2oy, RETYAERERS Y V5T ws U il g3 53149 § 3 4y5¢ ORNL- 1375 This document consists of 180 pages. Copy f?@fi of 237 copies. Series A, Contract No. W-T405-eng-26 AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT for Pericd Fnding September 10, 1952 R. C. Briant, Director J. H. Buck, Associate Director A. J. Miller, Assistant Director EDITED BY: W. B, Cottrell DATE ISSUED 0AK RIDGE NATIONAL LABORATORY Operated by CARBIDE AND CARBON CHEMICALS COMPANY A Division of Union Carbide and Carbon Corporation Post O0ffice Box P 0ak Ridge, Tennessee MARTINMARIETTA ENERGY SYSTEMS LIBRARIE i 3 445k 0353149 9 R DRNL-1375 P. Smith {(consultant) H. Snell _ I.. SBteahly Clewett Clifford - Progress INTERNAL DISTRIBUTION 1. G. M. Adamson | 40. REN. Lyon 2. R. Baldock : , fif D. Manly 3. & J. Barton B. McDonald 4. Billington £]. L. Meem 5. A. J. Miller 6. : " K. Z. Morgan 7. E. J, Murphy 8. M. F. Poppendiek 9. P, M. Reyling 10, H. W. Savage 11. R. W. Schroeder 12, Calliha¥ ~E. D. Shipley 13. Cardwel 0. Sisman 14, Center E. C. Smith - 15. Cisar L. A F. R. C. J E. - F, F. TEPTTHARVAOTEIZm EOEIEEO > T 18. Cottrell W. Stoughton 19, Cowen D. Susano 20. Eister . A. Swartout 21. Emlet (Y-12) H. Taylor 22, Ergen | Co Uffelman 23, Felbeck (C&CCC) 62. C. VonderLage 24, Fraas 63. J. M. Warde 295, . Gall 64, A, M. Weinberg 26. Graham 65. E. P. Wigner (consultant) 27. Grigorieff {co 66. H. B. Willard ‘ 28. R. Grimes 67. ~J. C, Wilson 209, Hollaender 68. C. E. Winters 30. S. Householder ANP Library 31. B. Humes (K-253 . Biology Library 32. J. Jones Central Files 33, W. Keilholtz Health Physics Library 34, P. Keim i Metallurgy Library 35. T. Kelley Reactor Experimental 36. Kertesz ;. Engineering Library 37. M. King chnical Information 38. B partment (Y-12) ™S }-.J wpmpfigpmmg>>fiépé?mér?u?nm?npbfifim?g?}mc' 39. 111 96-106, 107, 108-115. 116. 117-119, 120, 121, 122-123, 124, 125, 126-130. 131-155, 156-159, 160, 161. 162-169, 170. 171-174, 175-176. 177-179, 180. 181, 182-184, 185-188. 189, 190-191. 192-193, 194, 195, 196-197. 198, 199, 200-201, 202, 203-206. iv EXTERNAL DISTRIBUTION Argonne Natlomgl Laboratory (1 copy to Kermit Anderson) Armed Forces Sfi%c1al Weapons Project (Sandia) Atomic Energy Co ission, Wathngton Battelle Memorial Ynstitute Brookhaven Natlonal<¢aboratory Bureau of Aeronautic# (Grant) Bureau of Ships A : California Research and Development Company Chicago Patent Group % : Chief of Naval Research duPont Company i General Electric Company,AN¥P(3 copies to AF Engineering Office) General Electric Company, Righland Hanford Operations Office ; USAF-Headquarters, Office OfigAS&lStant for Atomic Energy (Atetn: Lt. Col. R, E, Greer) Idaho Operations Office (1 gopy ta,Phillips Petroleum Co.) Towa State College ¥ Knolls Atomic Power Laboratory Lockland Area Office i Los Alamos ; Massachusetts Institute Qf Technology (Banedlct) Massachusetts Institute @f Technology (Katk mann) Mound Laboratory £ National Advisory Commlfitee for Aeronautics,? A, Silverstein) £ National Advisory Committee for Aeronautics, Waghington New York Operations Office North American Av1at1qn Inc, Nuclear Development Adsociates (NDA) Patent Branch, Washington Rand Corporation (1 copy to V. G, Henning) Savannah River Opergtions Office (Augusta) Savannah River Operations Office (Wilmington) University of California Radiation Laboratory Vitro Corporation of America Westinghouse Electric Corporation g leveland (3 copies to 207-222, Wright Air Developfi? . 15 L. Hill ; Nielson ce Aircraft Corporation et Division @ opy to Pratt and Whictney 1 copy to Boeing Airplane Company 223-25fif Technical Information Service, Oak Ridgé *Tennessee vi Reports previously i1ssued in this series are as follows: ORNL-528 ORNL-629 ORNL-768 ORNL-858 ORNL.-919 ANP-60 ANP-65 ORNL-1154 ORNL-1170 ORNL-1227 ORNL-1234 ORNL.- 1294 Period Ending November 30, 1949 Period Ending February 28, 1950 Period Ending May 31, 1950 Period Ending August 31, 1950 Period Ending December 10, 1950 Period Ending March 10, 1951 Period Ending June 10, 1951 Period Ending September 10, 1951 Period Ending December 10, 1951 Period Ending March 10, 1952 Reactor Program of the Aircraft Nuclear Propulsion Project Period Ending June 10, 1952 FOREWORD ., . . . . . PART 1. SUMMARY AND INTRODUCTION . 1., e CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT Fluid Circuit Pressure-Shell Stress Analy81q' Beflector Coolant Instrumentation ® & Off-Gas System . . Reactor Control System Shielding . . . . EXPERIMENTAL REACTOR ENGINEERING Pumps . . . . .+ . Frozen-sodium-sealed pump in figure € loop & - - . TABLE REACTOR THEORY AND DESIGN . ] a OF CONTENTS a - 4 - - - Durco frozen-fluoride-sealed pump. l.aboratory-size maintained-level gas-sealed pump Durco pump with shaft packing ARE centrlfugal pump . Valves . . . e Bellows fype of valve stem seal . » » » » High- temperature valve-stem packings Sel f-welding of seat materials Canned-rotor-driven valve Heat Exchangers . - » ® - * » a " ® » * » * - Test of core element of sodium-to-air radiator Bifluid heat transfer loop . NaK-to-NaK heat exchanger Instrumentat1on . . Rotameter type of flowmeter Rotating-vane type of flowmeter . Mcore nullmatic pressure-measuring device Diaphragm pressure-measuring device Diaphragm pressure transmitters Moving-probe level indicator . Null-balance level control Strain-gage level indicator Fluid Dynamics ARE core mockup B * ’ - 12 . 13 11 12 12 i3 15 15 15 16 17 17 19 19 19 20 20 20 20 21 23 23 23 24 25 25 25 26 26 26 26 26 vil P Possible fuel-header manifold for the reactor core , Technology of Fluoride Handling ., . . . « « « + « « « & FlUOFide prOduction . . . s 2 . s . e e 8 s ¢ e Removal of fluorides from contaminated systems . . . Purity of pipe line helium . . . . « + . . . s e Fluoride plug-removal test from simulated ABE core . Injection of NaK into a flowing fluoride stream . GaS'linE‘plugging tests " s * 4 e 8 ® & v & 8 8 & » - . - Cooled-baffle vapor trap for zirconium-bearing fluoride fuel Descaling and pickling tests . . + + « + ¢ « « « + & 3. REACTOR PHYSICS . . . ¢« v ¢ v ¢ v v« 0 v o o o« o o o o Oscillations in the Circulating-fuel Aircraft Reactor . Effect of Gaps on Reactivity . . .« « ¢« & ¢« v o « « o & SlOWing'Down KETHEIS . ] . . . . » . . . . @ . . » . . 4., CRITICAL EXPERIMENTS . . . « & v v v o 4 v v 0 0« o o 4 s Direct-cycle Reactor . . + & v ¢« v ¢« « v ¢ o 0 « v o RBeflector studies « o « o o s & o & o s & o« o o o« Poison rod calibrations .+ . « o« o + &+ s o o &+ & & ARE Critical Assembly o + + v v & v & @ o o o & o « o PART 11. SHIEIDING RESEARCH . . . . SUMMARY AND INTRODUCTION . . . . + ¢« v v & o v o o o« v o« 5. LID TANK EXPERIMENTS . . . . . . « « v ¢« v ¢ o v v o & Air Ducts . . . e e e e e e e e e e e R-1 Reactor Gamma Shleldlng e e e e e e e e e e e Other Lid Tank Experiments . . . . .+ « « « « ¢« + « & Correlation of Neutron Attenuation Data « « « « o o« & 6. BULK SHIELDING REACTOR . . . . . . . « « .+ . « . . Mockup of the Divided Shield . . . . . « « « + « . . Air-scattering Experiments . , Reactor Power Determination . « + + ¢« v « + v o « o + & Irradiation of Animals . . . . .+ « & « ¢ v « 4 4 0 . Irradiation of Electronic Equipment . . . . . . . . . . Capture Gamma-Ray Measurements ., . . . . . . . . . . . 7. TOWER SHIELDING FACILITY . . . . . + + « « o v o o v o 8. NUCLEAR MEASUREMENTS . . . . s e e e e 4 s e s e e vilil Fission Cross Section of U234 and U?3%¢, . . . 29 29 29 30 31 31 32 34 34 35 36 38 40 40 40 41 42 43 45 47 48 48 53 56 63 63 64 64 64 66 69 70 70 71 71 Total Cross Section of N . . . . . . . .. Time'flf'Flight Sp&ctrometer > L] ° » . - » » [ a » PART 111. MATERIALS RESEARCH . . . SUMMARY AND INTRODUCTION . « & & o v v v 0 o v o o o o 9, CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS . . . . . . . . . Fuel Mixtures Containing UF, . . . . . . . . . . NaF-KF-ZrF, -UF, . . .« v v v o 0 o o v 0 0 o v o NaF-ZrF, UF s e e e e e e e e e e e e e e e NaF- Zr¥, BeF UE, o o s e s s e e e e e e e ZnF —UF4 s e s s s s e e sa e s s a e e e Fuel Mixtures Containing UF,. . . . . . + « + « « + « . NaF-KF-LiF-UF; . . . .« . . « v « o v o o v« o o KE-UF, . . « v o v 0 v« o o v 0w o o 0 s s v 0w s NaP'UF e e s e e e s a e e s e e a s e e e e e Preparatlon of cempounds of UP o 4 4 a4 Alkal: Fluoborate Systems . . . v ¢« « & & + & & « « o« & Vapor pressures of commercial flucborates . . . . . Thermal data for fluoborate systems . . . . . . . Differential Thermal Analysis . . . . . X-Ray Studies of Complex Plufirlde Sy%temq o 4 e s e Na ZrF -Na UF, . . .« o o o o o v o b o o 0 v 0 o NaQZrF Nd UF o & 3 e s m m e e 5 e e e s e s s e NaZrF, NaUP s e e s s e e e e e e e e s Spectrcgraphlv Analysws of Fluorldes O Simulated Funel Mixture for Cold Critical Experiment ., . Moderator Coolant Development . . . . . . « . +« « 4« 4 Coolant Development . . o . ¢« s & & @ ¢ v o o« « o o » RbF-AIF, . . . . v o v v o v e v o e e e e NaF-KF- A1F3 e b e e e s v e e s e e e a e e e NaF-ZrF,-ALF, . . . « o o o0 v 0 v 0 v e s NaF-ZrF, . . . 4 o o v s o oo v 0 o v v e e e e ZrF BeF o . ’ s e e s e e e e Hydrolybls and 0x1dat10n of Fuel MlXtUFEb ca e Heat of solution and heat of formation of uranyl fluorlde Free energy change for the reaction of uraninm tetrafluoride with oxygen . . e e e Fuel Purification Research . . . e e e e e s Purification of molten tlu011deb e e e e e s e e s Preparation of pure zirconium tetrafluoride Preparation of pure aluminum fluoride . . 71 72 73 75 76 17 77 78 - 18 79 79 79 - 81 81 - 81 - B2 - 82 82 84 85 - 85 86 86 87 87 88 89 -89 89 89 89 90 90 - 90 - 90 91 91 92 93 1x Reaction of Fuels with Alkali Metals . . + & + &« & « & + o o & o o o 93 Service Functions .+ .« o v o o t o s s s s e e 4 6 4 s 4w w4 e e 94 CORROSION RESEARCH + + + o v v v v e e e e e s s e s s e e s s s e 95 Parametric Studies of Fluoride Corrosion . . « ¢« v &+ ¢ v o o o v o« + & 96 Temperature of test . + « v v o ¢ ¢ ¢ v 4 e 0 e e e e s 4 e e e e 96 Length of test + « v v v &« o o o o o s o s o o s s o v 1 2 a2 & 4 96 Residual stresses in SPECIMEN +« « » o o o o s o &+ & o o o o s 1 o » 99 Carbon content of specimen . .+ + + ¢« « o ¢ ¢« « & o o 0 s 4 e e e 99 Corrosion inhibitoTs + « v « 4 « o o = o s o o« & o a o & s o « o » 100 Pretreatment .« o+ + « o o o o o o o o o o 0 o o 4 w0 e e o0 o0 owo. 101 Seesaw and Static Tests with Fluorides Containing ZrF, . . . . . . . . 103 Fluoride Corrosion in Thermal Convection Loops « . + « « ¢ & &« « « . » 103 Fluoride mixtures containing ZrF, . . . . . . .+ + « v « o « . . . 104 Effect of cleaning loops « « « v s 4 & o« o o o 0 o s v o+ + o« . 104 Corrosion inhibitors ..ce v ¢ 4 & & ¢ & o o o o o + o + + & + o« « . 108 Crevice COTTOSION « wsv o + o o o o o o o o & o o o o o o o o o o« 109 Temperature variations . + « + o & « o o o o « o o o o o« o o 2+« 110 Variations in alloy composition of the walls . . . . . . . . . . . 110 Postrun examination of fuels « . & « v ¢ v v v &+ o o o o & o & o . 112 Compatibility of Beryllium Oxide with Various Fluoride Mixtures . . . 112 Hydroxide COTroSion « « v o« o o o & & o o o o o & o o« & o o o+« + 4 114 e v e e e e e s 114 Temperature of test . + . + v v v v v 4 v e a0 v e Residual stresses in specimen . .« 4 + & o + & o & o o« « o+ + « « o o 114 Corrosion inhibitors . . . + v ¢ & & &« & o« o & o s &« « = o + & o« 115 Hydrogen atmosphere . + + « « v v 4 o o o o o & o o o o o v o o o o 115 Liquid Metal Corrosion . . . e e e e e e e e e e e e e e e e e s 1T Mass transfer in liguid lead e e e e e e e e e e e e e e e e 1T Effect of temperature on lead corrosion . o+ « « « & & o« o o o« « o« » 118 Lead-sodium miXtUTes .+ &« & &+ & o + o o & o s « & & o o 2 « « & + . 118 Spinner tests with sodium ., . . . . . . . . e e e e e e e .. 119 Type 1020 chromized steel in low-melting- p01nt alloys . .. 119 Compatibility of beryllium oxide with sodium, NaK, and lead « .. 119 Fundamental Corrosion Research . . . . + « & « « « &« & o o o o « « « . 121 Interaction of fluorides with structural mwetals . . . . . . . . . . 121 Synthesis of complex fluorides . +« + « « o o« & o & o o &+ o & o o o 122 EMF measurements in fused fluorides . . . « + + « « « « « &« « &+ « . 