3. Phase Equilibrium Studies Preparation of Standard Cooclant Samples Reaction of Fluorides with Alkali Metals M. Radiation Damage 1. o . 11. 170 Liquid Compound Irradiations in LITR Liquid Compound Irradiations in Cyclotron Liguid Compound Irradiations in MTR Fluoride Fuel Irradiation in Berkeley Cyclotron Liquid Metal Corrosion in X-Pile loops Stress Corrosion and Creep in LITR Loops Creep of Metals in LITR and X-Pile Thermal Conductivity of Metals in LITR and X-Pile Diffusion of Fission Products from Fuels Neutron Spectrum of LITR Irradiation of Water (GE) 9733-3 9766 MHI 3550 9733-2 9733-2 9733-3 3005 9201-3 3025 NAA 3001 3025 3025 3001 3005 3001 3001 3005 3550 Overholser Ketchen Nicholson Kertesz Croft Smith Banus Bredig Barton Blakely Bratcher Nessle Powers Love Blankenship Keilholtz Morgan Webster Robertson Klema Kinyon Keirlholtz Feldman Sturm Jones Keilholtz Klein Kinyon Pearlman Brundage Parkinson Ellis Olsen Sisman Bauman Carroll Sisman, Wilson Zukas Davis Cohen Templeton Keilholtz, Sisman Trice Lewis Taylor etc, etc, ;FOR PERIOD ENDING N. Materials Analysis and Inspection Methods 9. 10. 11. 12. 13. 14. Determination of Oxygen in Sodium and NaK Determination of Oxygen and Nitrogen in Lithium Determination of Trace Impurities in Sodium by Colorimetric Methods Analytical Studies of Fluoride Eutectics Determination of Oxygen in Inert Gases Composition of Diatomaceous Earth Trace Quantities of Boron by Colorimetric Methods Determination of Nickel and Nickel Oxides in Alkali Metal Hydroxides Clarity Tests of Borated Water Selutions Chemical Methods of Fluid Handling Metallographic Examination Coordination of Chemical and Analytical Data on Loops * Reaction Products of Chromium and Alkali Metal Hydroxides | ‘ ' lLead in Fluoride Eutectics and in Alkali Metal Hydroxides D. Heat Transfer 1. 1. Convection in Liquid Fuel Elements Heat Transfer in Circulating Fuel Reactor Heat Transfer Coefficients of Fluorides and Hydroxides Hgat Transfer Coefficients of Lithium Boiling Liquid-Metal Heat Transfer Sodium Heat Transfer Coefficients in Short Tubes Heat Transfer in Special Reactor Geometries Fluoride Bandling Fluoride Production DECEMBER 10, 9733-2 9733-2 9733-2 9733-2 9733-2 9733-2 9733-2 9733-2 9733-2 9201-3 2000 9733-3 9733-2 9733-2 9204-1 9204-1 9204-1 9204-1 5204-1 9204-1 9204-1 9201-3 1951 White Ross White Ross White Ross Baxter White Baxter Druschel White Brady White Ross White Ross White Baxter House Bronstein Gray Krouse Roeche Blankenship Metcalf White Baxter White Druschel Hamilton Redmond- Lynch Tobias Poppendiek Palwer - Hamilton Hof fman Lonas Claiborne Winn Farmer Harrison Claiborne Kackenmester Mann 171 ANP PROJECT QUARTERLY PROGRESS REPORT 172 2. Fluoride System Cleaning 9201-3 3. Fluoride Salvage and Disposal 9201-3 4. Preheating of Fluoride Systems 9201-3 5. Fluoride Pressure and Flow Measurements 9201-3 6. Experimental Joints for Hot Ligquid Systems 9201-3 Liquid Metal Handling 1. Equipment Cleaning Techniques 9201-3 2. Continuous and Batch Sodium Purification 9201-3 3. Sampling Techniques 9201-3 4, Sodium Vapor Trapping 9201-3 5. Liguid Metal Salvage and Disposal 9201-3 6. Liquid Metal Safety Equipment 9201-3 7. Blanket Gas Purification 9201-3 Dynamic Liquid LooODPS 1. Operation of Convection Loops 9201-3 2. High Flow-Rate Convection Loop 9201-3 3. Operation of Figure-Eight Loops 9201-3 4, Bi-Fluid