123 Polarographic studies in sodium hydroxide . . . . . « « « . . . . . 124 Electrochemistry of sodium hydroxide . . . + « ¢« & « &« « « + + o » 124 Mechanism of fluoride COTTOSION « & &« + & « & & & o o o s o + + « o 126 Tripositive nickel compounds from hydroxide corresion . . . « . . . 127 Solutions of metals in molten halides . . . « « « « .+ « « « +« « + . 130 Fluoride corrosion phenomena . .« + + + « o o« o o « o o« o o+ o « « o+ 130 11, METALLURGY AND CERAMICS . . . & . v v ¢ v 4 s e s s e e v e o o » o o« 131 Control Rod Fabrication . + + +ie v o « o v 4 ¢ & = o o o a o 2« o . 131 Mechanical Testfng of Materials . 4 o « v v 4 4 ¢ « & 4 o « o « + « » 132 Physical testing in fluorides . . . « + + « & « « o 52 v o+ » o 132 Creep and stress-rupture tests in ArgoOn . « + &+ & « » 4 + o+ » » . 133 Welding . . . e e e e e e e e e e e e e e e e e e e e e e e e w133 Cone-arc weldlng e b s s 4 e e e ah e e h e 4w a e e e e e .. 135 Specifications for ARE weldlng e e e e v e e a e e s e o a s oa ow . 135 Tests of Brazing Alloys . -+ . .« v & o & & + 4 « & & 4 « o« » « o« o » . 138 Static corrosion tests of brazed joints . . . . . . + .+ . . . . . 138 Tensile strength of brazed joints . . . . + + ¢« « ¢« + + &« +« « » . 138 Brazing of molybdenum . . « . & + © 4 & v . 4 4 v s . e . e . . . 140 Dry-hydrogen brazing furnace . . . .+ « « & & o « o o » o + « » & . 140 Reduction of Molybdenum Disulfide . . . . . . < « « . « « v « « » » ., 141 Ceramics Research . . . . « .« . v o 4 o o 4 o v v v e s s e e e .. 143 Ceramic coatings for metals . . . « & v « & + 4+ « « 2 & o« « + « . 143 Ceramic reflector . . &+ & & & =« & o 5 « 4 s s o & s o o + & v « . 143 12, HEAT TRANSFER AND PHYSTCAL PROPERTIES RESEARCH . . . . . . . . . . . . 144 Viscosity of Fuel Mixtures . . & + & « « » & o o 5 o o o s o « » » o 145 NaF-ZrF,-UF, fuels . . . . . v . .« o o o o v v v v v o v o o v . 145 Capillary viscometer . + « « v « o » v 4 & s + « & 5 «. 6 « o+ + « . 145 Thermal Conductivity of Liquids . . . « + . & & 4 « &« + « & o« o o o« o 146 Heat Capacify v = « v v 4 v v v 0 o v n o o 4 e 4 e e e e e e e e . LEB Density . . . . .o C e e e e e e e e e e e e e e el 14T Vapor Pressure of Fuel Conqtltuents P e e e e e e e e e e e e e e . 147 Zirconium tetrafluoride . . . . 4 . « o . . ¢ o 4 e e s . oa e s . 147 Zirconium-bearing fluoride mixtures .+ « .+ » + + 4 « o« 2+ . . . 147 Convective Heat Transfer in Molten Sodium Hydroxide . . . + . . . . . 148 Boiling Heat Transfer in Mercury . . soe s e b e s e . . o, 149 Natural Convection in Confined Spaces w1th Volume Heat Generation . . 149 Heat and Momentum Transfer Analysis of the Thermal Convection Loops . . 152 13. RADTATION DAMAGE . . . . s e e e e s e e e e e e e . 152 Irradiation of Fused Materlals e e s e e e e e s w e a e s w w183 In-Reactor Circulating LooPs v v v v v o« « 4+ o o o o s o 4 o o & o+ . 154 Creep Under Irradiation . . . ., . s e e e e e a e e e e 4 e s .. 154 Radiation Effects on Thermal Conduct1v1ty O X1 PART IV. APPENDIXES , . . . . . . . . . . . . 157 SUMMARY AND INTRODUCTION . & v v v v e e e e e e e e e e e N ¥ 1 14, ANALYTICAL CHEMISTRY . . . . . . .. e e v e e e e e e e e e e .. 159 Analytical Studies of Components of Fluoride Mixtures . . . . . . . . 159 Alkali metals o & ¢« & v v v 4 v 4 4 e e e e e e e e e e w e o« . . 159 ZITCONIUM v v & o & o & & & + o s & o s & o o s o o 4 o o o o« o 160 Analytical Studies of Impurities in Fluoride Mixtures . . . . . . . . 160 Chromium « & v v & o & &+ & o « o & a & s o o o o s s o o s 2 o+ « » 160 Nickel . v . v v v e e e e e e e e e e e e e e e e e e . )6] OXYEEN « v & v s & & & & & & o 4 &« 4 4 4 e 4 w4 e e e e e 161 Chloride . « « v & v 4 4 4 4 4 & 4 4 o & 4 o o 4 o o o o 2 o « . 161 Water o o o 4 v v 0 v v e e e e e e e e e e e e e e e e e e 162 Determination of Carbon in Zr¥, and Zr0,-NaF-C Mixtures . . . . . . . 162 Compatibility of Reactor Fuels and NaK . . , . . . . . . . . . . . . 162 Service Analysis + v 4 4 4 4 4 4 4 4 e 4 s s 4 4 4 4 s s e e e e o4 . 163 15, LIST OF REPORTS ISSUED . . . ¢ v v v v v v v v v s v v v s e v o v o o 164 ANP PROJECT QUARTERLY PROGRESS REPORT FOREWORD The Aircraft Nuclear Propulsion Project is comprised of some 300 technical and scientific personnel engaged in many phases of research directed toward the nuclear propulsion of aircraft., A considerable portion of this research is performed in support of other organizations participating in the national ANP effort. However, the bulk of the ANP research at ORNl. is directed toward the development of a circulating-fuel type of reactor. The nucleus of this effort is now centered upon the Aircraft Reactor Experiment ~ a 3-megawatt high- temperature prototype of a circulating-fuel reactor for the propulsion of aircrafeu, This quarterly progress report of the Aircraft Nuclear Propulsion Project at OBRNL records the technical progress of the research on the circulating-{fuel reactor and all other ANP research at the laboratory under its Contract W-7405- eng-26. The report is divided into four parts: I. Reactor Theory and Design; II. Shielding Research; III. Materials Research; and IV. Appendixes. Each part has a separate Summary and Introduction, SUMMARY AND The over-all concept and design of the Aircraft Reactor Experiment were set forth in the two preceding re- ports. The most significant modifi- cations during the past quarter have been those attendant to the specifi- cation of NaK as the reflector coolant and the NaF-ZrF, -UF, mixture as the circulating fuel. Most of the equip- ment for the experiment is now on order and some has already been received and installed in the ARE Building. The safety aspects of the reactor are being analvzed, particularly, the off-gas disposal system and the effect of a postulated fuel~tube rupture (sec. 1), Valves, pumps, and instrumentation for the fluid circuit of the Aircraft Reactor Experiment are being developed {sec. 2), Valves with both bellows seals and packed sesls have been suc~ cessfully used with molten fluorides, but the bellows seal appears to be the more reliable., Water tests with the maintained-level gas-sealed pump pro- posed for the ARE have been unsatis- factory. Since the packed seal 1s satisfactory when the back end 1is maintained at temperatures below the fluoride melting point, a pump 1n- corporating this seal has been specified for the ARE., Tnstrumentation for the ARE fluid circuits must be designed so as to be unaffected by the high vola- ti1lity and subsequent condensate of the ZrF,-containing fuel, Accordingly, flowmeters, pressure-measurement INTRODUCTION devices, and fluid-level indicators and controls are being redesigned and tested for this application, Data from the NaK-to-air radiator tests corres- pond exceptionally well with theoreti- cal values. A hydrodynamic mockup of the ARE core has served te illustrate the problems associated with fuel loading and draining. Also of signifi- cance to the ARE is ap experiment in which a fuel-tube rupture is simulated so that NaK {(the reflector coolant) is injected into the fuel stream. Theoretical investigations of the kinetics of the circulating-fuel reactor indicate that all subseguent oscillations will be moderate 1f the survives the first, short, The applicability of multigroup calculations to reactors with vastly different core and re- flector materials 1s uncertain pending critical experiments. An improved method of computing the effects of gaps on reactivity has been developed (sec. 3), reactor maximum excursion, The recent experiments on the critical assembly of the General Electric sir-cooled water-moderated reactor {R~1) have included reflector studies and poison-rod calibrations, The preliminary assembly of the circu- lating-fuel reactor for the ARE has been critical with a uranium mass in excellent agreement with that pre- dicted (sec. 4}, ANP PROJECT QUARTERLY PROGRESS REPORT 1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT E. S, Bettis R. W. Schroeder ANP Division The main core design and the general concept of the circulating-fuel air- craft experiment are essentially the same as reported 1in the preceding quarterly report.t1) The few changes made in the design and in the 1in- strumentation were necessary because of certain fuel characteristics; the most probable fuel (NaF-ZrF,-UF,) has higher vapor pressure and higher viscosity than expected, and some fluid-circuit design characteristics have had to be altered accordingly. NaK has been selected as the reflector coolant after an examination of the hazards introduced by a postulated fuel-tube rupture. Although a fuel- tube plug would almost certainly result 1f the rupture were large, studies on the simulator indicate that the reactor could be safely scrammed and drained. The off-gas system in- corporating a vapor trap and a low- temperature carbon absorber unit is being designed. Final drawings of the reactor assembly and fluid-circuit flow sheet are shown in Figs. 1 and 2. A detailed description and design draw- ings of the ARE were published in the ARE status report,(?’ FLUID CIRCUIT G. A. Cristy, ANP Division Some problems have arisen regarding the design of the fluid circuit as a result of the high viscosity of the NakF-ZrF, -UF, fuel. Since the viscosity of this circulating-fuel mixture 1s (I)H. ¥W. Schroeder, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending June 10, 1952, OBNL-1294, p. 9. (Z)W. B. Cottrell, Reactor Pragram of the Air- craft Nuclear Propulsion Project, ORNL-1234 (June 2, 1852), twice as high as originally anticipated, 1t caused a reduction of one-half in the Reynold’'s number within the heat exchanger tube system, that is, from approximately 8000 to approximately 4000, This, in turn, has caused a reduction 1n the tube-side heat- transfer coefficient, such that the calculated minimum film temperatures are in the vicinity of the freezing point of the fuel mixture (940°F), In order to re-establish the minimum film temperature sufficiently above the freezing point, two major changes were made. First, the operating temperature range was raised from between 1150 and 1500°F to between 1200 and 1500°F, which i1ncreased the flow rate by a factor of 7/6 and increased the minimum bulk fuel temperature by 50°F. Also, the heat exchanger tube arrangement was modified., In contrast to the previous arrangement with five tubes in parallel, the present arrangement has all the fluid passing through two tubes in parallel and thence through a second bank of three tubes 1in parallel. By incorperating these two changes, the minimum calculated film temperature was 1increased to about 1050°F. The changes also increased the system pressure drop and the pressure level within the pressure shell., The consequences of these differences are discussed in the following section,. The gas-sealed pump originally designed for use with Na or NaK has been found to be unusable for fluoride circulation without extensive modifz1- cation of the seal. Since the surge tank design is 1ntimately related to that of the pump, details on the surge tank will not be released prior to resolution of the pump problem, All other details of the fluid circuits FOR PERIOD ENDING SEPTEMBER REGUL.ATING ROD ASSEMBLY . \ SAFETY ROD GUIDE SLEEVE SAFETY ROD ASSEMBLY e 1952 D!! HA36 10, e e=TURBE EXTENSION /,,,-—THERMAL SHIELD CAP -~ THERMAL SHIELD TOP ~—FUEL INLET MANIFOLD T - RTINS g : ’1 ey ; S ' ; .‘.\?\&;. E 2 N, A NS N \ B W ¥ N AR i TR, | Bk PP IP DI AT './77>>" R |- il [ e g AT R .~\\\\ffl\ SN &k\:\ RN AR THERMOCQUPLE LAYOU gy s S T CAP I i | . X s Y N e e T y SRR LS s e P il 2 2t . A ».V%figfi,ixfp%gé iy > - TOP HEADER L N X S TOP TUBE SHEET CORE ASSEMBLY —— _ RN e \\* N ~—— ‘ AN 3 - 13 7 5 . Fe Z \ 7 i 7 R TN REFLECTOR COOLANT kX < By f§% AN T HEATERS TUBES --meeees ~—-§: e g N R vl N7 A b e i N~ Be0 MODERATOR ) B i/ B P B 3 AR AND REFLECTOR FUEL TUBES R R ¢ SR i VAN RS R N 2N AN .