Systems 9201-3 S. Fluoride Fuel Flow Transfers 9201-3 6. UO;-NaOH Slurry lLoop BMI 7. Operation of Thermal Convection Loops 2000 IV. ANALYSIS OF OTHER NUCLEAR REACTOR SYSTEMS Supercritical Water Reactor 1. Cycle Analysas 9704-1 2. Cycle Analysis NDA Helium-Cooled System 1. System Analysis NAA Na-Vapor Compressor-Jet System 1. System Analysis NAA Air-wWater {ycle 1. Survey of Air-Water Cycle 9204-1 Mann Mann Affel Coughlen Bailey Taylor Reber Wyld Mann Mann Mann Mann Devenish Mann Devenish Mann Mann Adamson Tunnell Coughlen Coughlen Wischhusen Ward Simons Cathcart Bridges Smith Fraas Cohen Gruber, etc. Schwartz, etc. Schwartz, etc. [Lane Noderer OFFICIAL USE ONLY 19_Chart of the Technical Orgaonization of ) THE AIRCRAFT NUGLEAR PROPULSION PROJECT ANP DIVISION DIRECTOR R. C. BRIANT, 8iDG. 9704.1 D. MILYER. SEC, ANP COORDINATOR ¢, B. ELLIS NOTE: This churt shews only the lines of technical coordination of L. BOND. SEC the ANP project, The verious individvals end greups of pacple listed ’ ' ’ ANP LIBRARY 5 ch and dasign which BLDG. $704.1 is coordinated fer the banefit vi the ANP projact in the monwer indis ADMINISTRATIVE ASS|STANT : ' *d coted en the thart. Eoch group, hewaver, is oluc responsible to it X group p L. M. COOK M. CARDWELL Division Dirsctar for the dutailad pregress of Its reses and for P. HARMAN® SEC M. BROWN wdministrative matrers, Peesennsi fram 13 differeny Divi of ths . . E. CARTER Oak Ridge Nationol Laboretary and Enginesring secrisns of the Y-12 §. REAGAW Plont ara Included on the chart without speciic indication of divie COORDINATING STAFF PROJECT EDITOR E. WEBSTER sionsl tines. w. B. COTVTRELL P. HARMAN® SEC. STAFF ASSISTANT ARE FROJECT CHIEF STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT FOR RADIATION DAMAGE Ww. M. BREAZEALE FOR EXPERIMENTAL ENGINEERING R, C. GRIANT FOR METALLURGY FOR CHEMISTRY €. 8. ELLIS A. J. MILLER R. WiLL!AMS,. SEC. E. S5, BETTIS ¥. D, MANLY . R. GRIMES SHIELD NG RESEARCH REACTOR PHYSICS GENERAL DESIGN GROUP RADIAT I ON DAMAGE ARE DESIGN EXPERIMENTAL ENGINEERING HEAT TRANSFER RESEARCH METALLURGY CHEM} STRY CHEMICAL ANALYSIS E. P. BLIZARD, 8LDG. 3022 W. K. ERGEM, BLDG. 9704-1 C. 9. ELLIS, BLDG. 9704-1% 6. S. BILLINGTON®, BLDW. 3025 R. W. SCHRDEDER. BLDG. 9201-3 H. W. SAVAGE, BLDG. 9201-3 H. F. POPPENDIEK. 8LDG. 9204-3 W. D. MANLY. BLDG. 2000 W. R. GRIMES, BLDG. 3733-3 C. D. SUSANG, BLDG. 9733.2 J. t. MEEM R. R. COVEYOQU A. G. H, ANDERSEN C. D. BAUMANN A. W. ALEXANDER G. M. ADAMSON® . J. M. CISAR® G. M. ADAMSON® C. J. BARTON L. J. BRADY LiD TANR N. EDMONSON E, S&. BETTIS"® W, £, BRUNDAGE /. K. BROWNING R. G. AFFEL H. €. CLAIBORNE E. S. DOMAR 4. P. BLAKELY H. P. HOUSE® C. E. CLIFFORD, BLDG. 3001 8, 1. MACAULEY. USAF J, M. CISAR"® R. M, CARROLL G. A. CRISTY" M. B, BRIDGES L. CCOPER A. D. BRASUNAS F. F. BLANKENSHIP W, K. MiLLER €, H. MILLS A. P. FRAAS® A. F. COHEN G. W. ECKERD H. R, HRONSTEIN W. S. FARMER W. H. BRIDGES R. A. BOLOMEY J. W, ROBINSON® T ;' :'-3555“ %. K. OSBORN M. E. LEE W. W. DAVIS J. Y. ESTABROOK K. E. BURMASTER D. Ot HAMILTON J. V. CATHCART b. R, CUNED R. ROWAN® ;. C. LYNN F. G. PROHAMMER® K. M. MARTIN CHESTER ELLIS B. L. GREENSTREET w. G. COBB W. B. HARRISON J. H, COOBS E, E. KETCHEN J. €, WHITE L. MARMEY ec S. TAMOR A. S. THOMPSON® J. T. HOwE® L. F, HEMPHILL €. P. COUGHLEN H. W, HOFFMAN R. B, DAY R, P. 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