~ THERMAL SHIELD w 7 ool b : AN £ | NZEN SRR B U : F% VS ASSEMBLY - PRESSURE SHELL-—tZaad DRSS Y N R AR E A N7 7N i 1 B AR o - A 7:'741 ; 3 BN % < E ’/E‘}{ ;/. N “—(; N8 ,’j,;§42% N i AN N7 LIREC [ > N 37 CANRA Y R b 3 R SR B A B il ¥ ARy _moTTOM TuBE NN s bt 7 - SHEET o N o ;.Z ; z /4/ Q\\ R /".‘_\ N ! iy /f N ’N L\ © AR AN e STUD N e 7R S £od / L4 e }\f 3 S o - > . e | gk = VAR —-5UPPORT ASSEMBLY BOTTOM HEADER : . ‘ j SR ; el FTERGE L SR FUEL QUTLET i N, || e || e il \fiéf QI MANIFOLD NS e s R A IR 5 -‘ :]? ST I'I’. o \S\\\/‘?-)> >?f 1 ; } R Bt % __/24/‘_ 4_/_( SRR SRR BIORRNR ‘/ L i .‘/f A A }i‘ LA ‘ N Sl N . T A RS = &> g ) NN & ARl e ~THERMAL SHIELD THERMAL SHIELD CAP-— 27 BOTTOM r L 1l o £ X ifl\‘ M ol ) Fig. 1. are proceeding as rapidly as drafting man power permits. The fuel-injection system has been completely detailed, and the egquipment is now being fabri- cated in the local shops. The bellows valves for use in the fluoride circuits are being supplied from an outside vendor. It was found that the welding of these units could be done better by SCALE IN INGRES Experimental geactor (Elevation Section). OBNI, and arrangements for doing this welding for the vendor were completed. STRESS ANALYSIS J. W, Walker ANP Division PRESSURE -SHELL B. L. Maxwell Consultants, For a pressure of 100 psi in the pressure shell, the maximum bending T oaRacN ' v 2-in. GATE VALVE - €-in, WATER SUPFLY "IA f fl"\ owg, 16337 pm—— o EXISTING EQUIPMENT SHOWN DOTTED e )_, o \ { / ! e — ——— ( T { * ~in. ©B r—+ RESERVOIR ‘l;i————r—“—fi_iflflp's_ _________ T 1 .J_\ ‘I» = ‘ . [ fox w L8 — N Y 4in. 2 L ) ! o [ N T - / GLYCOL SURGE TANK v SR . X G-in. GATE VALVE ¥ Gan 1\4/1 1 23 cfi 2 ; Ej 2 *0\ FROM Ny SUPPL F ' ’_Ll;__l T o vawves; ; ; G135 - L2 - i Ly -in PIPE . ) i ! ' ' ¥ o= 8 i3 i 8 [ T r 5 : : L =3 i ! g7 18ie 1005 ! & | } : | ! - n | o 1 w o ' pwalE oa i | o ! PRESSURE : LgiE -ae |§ a e E -~ i SWHTCH 1 Tapono [ Gix (oD o | :, | 5 _} iy Lede G : ! L= | Z —;F gfi RADIATION | MONITOR | ‘t VACUUM b puMe | ] COLD TRAP y VACUUM PUMP s ! . ' : L W ! ! i : : ‘ : 5D : i < ; ' : : . | ! — 7 .._,A_,I, f 5 : : : \JFILTER i X A MAINTENANCE LMAINTENANCE DRAIN ! DRAIN P = I @ ! ! T ' DLl u 8,900 § L . Z Yo | : : Wt : g s N | v x _ . /750 2 = i He ol T ol > = - @ N | ‘ z o i TO DRAIN 5 3 3- 2 < ‘ '.‘)1 A f: TC DRAIN I_f! . FILL i i | T - ‘ rie LINE . : . : ho_ = ¥ : i TR & 5 F!LLX\Q'D FILL ( CO r:lflL‘ ' ‘ ( 1 VT LINE ™ LINE [vT NE ‘ ; x T /{"q o | i o u ~| 7 i YD) | FILL ARD FLUSH = P P P 4 [l TANK I | ! 1) i i [ | : i He K @ @ @ | ; > 2 | FLUSH TAMK : FILL AND DUMP | FILL AND DUMP y FLUSH TANK 3 @ i M (e s : TANK i TANK : ! : N i ‘ : | @ ‘ - | { ‘ 1 ~ | & . _} -~ ; : : o) : | [ . ____. 5 oo J P! ‘ i . He '/, MAKEUP HEADER _—— i ! RESERVE MANIFOLD - . NOTES: e - | He SCRUBBER | O T O FLOW IN gpm \:] FLOW 1M cfm —(F y—— FUEL FLOW CONDITIONS ARF BASED ON THE ASSUMPTION THAT L j ! J‘W E r = ] 7 B S INERT SALT THE COOLANT HAS THE FOLLOWING PROPERTIES: T L ‘ P i ] PRESSURE IN psig | fl PRESSURE N INCHES OF WATLR (9 b= 475 /143 FROM He P _ gy HELIUM L= 8 TO 4G centipoises AT OPERATING CONDITIONS O SUPPLY TRAILER ij TEMPERATURE IN °F C WATER Cp=10.28 Bru/lb- F o GLYCOL VOLUME OF MAIN SYSTEM [APPROX )i J— TERNA : 423 - Vfl;’ VAPOR TRAP INTERNAL: 1.5 1 EXTERNAL: 5.0 #° TOTAL: 75117 Fig. 2. Fluid-Circuit Flow Sneet. o stress 1n the heads of the shell would be 8500 psi. At a temperature of 1150°F this gives a creep rate of about 1% in 10,000 hr and a factor of safety of 1.6 based on stress-rupture in 1000 hours. By plate reinforcing, the maximum bending stresses are reduced to 6000 psi. This 1s slightly less than the stress for a creep rate of 0,10% in 10,000 hours. As stated in the previous report, all holes that occur 1in areas of high stress are re- inforced. With added reinforcing, the stresses in the side walls have also been reduced to a maximum axial stress 1in the cylinder of about 6000 psi. With- out added reinforcing, the stress at 1006 psi is 7430 psi. The reinforced side walls give a creep rate somewhat in excess of 0,10% in 10,000 hr and a factor of safety of 1.8 based on stress-rupture in 1000 hours. REFLECTOR COOLANT E., S. Bettis, ANP Division The fuel-carrier NaF-ZrF, was originally considered as the most probable choice for the reflector coolant., However, recent dynamic tests of the corrosion of the beryllium oxide reflector by a similar fluoride mixture, NaF-KF-ZrF,, showed that the beryllium oxide was dissolved to an extent that would prohibit the use of that fluoride as a reflector coolant, Two alternative secondary coolants have been considered: NaK and the fluoride eutectic NaF-BeF,. The choice between these two coolants resolves to an analysis of the hazards attendant to a fuel-tube rupture that would connect the fuel and reflector-coolant circuits, Although i1t 1s unlikely that a rupture will occur and allow the coolant te leak into the core, the possibility of such a leak must be admitted., Furthermore, if such a leak does occur, the reactor must and will be shut down no matter what reflector coolant is used. The fluoride eutectic NaF-BeF, has acceptable physical characteristics and 1s compatible with both the beryllium oxide reflector-moderator and the fluoride fuel. An objection to the use of this coolant 1s the expense and time involved in producing a sufficient quantaity for the ABE. Another objection, which would net be present 1f NaK were used, emerges from the consequences of a fuel-tube leak. With the relatively inert NaF-BeF, mixture as the reflector coolant, in the case of a fuel-tube leak there is the possibility that sufficient fuel could seep into the moderator inter- stices to overcome all contrel rod effects. NaK, however, is known to reduce fluorides such as ZrF, and UF, to lower valence states of the metal. Since these reduced compounds have higher melting points than the original material, a fuel-tube plug could readily occur. That this would actually happen with a sufficiently Jarge rupture has been demonstrated experimentally (cf., sec. 2, “Experi- mental Engineering’ ), The reactor system simulator was used to study the effect of a leak of NaK into the fuel channel of such magnitude that the channel 1s com- pletely plugged. When this occurs the temperature of the plugged fuel passage begins to rise at a rate of about 36°F/sec and will rise to an asymptotic value of about BOO°F above the fuel temperature at the time of the plug- ging., This condition obtains by the temperature coefficient alone 1f no control rod correction is employed. When the plugged tube begins to heat the reactor goes on a 25-sec negative period. The outlet temperature of the 11 ANP PROJECT QUARTERLY PROGRESS REPORT five unplugged passages drops at the rate of about 15°F per second. This drop in temperature of five out of six outlet fuel tubes provides a unique symptom of plugging that will enable the reactor to be shut down by scram- ming and subsequent draining of fuel. Since there appears to be some additional safety to be realized from the use of NaK and NaK has the most favorable physical properties of the coolants considered, 1t will be used as the reflector coolant in the ARE. The secondary heat exchanger and associated gear are being designed for use with this coolant. INSTRUMENTATION S. A. Hluchan, Instrument Department Some ARE instrumentation has had to be modified because of the high vapor pressure of the ZrF, component of the fuel mixture. Sufficient ZrF, con- densate would accumulate in the gas space above the hot liquid to com- pletely clog the open instrument lines in a few hours i1f the liquid tempera- ture was around 1200°F, the instrument lines contemplated for use 1in liquid level, pressure, flow had to be replaced by devices with no open lines. Bellows were Consequently, previously measuring and liquid substituted as pressure indicators and a fluid-immersed inductance type of instrument was substituted for measur- ing flowand liquid level. A discussion of these instruments appears ain the ‘Experimental Engineering’’ section of this report (sec. 2). OFF-GAS SYSTEM T. Boseberry, ANP Division A carbon absorber unit at liquad nitrogen temperature 1s being used to completely remove the fission gas 12 (Br,, I,, Kr, Xe) from the helium stream coming from the NaK vapor trap before it is released to the stack. The unit consists of four 4-im.-IPS3S stainless steel tubes connected 1n series by 2-in.-IPS stainless steel tubes. The first and second tubes cool the gas stream and solidify the Br, and I,, and the third and fourth tubes, which contain approximately 700 in,? of 20-28 mesh CXA carbon, absorb Kr and Xe. With normal off-gassing (unit designed for 10 fe*/hr; probably the flow will be less and not continuous), the carbon bed has a break-through time of at least 800 hours., If it becomes necessary to purify the helium atmos- phere in the heat exchanger pit be- cause of a hot leak, each of the two units can take a flow rate of 100 ft? /hr for at least 70 hours. ' The break-through time of the unit was obtained as follows:(?) 1 A% t”? = 2.30 + 26.1<-~-~> Q where [} weight of carbon (g), flow rate of helium (cc/min), o~ i time of break-through (hr),. REACTORE CONTROL SYSTEM E. P. Epler Research Director’s Division The reactor control system 1s essentially the same as outlined in previous reports. (%) Detailed drawings (B)T. S. McMitllan and W, L. Johnson, Pissolver Off-Gas Processing, ORNL-1309 {(to be published). (4)R. ¥. Schroeder, op. cit., ORNL-1294, p. 12; R, W. Schroeder and E. S. Bettis, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending Marck 16, 1952, ORNL-1227, p. 10, FOR have now been completed for the ionization~-chamber liner-and-shield installation and for the ionization- chamber gas system. Layout and wiring diagrams for the relay cabinet are being prepared. The construction of ionization chambers, preamplifiers, amplifier cabinets, relay cabinets, contral console, and recorder panels by the Radiation Counter lLaboratories is on schedule for October delivery. Components for a simplified servo control have been assembled for testing with the analog computor. The regulat- ing rod speed has been established at 0.1% Ak/k per second, with a total of 0.4% Ak/k in a total rod travel of 12 inches. The pilot model of the high-temper- ature fission chamber 1s ready for test. Two additional assemblies are being constructed with completion ex- pected in October. SHIELDING H. I_Ji P‘I Enlund, Physics Division The adequacy of the shielding and the anticipated radiation levels in PERIOD ENDING SEPTEMBER 10, 1952 various components of the ARE are being studied. To date nothing untoward has been brought to light. The servicing of equipment in the heat exchanger room will be governed more by radiation that comes directly from activated equipment and fission- product deposition than that which comes through the concrete wall from the fuel dump tank or the shut-down reactor. The rate of deposition and the retention of fission products on Inconel surfaces are not readily pre- dictable, except by experiment. Tt would appear that satisfactory removal of fission products may be possible 1f methods analogous to these for the decontamination of solid-fuel process- ing equipment were used. An upper limit to the activation of the Inconel by thermalized delayed neutrons has been calculated by assum- ing all are captured in the Inconel. The cold trap in the helium scrubber system is being investigated for the heat generation caused by the retention of radiocactive xenon and krypton on the activated carbon,. 2. EXPERIMENTAL REACTOR ENGINEERING H. W, Savage, ANP Division Continued emphasis by the ex- perimental engineering group on the development of components for high- temperature fluid-circulating systems has resulted in improvements in pumps, valves, seals, and instruments such as those for indicating and controlling liquid levels, measuring system pressures, and measuring flow rates. Valves for use in high-temperature fluid systems, such as the ARE, must operate with complete reliabiiity for long periods over a temperature range of from 1200 to 1500°F. The stem seal 1s considered the most critical problem in the operation of valves at high temperatures; conventional sealing materials are inadequate. Consequently, the major effert of the valve progranm has been on seal development, and progress has been made in sealing valves with both Inconel bellows seals and stuffing-box seals packed with 13 ANP PROJECT QUARTERLY PROGRESS REPORT Inconel braid and graphite and nickel powders. Some time has been spent, also, in considering several valve configurations and 1n determining whether self-welding of valve seating materials may be encountered in the presence of the ABRE fuel at operating Lemperatures. Seals are also the critical problem in pumps. Gas seals, packed seals, frozen seals, and combination packed and frozen seals have been tested with varylng degrees of success. The combination of packed and frozen seals appears to furnish the most positive sealing action against high-temperature fluoride fuel mixtures. The hydro- dynamic tests of the first model (Mcdel DA) of the ARE pump, with water as the circulated fluid, indicated that the pump sealing arrangement had to be modified., Two avenues of modification were then open: a sump type of pump could be uwused in which the fluid 1s not 1n contact with the shaft seal but is separated by an inert gas volume,or a packed and frozen seal could be used. The latter was selected for use in the ARE. combination Heat transfer tests have continued with sodium-to-air radiators with varying fin spacing. The test results correlate with the changes in fin spacing made 1n the radiator configu- rations tested and are in excellent agreement with the theoretical heat transfer. A test loop is near com- pletion to make measurements of over- all heat transfer coefficients of fluorides to liquid metals. Instrumentation for the ARE cannot incorporate gas-sensing lines that contact the liquid fluoride surface because of the high-vapor pressure of the zircouium fluoride. Accordingly, flowmeters, pressure-measuring devices, and fluid-level 1ndicators and controls are being designed with closed liguid 14 surfaces. Two types of high-tempera- ture fluid flowmeters are being developed. A rotameter type of flowmeter incorporating a tapered iron core and an i1nduction coil as the sensing element is being tested. A rotating-vane flowmeter that uses a permanent magnet and a pick-up coil as the sensing element 1s being prepared for test. Several pressure- measuring devices are being developed. These are divided i1nto two general types in which the sensing element, either bellows or diaphragm, operates in a cooled, trapped-gas volume or is completely submerged in the high- temperature fluird. A float type of level indicator and controller that uses a tapered iron core and induction coil to sense and control levels has been developed that will reliably main- tain two levels within 0.1 in. of each other when used in connection with a venturi for flow measurement. Other level indicators are also being developed. Hydrodynamic tests have been con- ducted with a glass mockup of the ARE core. With 45-gpm flow 1t was demonstrated that the system could be filled with certainty when evacuated to a pressure of 28 1in. Hg, but under o circumstances can all the fluid be removed from the core structure, Later tests i1ndicate that with 50-gpm flow,vacuum is not required for com- plete filling. A possible reactor header ar- rangement has been developed that gives an optimum flow pattern through a reactor core that has single-pass, parallel fuel tubes. greatly reduces the amount of stagnant fuel that would be held up in the header as compared with that held up in the conventional plenum chamber. This design The technology of fluoride handling has progressed. Fuel production equipment capable of meeting current FOR fuel needs is 1in operation. The equipment for producing the ARE fuel and moderator coolant is being designed. Methods have been developed for cleaning fluoride-contaminated systems with high-pressure steam jets. Tests were conducted 1in an attempt to determine whether a plug of solid fluoride fuel mixture could be remelted in the ARE core-tube bends without rupturing the tubes. Repeated cycles of increasing severity failed to rupture the tube under test, but a slight swelling was noted in the bend at the conclusion of the experaiment, Fuel-flow stoppage resulted when rapid additions of NaK, which 1s one of the proposed ARE moderator-coolants, were made to a stream of fuel, NaF- Zr¥,-UF,, at 1500°F in a small, forced- circulation loop. This test simulated a large fuel tube rupture that, 1f developed in the ARE fuel tubes, would permit the moderator coolant to leak into the fuel system. The vapor pressure of NaF-ZrF, -UF, (46.0-50,0-4.0mole %) is high compared with other fuel mixtures considered, and at ARE operating temperatures some Zr¥, 1s sublimed from any {free sur - faces. This sublimed material plugs cool control-gas lines in dynamic fluid systems and makes their operation difficult. The vapor pressure problem with this fuel 1s being studied, and equipment that may alleviate the problem is being developed, PUMPS Frozen-Sodium-Sealed Pump in Figure 8 Loop (W. R. Huntley, ANP Division). The Worthite frozen- sodium-sealed pump test first re- ported in ORNL-1154(!’ was terminated (l)fi’. B, McDonald, Afrcraft Nuclear Propulsioen Project Quarterly Progress Reporit for Period Ending Septeaber 10, 1951, OBNL-1154, p. 21. PERIOD ENDING SEPTEMBER 10, 1952 after more than 4000 hr of pumping sodium at temperatures from 800 to 1200°F. During the major portion of the test the sodium temperature at the pump was approximately 1100°F. A weld failure and a simultaneous flange failure in the test loop caused the termination of the test. : Shaft speed during the greater part of the test was 2000 rpm. Pump operation was smooth during the test except for a short period during which there was interference between the impeller and the pump bhousing. An inspection of the pump shaft after disassembly showed the shaft to be in excellent condition except for a build-up of nickel on the shaft at the so0lid sodium-liquid interface. The build-up of nickel was caused by mass transfer from the nickel sealing ring at the parting faces to the cold region of the seal. Although this build-up caused no operational diffi- culty, it indicates that nickel waill not be entirely satisfactory for use as a sealing gasket for parting faces for high~temperature pumps that are to be operated for extended periods. The loop is to be disassembled for complete examination and oxygen analysis of the sodium in the system and in the cold stub that was placed in the loop for trapping oxides from the system. Durco Frozen-Fluoride-Sealed Pump (W. B. McDonald, W. G. Cobb, P. G. Smith, ANP Division). The Durco centrifugal pump (Model H34MDVX-80) incorporating an improved design of the frozen-fluoride seal has been constructed and i1s being assembled into a test loop. Previous tests of a frozen-fluoride seal on a similar pump, which operated for over 500 hr at 1200°F, indicated that such a seal will furnish a positive sealing action against high-temperature fluorides. Failure occurred at the parting-face 15 ANP PROJECT QUARTERLY PROGRESS REPORT seal and some scoring of the shaft occurred in the sealing area. The operational characteristics of the seal are considered to be the most perplexing problem encountered. The pump in the first test could not be stopped for periods longer than 5 min without considerable difficulty in restarting. Upon stopping, the frozen fluorides in the seal bonded the shaft tightly to the sleeve and 1t was necessary to apply sufficient heat to melt the fluorides before operation could be resumed. The improved design incorporates a 2000-watt Calrod heater by means of which the heat input for melting of the fluorides in the seal can be carefully controlled so that the pump can be more easily restarted after shutdown. The cooling fins on this seal are baffled so that the air flow across these seals can be ac- curately controlled. This degree of close control over the temperature at which the seal is operated should enable a better determination of its optimum operating conditions. The shaft for this pump is coated with Stellite No. 3, which 1s harder than the Stellite No, 6 used in the previous test, to reduce shaft wear and scoring. The weakest point of the pump is the high-compression sealing joint at the parting faces; nevertheless, it 1s expected that this pump can be operated with the fluoride temperature approaching 1500°F. Laboratory-Size Maintained~-Level Gas-Sealed Pump (W. G. Cobb, P. W. Taylor, G. D. Whitman, ANP Division). The redesigned gas-sealed pump re- ported previously®?) has bheen con- structed and placed in operation. The (2)y, g, Cobb, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending June 10, 1952, ORNL-1294, p. 17. 16 redesign consisted of the following modifications: 1. The ring-joint parting-face seal was moved from the region in which 1t was in contact with high- temperature fluorides to a higher position where it now seals against helium only. 2. Heat-radiation baffles were placed between the high-temperature liquid-gas interface and the top flange of the pump. 3. The stuffing-box shaft seal was replaced by a carbon-ring hardened- tool-steel face seal. The fluoride fuel NaF-ZrF,-UF, has been pumped for over 250 hr at 1200°F with this pump. The flow rate 1is approximately 10 gpm, the shaft speed is 2500 rpm, and a 16-ft head of fluid is developed. Some operational diffi- culties have been experienced which, although not severe, indicate that some further minor design changes are needed. Turbulence of the free liquid surface i1n the pump 1is rather violent and prevents true level indication by the spark-plug probes. The shaft speed is limited to between 2500 to 3000 rpm because of this disturbance. Some o0il leakage has occurred from the bearing housing past the carbon-ring face seal and/or the '0”-ring oil seal 1nto the o0il catch basin im- mediately below. Although this basin 1s drained continuously and no o1l 1is permitted to drain into the fluorides, the system 1s probably contaminated by o1l vapors. The extent of this contamination will be determined when the pump is shut down and analysis of the fluorides made. With only the above difficulties, which are not considered to be serious, this pump continues to operate very smoothly and 1ts success to date 1s very en- couraging. FOR PERIOD ENDING SEPTEMBER 10, 1952 Durco Pump with Shaft Packing (W. R. Huntley, P. W, Taylor, H. H. Johnson, ANP Division). Althoug packed seals that will operate satis- factoerily have been developed for high-temperature pump shafts, such seals must either be tightened at all times to the point where shaft wear 1is likely to Tesult or maintained at a temperature below the freezing point of the circulated fluid. During the operation of a standard Durco pump that was wodified to incorporate a stuffing-box seal packed with Inconel braid and nickel and graphite powders, it was found that the pump could be operated for long periods with zero leakage of the stuffing-box seal only when the back end of the seal was permitted to operate at a temperature below the melting point of the fluid pumped. When only the metallic packing is at a temperature above the melting point of the fluid, a small amount of fluid penetrates the packing inte the cold regilion near the compression member, where 1t freezes to seal the pump (leakage rate approximately 1 g of solid material in 24 hr). This pump has operated for 450 hr, pumping the fluoride fuel NaF-Zr¥F,-UF, at temperatures from 1150°F to 1300°F. The flow rate 1s 16 gpm, and an ap- proximately 35-ft fluid head 1s de- veloped at a shaft speed of 1500 rpm. This sealing method was derived because when the entire stuffing box, including the compression member, was heated to a temperature substantially above the melting point of the fluid, considerable leakage at a rapid rate resulted. When the heat source was remocved from the back end of the seal, the leakage i1mmediately stopped and the pump continued to operate smoothly without leakage. Operation of this Durco pump with shaft packingis smoother than operation o0f the Durco pump with the frozen “seal. When the pump operating power 1s measured, power surges of ap- proximately 0.5 kw and about 1/2-min duration are found to peccur at intervals of 2 to 3 minutes. Similar power surges occurred in operation with the frozen seal that contained no packing, but they were greater in magnitude because of the longer frozen section. The amount of wear on the shaft, which 1s hard-surfaced with Stellite, cannot be determined until the pump 1s shut down for examination; however, the pump operation is sufficiently smooth te indicate that the shaft wear may not be severe. Two lubricants (tricresyl phosphate-molybdenum disulphide mixture and lead borate glass) have been introduced into the seal in an attempt to eliminate the power surges. Either lubricant reduces the amplitude of the power surges temporarily but does not reduce their frequency,. ' ARE Centrifugal Pump (W. G. Cobb, A. G. Grindell, G. D. Whitman, ANP Division). The Model DA gas-sealed pump originally designed for the ARE has undergone hydraulic tests 1in a closed loop with water as the circu- lated fluid. As designed, a combination surge tank, de-gasser, and pump fluid- level contrel has been placed in the loop at the pump discharge. Figure 3 shows a schematic outline of this circuit. The tank carries full flow and 1n addition is connected to the pump with liguid and gasequalizer lines. The liquid connection leads from the bottom of the surge tank, enters the pump above the 1impeller housing, and feeds a narvow, annular volume surrounding the shaft to form a liquid seal for the impeller shafrt. The gas equalizer interconnects the two gas volumes in the pump and surge tank. ' Successful pumping has not been obtained except when the gas equalizer is valved shut. With this inter- connecting line open as designed, the 17 ANP PROJECT QUARTERLY PROGRESS REPORT GAS SOURCE MODEL DA CENTRIFUGAL PUMP7 Teo—VENT UNCLASSIFIED OwG, 16338 DEGASSER -AIR SPACE TO GAS CONNECTION SURGE -DEGASSER TANK H ’-)/ VENTURI T [ ADJUSTABLE VARIABLE - SPEED DRIVE SUCTION TO GAS SPACE CONNECTION DEGASSER-AIR SPACE TO LIQUID CONNECTION THROTTLING VALVE DRAIN VALVE ~FILL AND DRAIN VALVES L G | Fig. 3. pump loses 1ts prime and, as observed in a glass pipe section in the pump discharge, the flow is almost wholly gas. With the gas equalizer closed, but with the liquid equalizer either or closed, satisfactory pump performance data have been obtained. The performance data are presented in Fig. 4. open LLiquid-level detection in the pump has been attempted by means of a sight glass, and satisfactory indicated levels have been maintained with separate gas-pressure supplies on the pump and surge tank through a wide range of flows. However, small, rapid, pressure perturbations in the system communicable to the pump will cause the liquid level to rise or fall unduly 1in the shaft annulus without these changes being indicated on the pump-level sight glass. Thus over- filling of the pump bedy or gassing of the liquid being pumped can occur. 18 el e Loop for Water Test of ARE Pump. UNCLASSIFIED DWG. 16339 00 100 - o o e am ] 50 HEAD (ft OF Hy0) FRICIENCY (%) o 0 .~ —_— o 0 50 100 150 FLOW (gpm) Fig. 4. Water-Test Performance Data for ARE Pump. FOR PERIOD ENDING SEPTEMBER 10, 1952 Either overfilling or gassing would result in flow stoppage if the fluid involved was a fluoride. Although 1t may be possible to render perturbations of this type harmless by some combination of stationary vanes 1n part of the annulus, by increased flow (or de- creased head loss) from the surge tank through the liquid equalizer line, or by packing dampers 1in the annulus, the development work required appears to be considerable. In view of the requirements of the ARE {for high mechanical certainty of operation and the progress on packed seals, further development work on the Model DA pump has been postponed indefinitely. Modifications for using the same impeller and housing in both the sump type of pump and packed-seal pump are being carried out. VALVES Bellows Type of Valve-Stem Seal (P. W. Taylor, ANP Division). One of the most promising valve-stem seals appears to be that constructed of three-ply Inconel bellows, each ply in the bellows being approximately 0.009 in. thick. Results of tests of two such bellows have been so highly satisfactory that this type of seal has been incorporated in the design of 1 1/2-in. throttling valves for use in the ARE system and in pump test loops. One test was terminated for ex- amination of the bellows after operation for 1000 hr submerged in a bath of fluorides at 1500°F, The bellows was cycled once each 12 hr through 1ts maximum rated travel of 9/32 inch. Upon examination it was found that only the first ply of the bellows has been penetrated by the fluorides. A second such bellows has been operated under similar conditioens for over 1600 hr with no indication of failure. This bellows has been cycled 127 times during the test, which will continue until failure. : High-Temperature Valve-Stem Packings (D. R. Ward, H. B, Johnson, ANP Division; R. N. Mason, Engineering and Maintenance Division). A program for investigating high-temperature valve- stem packing materials is under way. Considerable operating experience has been accumulated with valves packed with Incenel and graphite powder and lubricated with tricresyl phosphate- molybdenum disulfide mixture. Such valves have proved quite successful in laboratory operation; however, some stem seizure has been encountered that required heating of the valve bonnet to free the stem, and in some instances leakage of fluoride around the stem has occurred. At the present time reliability of a valve packed in this manner is not completely assured for long~time maintenance-iree operation with fluorides at 1200 to 1500°F. It is possible that powdered packing materials that are not wetted by the molten fluorides may give better stem sealing and better valve operation. A test apparatus has been constructed to investigate the wettability of variocus powdered materials by the fluorides, and tests are presently being con- ducted. Experience to date with high~temper- ature packing materials indicates that in almost every instance stuffing boxes have required retightening after reaching operating temperature. For example, with powdered-graphite packing a tightening motion of 0.030 in. per inch of packing length 1is required Lo re-establish the original degree of compression after the packing temperature has been elevated to 1100°F. A packing-expansion testing 2 19 ANP PROJECT QUARTERLY PROGRESS REPORT rig has been constructed and the expansion rates of various packing materials will be investigated in the search for a powdered material that will show minimum dimensional change upon being heated. Tricresyl phosphate has proved valuable as an antigalling lubricant. The application of this fluid teo threads during assembly has permitted greater ease of disassembly after valves have operated for many hours at 1500°F, Self-%¥elding of Seat Materials (D. R. Ward, ANP Division; R. N. Mason, Engineering and Maintenance Division). Preliminary tests were conducted to determine whether self- welding 1s encountered in the presence of high-temperature fluorides with the following combinations of materials: type 315 vs. type 316 stainless steel, type 316 stainless steel vs. Inconel, and Inconel vs. Inconel. Test specimens were clamped tightly together and submerged i1n fluorides for 187 hr at 1300°F. Although visual examination showed no evidence of self-welding in any of the specimens, metallographic examinatlon showed some evidence of interdiffusion of the specimens of similar materials. Investigation of self-welding is being continued. Cauned-Rotor-Drivea Valve (W, B. McDonald, A. L. Southern, ANP Division). Preliminary tests of the driving elements of the canned-rotor-driven valve(?®? indicate that an operating mechanism can be built that will operate such a valve satisfactorily. A design of a canned-rotor-driven valve has been completed, but the priority given to the bellows valve hasdelayed construction. The favorable performance of the three-ply Inconel bellows indicates that a canned-rotor type of drive may not be required. (3)!7. B. McbBonald and A. L. Southera, op. cit., ORNL- 1294, p. 20. 20 HEAT EXCHANGERS Test of Core Element of Sodium-to- Air Radiator (G, D. Whitman, A. P. Fraas, M. E. LaVerne, ANP Division}. The second performance and endurance test of the core element of the sodium-to-alr radiator was initiated on June 6, 1952, and 200 hr of operation was logged before failure. The average sodium inlet temperature during the test was 1200°F, and a maximum temper- ature of 1600°F was held for 74 hr before failure. As 1n the first radiator core test, failure occurred in a tube between the top fin and header on the sodium inlet si1de of the radiator. Metallographic examination of both core elements indicated failure at a brazed tube-to-header joint because of local porosity in the joint. In both cases the failure was accelerated by the attack of the sodium on the outer surface of the tube. When the second test had been 1in progress (2 hr a building power-supply failure occurred and the sodium froze 1in the radiator before 1t could be drained. To resume the test, sufficient preheat was obtained through normal methods to restart sodium flow without difficulty. FElectric strip heaters were used on the outside of the air duct. The second radiator core element contained Nicrobrazed Joints and was identical in construction to the first except that the fin spacing was altered to give 15 fins per inch instead of 10.5 fins per inch as on the first core element. Some fabrication difficulties have been encountered in producing these radiators, Since the furnace at Y-12 is too small to accommodate brazing of air radiators for the turbojet project, an attempt was made to FOR PERIOD ENDING SEPTEMBER 10, 1952 es tablxsh Jarger furnace located im the rolling mill at X-10. The second core 1s an excellent example of successful brazingof an intricate design; however, to complete the radiator as planned, the header caps had to be heliarc welded in place. From the results of several uwnsuccessful attempts to do this it may be concluded that the heliarc method does not produce a in the presence the techmigue i1in a new and sound welded joint of Nicrobraz metal. Heat transfer data from the first and second cores (10.5 and 15 {fins per inch, respectively) were plotted as Nusselt number vs. Reynolds number. When the physical properties of air at the mean temperature of the free calculate these stregm were used tao parameters, the points were segregated on a temperature basis with a large spread between the data taken with sodium temperatures of about 500°F and those taken with sodium temperatures of 1500°F. Since NACA RM-No. EBLO3 recommends basing physical properties of the air on the boundary layer rather than on the mean free-streanm temperature, the data were replotted. (The mean sodium temperature in the radiator was taken as being ecquivalent to the ailtr temperature in the boundary layer.) As v Fig. 5, this gave excelleat correlation of the data, it should be noted that 1in calculating these data full allowances were made for variations in fin efficiency resulting from variations in both heat transfer coefficient and thermal conductivity of the fins. shown The air-pressure-drop data were correlated on the basis of the mean temperature of the free air | As shewn in Fig. 5, this gave good correlation of the data for each of the two [1n spacings tested. stLream. A third radiator with interrupted ting has been assembled and brazed and 1s now being tested. The fins have been cut through and slightly offset every 2 in. {about 50 hydraulic radii) in the direction of the air flow. Preliminary data show that interruption of the fins vields an increase in heat transfer coefficient of approximately 30% depending on the Reynolds number. The concomitant increase in pressure drop is only 10%. Not only do the interrupted fins give an increased heat transfer coefficient, but 1t appears that they make the heat transfer coefficient almost independent of fin temperature at a given air-weight flow rate instead of falling off rapidly with an increase in fin temperature, as was found to be the case for the plain fins. The use of pure wnickel in place of stainless a fin material should produce asubstantial 1mprovement 1n fin efficiency because i1ts thermal conductivity is nearly three times as high. A ceramic coating 1is desirable, however, to protect the nickel {from oxidationif 1t 1s used at temperatures steel as above about 1200°F, The ceramics group have developed a suitable coating (mainly chrome oxide) and are able to get pood adherence 1f the nickel 1= first bright-annealed ina wet-hydrogen atmosphere. A test on a radiator of this type 1s plarnned as priorities permit. soon as Bifluid Heat Tramnsfer Loop (D. ¥, Salmon, ANP Division). Fabricatien of components for a lnop te measure heat transfer from flucride fuel mixtures to NaK has been completed, and the components awalting assembly (Fig. 6). are The tube side of the double-pipe heat exchanger has been calibrated with water. The friction factor was determined over a range of Reynolds numbers 3,000 to 90,000, @ An average value of 0.032 was obtained. The calibration of pressure drop vs. from 21 ANP PROJECT QUARTERLY PROGRESS REPORT S I FIN SPACING RS . Na INLET [__FIN SP i TEMPERATURE (°F) | 0.5 per in. | 15 per in 700 | i | | & 900 1oo ; 13C0 ‘ 1500 1600 & O * = h0 s 10 [— | 5, | | 27 —_ I 1ol 0% R, Fig. 5. 0.010 in. to center line square spacing. thick; stainless steel fins, flow will be used as a secondary method of measuring tube-side flow rate. Venturis for flow measurement 1n both primary and secondary systems have been calibrated with water. A calculated prediction of the heat transfer performance of the system has been made to determine operating characteristics. The original calculations for the loop were based on the fluoride mixture NalF-KF-TLaF (11.5-42,0-46.5 mole %}). The fluoride fuel NaK-Zr¥F,-UF, (46-50-4 mole %), 22 Performance of Core Element of Sodium-to-Air Radiator. DWG 163540 | — L | 4 — + - 5 ! | | o . i i : T ! P i | L S S S | | | b | w 5 Type 304 tubes on 2/3-in. center line 3/16-1in. which has approximately 100% greater density and viscosity, will now be used for tests aund, consequently, the range of operation will be reduced. The pressure drop with the NaK-Zr¢F,- UF, fuel will be greater for the same flow rate than with the NaF-KF-LiF mixture, and as a result the operating pressures for the system will be higher. The maximum flow rate probably will be reduced so that the maximum Reynolds number will be 1in the order of 30,000 instead of 80,000 as origi- nally reported. FOR PERIOD ENDIING SEPTEMBER 10, SURGE TANK ~=--— FLUORIDE SYSTEM e _IQUID METAL SYSTEM 1952 UNCLASSIFIED WG 1834 ELECTROMAGNETIC FLOWMETER e COOLER Lt HEAT EXCHANGER ames \ )“/ FILTER~ ] | -~ CENTRIFUGAL PUMP % Mt — e e e A i I i (f”'_“ L e e e~ U o N e e l HEATER SECTION }Z l 1 ( wmwaL{:}~h_L ~u4 AIR o VENTURI SLOWER e - FLOWMETER VENTURI FLOWMETER SUMP };] ~ Fig. 6. NaK-to-NaK Heat Exchanger (M. E. Il.aVerne, A. P. Fraas, ANP Division; E. E. Hoffman, Metallurgy Division). Metallographic examination of the ction of the NaK-to-NaK heat ex- changer described i1n the previous quarterly reports®*’ has been completed. Two tube failures were discovered in the Nicrobrazed tube-to-header joint that were not detectable from sepa- ration. It was concluded that these failures were caused by the vibration in the system and the etching effect the circulating NaK had on the tubes in the hot zones. These failures were encouraged by the embritctlement of the areas near the joint by the brazing alloy and the large grain size of the 16-mil-wall, type 304 stainless steel tubes. No fractures were detected in the 19.5-mil-wall, type 347 stainless steel, Nicrobrazed, tube-to-header loints. Upeon sectioning the heat exchanger, 1t was found that the hot (inlet) end was very bright, whereas the cold (434 p, Fraas, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending March 10, 1952, ORANL-1227, p. 32; M. E. LaVerne and A. P. Fraas, op. cit., OBNL 1294, p. 22, . EL&CTROMAGNE? IC X g:@dmp Schematic Diagram of Bifluid Loop. {outlet) end was covered with a dark, powdery film. Spectrographic ex- amination of the dark and bright surfaces revealed the only detectable difference to be a manganese content present in the dark (cold) surface that was two to three times the ex- pected value for type 347 stainless steel. The x-ray powder pattern of the dark powder scraped from the surface of a tube 1n the cold zone showed Ni0O, Cr, and type 347 scainless steel base material. INSTRUMENTATION RBotameter Type of Flowmeter (P. G. Smith, A. L. Southern, ANP Division). A rotameter type of flow-measuring device has been designed for measuring the flow of a fluoride stream at 1200 teo 1500°F. This instrument consists of a conventional rotameter float operating in a tapered barrel, The float position is indicated by a tapered iron core, chrome plated to resist corrosion, which operates in a cylinder of static fluorides extending downward from the flowmeter (Fig. 7). An inductance coil that is capable of 23 ANP PROJECT QUARTERLY PROGRESS REPORT rfl~/13fig|nREF ’~—m~l¥fflw- | | 134 in DRILL THROUGH '1| n. ONE WA LL~W~~\:S:fi 3/8 n. | [ X 3/, -in-1PS SCH. 40 PIPE 29 ¥4~ in-LONG - | i l —~0.250-in-D % 1¥%-in-LONG PIN UNCLASSIFIED CWG. DSK- 157314 Iy i, e LAVA COil. FCRM ( FIRED AFTER MACHINING) 5V2ll’|,’§ n ; _L-, TR RN 7 / / / L34 inTHICK X 3% in-D ROTAMETER FLOAT INCONEL PLATE TAPERED BARREL Fig. 7. being operated at a constant tempera- ture of 1300°F is placed arocund this cylinderin the vicinity of the tapered core, and as the core moves 1n the field of the inductance co1l, the movement 1s sensed by suitable 1n- strumentation. The core movement may be calibrated to very accurately indicate the flow rate of the fluid through the rotameter barrel. The initial test of this instrument 1in a high-temperature dynamic fluid system was somewhat disappointing. Thermal distortion of the cylinder in which the tapered core 1s located resulted in binding of the moving parts and preventing further operation of the instrument. A new design, which should eliminate this difficulty, has been completed that allows greater clearance between the tapered core and the containing cylinder and requires stress relieving of the parts before assembly. The new instrument 1is being constructed and will soon be 1incorporated in a test loop. BRotating-Vane Type of Flowmeter (W. B. McDonald, J. M. Trummel, ANP Division; F. A. Anderson, Research A rotating-vane type manufactured by the Participant). of flowmeter 24 Yg=in-THICK X 3 ¥,-in-D PLATE L 3-in-IPS SCH. 40 PIPE 2 Va-in-LONG 3 -in-D x 12-in-LONG INCONEL BAR Yy -in-THICK % 1 3%4-in-D INCONEL PLATE —-- “TAPERED CORE (#g~in-Dx 15-in- ‘\ LONG CHROME~ PLATED MILD \\ STEEL BAR) INDUCTANCE COIL Rotameter Type of Flowmeter with Variable Induciance Core. Potter Instrument Company and modified for expected high-temperature appli- cation has been calibrated on water and 1s awaiting test i1n a high-tempera- ture fluid system. This instrument consists of a permanent magnet rotated in the fluad stream by a small turbine. A pick-up colil 1s placed outside the flowmeter in close proximity to the rotating This co1l, when connected to a Hewlett-Packardelectrical tachometer through a preamplifier furnishing 60 db of amplification, will very accurately sense the rate of rotation of the permanent magnet. At room temperature this arrangement was calibrated in terms of accurate flow measurement with a pick-up co1l located as much as 3 i1n. from the magnet. It 1s expected that this instrument can be operated at high temperatures with the pick-up coil located as close as 1/2 in. to the magnet. The limi- tation of this flowmeter is expected to be the curie point of the permanent magnet, although some operational difficulty may be encountered with the Carboloy bearings on which the magnet and the rotating vanes Operation of this instrument wi1ll be examined at fluid temperatures to between 1250 and 1300°F. magnet, are mounted. FOR Moore Nullmatic Pressure~Measuring Device (P. W. Taylor, ANP Division). The design of the Moore pressure transmitter, reported previously, (%) has been modified to prevent the high- temperature fluid from contacting the pressure-sensing bellows at high pressures. The modified instrument is rated at 0 to 60 psi at 1200°F maximum temperature. The temperature of the fluid stream at the point of pressure determination may be as high as 1500°F. This instrument incorporates the trapped-gas principle used in the previous design and must be operated in a vertical position, Further designs are under way that will in- corporate three-ply Inconel bellows. The trapped-gas space will be eliminated and the bellows pressure-sensing element will be completely submerged in the high-temperature fluid. Such an instrument may be located 1n any pesition and at any point 1in the system. ' Diaphragm Pressure-Measuring Device (p. W. Taylor, ANP Division). The diaphragm pressure-measuring device, Fig. 8, utilizes a linear differential (5)p. W. Taylor, op. cit., OBNL-1294, p. 26. PRESSUREROD*\\ I's Y Wt Yai - Yain D . 2% in. ‘/2 in. \‘“DIFFERENTIAL TRANSFORMER SPRING BLOCK I-*3-'hint-J T PERIOD ENDING SEPTEMBER 10, 1952 transformer to measure the deflection of a 1/16-in.~-thick Inconel diaphragm. The deflection will be calibrated 1in terms of the applied pressure. This instrument, which is being procured from an outside vendor, incorporates the trapped-gas~-space principle; however, test results indicate that the high-temperature fluid may be 1in contact with the diaphragm without resulting instrument failure. The instrument is rated at 0 to 60 psi at a maximum diaphragm deflection of 0.010 inch. The operating temperature is limited to 1000°F at the diaphragm to minimize the creep rate of the Inconel; however, by using a riser leg to the pot and diaphragm and by cooling either the trapped gas or the fluid te this temperature, the instru- ment should accurately measure the pressure of a fluid stream at 1500°F. Diaphragm Pressure Transmitters (P. W. Taylor, ANP Division). Six pneumatic null-balance pressure transmitters are being fabricated. These pressure transmitters will be constructed of Inconel and will have 0.014-1in. Inconel X diaphragms. The instrument is rated at 0 to 100 psi at 1000°F; however, since the pressures UNCLASSIFIED DWG. DSK~15608A r\" DIAPHRAGM HOUSING N, \ Yy~in. ORIFICE %mfi Mein‘ 3"15"" ¥gin. D 7-’fain. —2in. —+— FLUCRIDE STREAM 134hgin. WELD NCONEL OIAPHRAGHM Fig. 8. Diaphragm Pressure-~Measuring Device. 25 ANP PROJECT QUARTERLY PROGRESS REPORT on opposing sides of the diaphragm are balanced, 1t may be run at higher temperatures without sacrifice of accuracy. mitters will be tested completely filled with high-temperature fluid. too great a These trans- Moving-Probe Level Indicator (P. W. Taylor, A. L. Southern, ANP Division}. A level indicator has been designed that incorporates a conventional spark-plug probe moving through a high-temperature packed seal. For the first test the probe level 1is controlled manually by means of a lead screw mechanism., If the packed proves to be an effective gas seal and 1f the first test 1is en- couraging, an automatic control for the probe will be aincorporated into the design. seal Null-Balamce Level Control (P. W. Taylor, A. L. Southern, ANP Division). Fluid flow in high-temperature systems is presently measured by means of a venturi with trapped-gas pressure pots to sense the pressures of the venturi entrance and throat. Unequal levels in the pressure pots result in con- siderable flow measurement error, particularly with fluids of high density. A null-balance level con- troller (Fig. 9) has been developed that equalizes the level 1n the two pressure pots and greatly 1ncreases the accuracy of flow measurement. This level controller consists of a tapered core operating in the field of an induction co1l. The core 1is attached to a float that rides on the free of the fluid 1n the pressure pots. This instrument will maintain the levels 1n the pots within 0.1 in. of each other. surface In operation, when flow 1s started in the system, fluid rises 1in the venturl entrance and throat pressure pots, with the fluid rising to a higher level in the high pressure pot. The level device causes the 26 voltage output of circuit No. 1 to be greater than that of circuit No. 2. This voltage differential appears partially across variable resistor (A) and 1s fed to a Brown controller. The signal in the controller operates a limit switch to open a solenoid valve, which bleeds gas into the high pressure pot and forces the level down. When the levels equalize, the voltages in the two circuits buck each other and the signal from the con- troller drops to zero. The solenoid valve then closes and shuts off the gas supply. This sequence 1s repeated at any time the levels in the two pots become unequal. Strain-Gage Level Imdicator (A. L. Southern, ANP Division). A level indicator 1s being designed that incorporates a three-ply Inconel bellows to transmit pressures exerted by rising levels to a high-temperature strain gage. This device will be limited to use in systems pressurized within the limits of the bellows. A questionable feature, which must be tested, is the long-time reliability of high-temperature strain gages, The advantage of such an i1nstrument is that the bellows sensing element 1is completely submerged in the high- temperature fluid and 1s not subject to failure or loss of calibration as a result of solid deposits sublimed from the free surface of fuels having a high vapor pressure, such as the fuels containing zirconium fluoride. FLUID DYNAMICS ARE Core Mockup (L. A. Mann, ANP Division). A full-scale mockup of the fuel passages of the ARE core, as reported and illustrated previously, (%) was completed and tested for flow (6)1.. A. Mann and D. F. ORNL-1294, p. 28, Fig. 10. Salmon, op. cit., FOR PERIOD ENDING SEPTEMBER 10, 1952 UNGLASSIFIED OWG 16342 250 n IRON CORE 250 N o TVVVA A ~ AAAN~ D _ b H D \ P FLOAT HIGH LOW CIRCUIT PRESSURE PRESSURE CIRCUIT NO. 1 NO. 2 . SOLENOID - GAS CONTROLS BROWN CONTROLLER Fig. 9. Null-Balance Level cController. 27 ANP PROJECT QUARTERLY PROGRESS REPORT characteristics, ease of filling, and ease of emptying. The tubes were assembled from l1-in. glass tubing and bends with flanged joints, and the headers were constructed of copper sheet and tubing. A centrifugal pump, a rotameter, a vacuum pump, and a 50-gal reservoir were the major pieces of auxiliary equipment. Both water and a zinc chloride solution were circulated 1in this mockup. The zinc chloride was used to simulate the fluorides because of 1ts density (1.7 g/cc) and viscosity (10 cp) at the operating temperature {(90°F) of the mockup. The operating charac- teristics of the mockup on both water and zinc chloride are given in Table 1. Table 1 PRESSURE AND FLOW IN THE ARK CORE MOCKUP TAP WATER | ZnCl, SOLUTION Pressure at pump (psi) 20 38 Pressure at core entrance (psi) 10 18 Rotameter reading (gpm) 43 >50 Evacuation of the core prior to filling was a definite aid, as 1in- dicated by the percentage of tests 1in which all six parallel circuits were filled completely with zinc chloride solution at three core pressure levels (Table 2). When the core was first evacuated to 28 1n. Hg and the bottom header then opened to the liquid (ZnCl, solution) supply with the vacuum pump continuing to operate, all tubes filled almost completely. When the vacuum connection was shut off, the 28 Table 2 PERCENTAGE OF TESTS IN WHICH CORE FILLED COMPLETELY TESTS FILLED COMPLETELY CORE PRESSURE (%) Atmospheric 80 26 in. Hg vacuum 95 28 in, Hg vacuum 100 top header opened, and the pump started, the remaining gas was swept out and the core filled completely with liquid. In all tests with zinc chloride solution, many small bubbles remained in suspension in the liquid throughout all test runs, with some runs being as long as 20 min; however, this egquipment did not utilize a surge and degassing tank as incorporated in the ARE design. All attempts to completely empty the core by blowing out the fluid with air failed. When 20-psig air was applied to the top header an estimated 15% of the liquid remained in the tubes, unevenly divided hetween circuits, The largest rematning volumes were approximately 18 1in. (of tube length) in some tubes. in each test one or two tubes emptied almost completely, and it appears that no reasonable gas pressure will remove all the fluid from all circuits, Tests of evenness of flow in the six parallel circuits, with both water and zinc chloride solution, were made by suddenly adding ink to the fluid reservoir and observing the relative flow rate through each of the six circuilts. With water, no difference in flow rate was observed, With zinc chloride solution, one circuit lagged 2 to 3 sec behind the others. It 1s planned to install FOR PERIOD ENDING SEPTEMBER 10, greater pump and rotameter capacity - for making further tests at 80- to 100~gpm flow rates. Surge tanks and a water mockup of the ARE fluid circuit are being constructed that will be combined with the core mockup for hydrodynamic tests of the combined systems., Possible Fuel-Header Manifold for the Reactor Core (W. C. Tunnell, ANP Division)., A possible reactor core for ANP use would comprise upper and lower fuel headers joined by fuel tubes in parallel. Such an arrangement has the advantage of complete drain- ability and allows some increased flexibility 1n the external fuel circuit, The major problem of design 1is reducing the header volume to avoid large, relatively stagnant volumes of fuel 1n regions adjacent to the high- neutron flux of the core or in regions subjected to delayed-neutron effects, and at the same time providing for optimum flow conditions im all the parallel fuel tubes. Cptimum flow in a right-cylinder reactor core will provide uniform flow in all tubes 1f the reactor flux is uniform throughout the core volume or the flow conditions vary inversely with the flux. For purposes of investigation, a half-scale plastic model of the bottom half of the ARE core was built in which all of the tubes were headered. This was tested with water at flow rates up to 125 gpm. The header design started as two plates. One of the plates served as the tube header sheet and had grooves milled in 1t to provide communication to all tubes, and the other plate provided the tube sheet closure and had a circular milled groove in open communication with the network of grooves 1in the tube sheet. No location 1952 of inlet with respect to the tube- sheet closure passage was f{ound that resulted in optimum flow. After a series of tests it was found that the insertion between the original tube and closure sheets of a third plate with appropriately po- sitioned small holes, like a salt shaker, and a thin annular plenum chamber in the closure sheet over all the holes in the middle added plate resulted in optimum flow conditions. It is also apparent from the tests that variations 1in the flow pattern can be obtained by variations in the arrangement of holes in the middle plate. 3 As a point of interest, the volume of fuel 1n an ARE-size header and supply pipes would be approximately 1942 in.?, as compared to 2780 in.?3 of fuel within the tubes. TECHNOLOGY OF FLUORIDE HANDLING . Fluoride Production (G. Nessle, Materials Chemistry Division). The design of equipment capable of producing 3 kg of treated fuel was completed July 1. Briefly, the treatment consists of bubbling hydrogen through molten fuel for 1 hr followed by hydrogen fluoride for 2 hr at 1550°F. The initial heating and melting of the fuel mix is done under a hydreogen fluoride atmosphere. The equipment has operated with only one failure, which was due to faulty welding, since being placed in operation on July 14. Following the success of the 3-kg- batch equipment, the design was scaled up for construction of equipment to produce 25 to 30 kg per batch (50 to 75 1b). Unlike the smaller apparatus, the 25-kg equipment requiresa receiver 29 ANP PROJECT QUARTERLY PROGRESS REPORT furnace that can be lowered away from the receiver to facilitate removal of the filled receiver without disturbing the remainder of the apparatus. Since some zirconium tetrafluoride (around 50 g in 5 hr) 1s evolved during the treatment of the fluorides at 1500°F, a trap containing copper servo has been introduced to prevent clogging of the hydrogen fluoride and helium purge lines and damage to the valves 1in those lines. The trap signal, which is sized according to the size of the production equipment,has been quite satisfactory and requires only minor attention. Calrod heaters were used to heat the 3/8-1in.-0D nickel tubing transfer line, The treated fuel is transferred from the treatment reactor to the receiver by 1inert gas pressure. Construction of this apparatus began July 16 and the first run was suc- cessfully made on July 25. At the end of August the total production was 128.3 kg of treated fuel. Table 3 gives a break-down of the production in the two units, Although the equipment, which 1s capable of producing 60 kg per week 1f needed, 1s not operating at maximum capacity, production 1s well ahead of the demand. Equipment 1s now being designed to produce the mixture NaF-ZrF,-UF, (46-50-4 mole %) in the quantities required to meet ARE fuel requirements. The production of fluoride fuel is dependent upon the production of sublimed ZrF,. Arrangements have been made to obtain sufficient hafnium- bearing ZrF, to permit production of fluoride fuel for experimental needs. It i1s now possible to produce 300 1lb of sublimed ZrF, per week starting from the base material ZrCl,. adequate 30 Table 3 FLUORIDE PRODUCTION SMALL APPARATUS LARGE APPARATUS (3 kg) (25 kg) 1.869 21. 054 1.970 21.509 1.837 20,300 2,003 26.900 1.984 27.138 1.770 11.433 kg 116.901 ke Every batch of ZrF, fluoride produced 1s now being crushed and sampled prior to use in the fuel mixture. The Zr¥F, thus far obtained has been of very good quality, but it 1s grayish in color because of entrained carbon. After September, hafnium-free ZrF, will be available for the production of ARE fuel. Removal of ¥luorides from Comn- taminated Systems (L. A. Mann, ANP Division). High-pressure (140-psig) steam jets are used successfully far removing fluorides from contaminated systems. Heavy fluoride deposits are removed completely without damage to the equipment. It has been found desirable to select a steam-jet size compatible with the size of the equipment to be cleaned so that high- velocity steam will always impinge directly on the deposits to be re- moved, and there will be sufficient clearance for the spent steam to escape from the container. 1In several instances this method has reduced time of cleaning from 48 to 4 or 5 hours. In other cases parts have been cleaned FOR PERIOD ENDING SEPTEMBER 10, by this method that could not be economically cleaned by methods used previously. When systems or components are contaminated with a thin fluoride coating, steam jets furnish no par- ticular advantage over washing with water and scrubbing with a wire brush. Purity of Pipe Line Helium (L. A, Mann, ANP Division). Installation of the helium line direct to the tank car located outside the building is complete. After the lines were purged the oxygen content of the helium was reduced to a level not detectable by analyzing equipment sensitive to 1 ppm. Helium of this high purity 1is now distributed to all major points of usage in the experimental engineering areas of Building 9201-3, and the requirements for bottled helium and argon have been greatly reduced. ' Fluoride Plug-Removal Test from Simulated ARE Core (L. A. Mann, ANP Division). Some operational experience indicated that it would be very difficult to remelt a plug of frozen fluoride in a pipe line without rupturing the pipe. Since the ARE core as presently designed cannot be completely drained, tests have been coniducted to determine whether the residual fluorides can be remelted without damaging the core structure, A type 316 stainless steel tube bent to simulate an ARE core U-~tube was filled with NaF-ZrF,-UF, fuel, which was then permitted to freeze. The assembly was placed in a high-tempera- ture pot furnace and cycled three times at a heating rate of 1R°F/hr between 850 and 950°F. No failure occurred. More fuel was then added te the tube and insulation was placed arcound the upper legs of the tube to simulate the beryllium oxide in the ARE core. The assembly was again taken through eight temperature cycles, and again no failure resulted. 1952 Additional cycles over the same temper- ature range were then made at in- creased heating rates, with the final cycle being made at full furnace power. Again no failure resulted. Final examination of the tube showed that it had expanded 0.030 in. at one point and. 0.050 in. at another point. These tests did not entirely duplicate ARE conditions but were probably just as severe. Injection of NaK into a Flowing Fluoride Stream (L. A. Mann, ANP Division; F. F. Blankenship, Materials Chemistry Division; T. N, McVay, Consultant, Metallurgy Division). NaK is one of the coolaunts being considered for the ARE reflector and moderator. Since the fuel tubes will be immersed in the coolant, i1t 1is necessary to determine the extent of the NaK-fluoride reaction that would result from a fuel-tube rupcure. A test apparatus to determine the extent of this effect consisted of a small, forced-circulation locp with an Fastern centrifugal pump to circulate fluoride fuel, Nab-ZrF, -UF, (46-50-4 mele %), at 1500°F and a flow rate of approximately 3 ft/sec. The effect of both slow and rapid injection of NaK into the fluid stream will be measured. A NaK injection unit was attached to this loop and when the temperature of both the fluorides and the NaK reached 1500°F, & quantity of NaK (approximately 5% of the volume of the fuel in the system) was injected into the stream. After the injection the pump was permitted to operate for approximately 6 min, at which time all heat was turned off, the pump was stopped, and the entire assembly was rotated through 90 deg to a horizontal position and permitted Lo freeze. According to the ‘exothermal re- actions Na + UF, ——® NaF + UF; and Na + ZrF,———®= NaZrF,, heat evolution 31 ANP PROJECT QUARTERLY PROGRESS REPORT was expected and did occur. This was indicated by an immediate tempera- ture rise of 200°F that was measured by a thermocouple immediately above the NaK injection point. Fxamination of the system showed incomplete mixing of the NaK and the fuel and indicated that the system was almost completely plugged by the reaction products., Possibly flow had stopped completely. This test in- dicates that large fissure type of break) occurred in an ARE core fuel tube during actual operation and permitted a similar NakK injection, the flow 1n the tube would be completely stopped or considerably slowed down. The fuel residence time in the tube would thus be increased to the point where very high temperatures would result and perhaps bring about complete rupture of the tube. The possibility that the plug would heal the leak and prevent the flow of fuel into the interstices of the moderator 1s still a matter of speculation. The loop was sectioned into seven large pieces and most of the pieces were subsequently subdivided as in- dicated in Fig. 10. Also, the general appearance and the complexity of the products observed are suggested in Fig. 10. The extent and distribution of the altered fuel suggests that complete or nearly complete plugging of the loop occurred very soon after the NaK was injected. Material in the surge tank had been attacked omly slightly and even less reaction had occurred on the side of the loop opposite the injection point. The plugging and consequent lack of circulation were, of course, responsible for the system being very far fromchemical equilibrium and for the large differences observed between this experiment and the static systems described in the section on 32 1if such a failure (a - “Chemistry of High-Temperature Liquids” (sec. 9). Samples of material from each section of this loophave been examined by x-ray diffraction, chemical analysis, and the petrographic microscape. Since the material found varies con- siderably from the center to the edge of each section, as well as along the axis of the pipe, complete characteri- zation of all the material has proved virtually impeossible. Such phases as the crystal solution of NaUF, 1in NaZrF,, Na,ZrF., UF,, and numerous complex compounds of KF occur 1n abundance. In addition, metallic zirconium occurs 1in sections 3B and 4 just above the NaK injection point. At least two additional phases, which are unknown, are recognized 1in considerable quantity in this loop. One 1s the dark-brown phase that is pleochroic and has an average index of refraction at about 1.556. The other 1s an orange-red phase generally associated with, and perhaps formed from, the brown phase. This mineral has an average 1ndex of refraction of about 1.588, with low birefringence, and i1s probably monclinie. It 1s probably a compound of UF,, since some crystals are available that seem to indicate transitions from this red phase to pure UF,. A second test will be conducted during which NaK wi1ll be injected into the fuel stream at a much slower rate to determine the effects of a very slow leak. It 1is thought that fuel circulating at highvelocity will perhaps sweep the reaction products from a very slow leak out of the reactor core. Gas-Line-Plugging Tests (W, B. McDonald, P. W, Taylor, ANP Division). The vapor pressure of the fluoride fuel NaF-ZrF,-UF, (46-50-4 mole %) at FOR PERIOD ENDING SEPTEMBER 10, 1952 E.40U, .74 colg Hy , 37.6Zr, 41.9 Zr 349K LEMON-YELLOW PHASES SIMILAR TO NaUFg - NaZrFg SOLUTION, AND TANK NapZrfg . TRACE UFy, A. DWG. 16343 ‘SURGE 68 6 A SEE &8 A oot e - _ CRYSTALS ON WALL: SOME T _ 5 NaUFg -~ NaZrfg , Nap ZrFg OR _ 17 K AMALOGS. LARGE CRYSTALS zZ MUGH A 5 OF UF3. 19.3U, 9.65 cc/g Hy , 7.8 U, 1.22 cc/g Ho = i | TRACE UFg = ! o 2 5 T 30.2 Zr, 37.4F, 228K 36.9Zr, 41.6F, 237 K— = | |5 : & T }-NGZPF5(?) . CONSIDERABLE B BO2U, 140 cerg Hy, 2 tA ; Ns - ' wflj 369 7r, 41.4F, 22 K —— O 3 ol gy o »alar, ez & ~1 TV—VERTWCAL VEIN OF A, PLUS CRYSTAL 0 : e e e - ~ SOLUTION OF NaZrFg~KZrFs SOME L 1 ngSFB’ Neatrts Ky ZrFg OR NOEZPFES ” 2 1B 3 3c MUGH DARK BLUE-GRAY TO PURPLE, - MOSTLY UFz, GRAINS SHOWING A-=UFj b L 6410 245 onre 1 gae TRANSITION. SOME NaZrfs, SMALL M . cc o - - 2 SOME A AND B8 i eI AMOUNT Zr METAL. VERY SMALL AMOUNT ~ lig] MN2irfe 2 2n, SREF, 32tk T YELLOW PHASE ' " NG UF 1) % 35| — Nar 8.18U,1.28 ec/q H2 ) g e ) , . 381520,4L4F,255K'””/\ e EARLY AGRA 38a MAINLY NaZrFg, NO GREEN COLOR N APPEARING FUELL e 8.30 U, .35 ccrg H,, 38.9 2 4B F, 054K e B CHANGING: TO A NagyZrfg OR KpZrfy SOME A, UFy BLEACHED NaZrfg -MNalifg MUCH A, B AND A, NO UFg ~0.98K, 5.90 U, 1.60 cc/g vi,,40.2 Zr, 42 4F A= RED-ORANGE UNKNOWN, AVG n = 1.588; PROBARBLY MONOGLINIC, B= DARK BROWN UNKNOWN , PLEOCHRQOIGC, AVG n = 1.556 . MAY CHANGE GRADUALLY TO A AS REDUCTION PROCEEDS SINGE SOME GRYSTALS ARE BROWN ON ONE END AND RED ON THE OTHER Fig. 10, Fluoride Loop. 950°F, or slightly above, 1s 0.016 mm. At 1250°F it is 1 mm and at 1500°F 1t i3 13.26 mm. Tests were conducted to determine the rate at which gas lines are plugged by sublimed ZrF, rising from the free surface of this fluoride fuel at various temperatures. Gas lines ranging from 1/8 to 1/4 in. in Distribution of Reaction Products from Injection of MaK inteo diameter were connected toc a pot containing the fuel, which was heated to 1500°F. High-purity helium was introduced into the pot, swept across the free surface, and exhausted threough the gas lines at relatively slow flow rates. All lines plugged solid in less than 100 hr of operation. L (o8] ANP PROJECT QUARTERLY PROGRESS REPORT The fluoride temperature was reduced to 1050°F and a second test was started under conditions corre- sponding to the first test, This test has logged more than 1300 hr with no evidence of line plugging. Cooled-Baffle Vapor Trap for Zirconium-Bearing Fluoride Fuel (W, B. McDonald, J. M. Trummel, ANP Division; F. A. Anderson, Hesearch Participant). A cooled-baffle vapor trap was de- veloped to condense and collect the ZrV, vapor that evolves from the Na¥F-ZrF,~UF, fuel. The vapor trap consisted of a 3-in.-diameter pipe 24 in. long containing 28 baffles spaced at 3/4-in. intervals, with the openings in the baffles alternated from one side of the pipe to the other. A cooling coil through which water was circulated was placed around the complete length of the baffled pipe. High-purity helium was passed across the free surface of the 1500°F fuel and exhausted through the Laffled pipe. Filter paper was placed at the exhaust port to collect any solid particles not trapped by the baffles, The flow rate of the helium across the free surface was 15 cfh. approximately After slightly more than 50 hr of operation the baffled pipe was so completely plugged by the material sublimed from the fuel that greatly increased pressure was required to maintain the specified gas flow rate. This pressure 1ncrease resulted in the bulging of the top and bottom of the flat fuel container to such an extent that the test was terminated. Ex- amination showed that the baffle pipe was completely plugged by a hard crystalline deposit at the point of entry into the fuel container. The remainder of the pipe contained heavy deposits of fine, powdery material. 34 This test indicates that a baffled vapor trap 1s cntirely umnsatisfactory for the preventionof gas-line plugging. Other types of liquid vaper traps are presently being tested. One of these tests consists of passing the gas from the free surface of zirconmium-bearing fuel through a bath of Nak-KF-LiF (11.5-42.0-46.5 mole %) at 1500°F, and there 1s a similar test in which from zirconium-bearing fuel are passed through a bath of NaK at 1500°F, These tests are still in operation and the results have not yet been determined. the vapors Descaling and Pickling Tests (D), C. Vreeland, E. E. Hoffman, R. B. Day, L. D. Dyer, Metallurgy Division). In connection with the problem of removing welding scale from ARE components some tests have been run to check different methods of de-~ scaling and pickling oxidized Inconel. Sodium and NaK were mentioned as being two of the most promising descalers (from a handling viewpoint) to be tested. Oxidized Inconel specimens were treated, 1n corrosion test tubing, for various times and temperatures with sodium and with NaK. The results of these tests are best summarized bv reference to Fig. 11. It is quite apparent that in these tests neither sodium nor NaK efficient as an oxide scale remover until the test temperature was railsed to approximately 800°C. 1In order to determine whether a dynamic NaK system would be any more efficient as a scale remover, an oxldized Inconel specimen fastened i1nto the bottom of an Inconel tube by crimping. The tube was then half filled with NaK, sealed, and run in the rocker furnace for 4 hr at approximately 700°C. The oxide scale was not removedby this treatment and so, apparently, the temperature of test is the deciding factor in scale removal. static was was FOR PERIOD ENDING SEPTEMBER 10, 1952 UNCLASSIFIED PHOTO Y-68993 INCONEL, AS OXIDIZED Na Lhr AT 200°C ~ {br AT 300°C fhr AT 500°C 5 hr AT 600°C Ne 10 hr AT 700°C 3 hr AT 800°C NgK 2 hr AT 300°C 2 hr AT 500°C 2 hr AT 700°C 2 hr AT 800°C Fig. 11. Bescaling Properties of Na and NaK on Oxidized Inconel. 3. REACTOR PHYSICS W, K. Ergen, ANP Division The theoretical investigations of mechanical vibration. With this the kinetics of the circulating-fuel simplified concept, a large temporary reactor continued, still disregarding disturbance was assumed to be applied for the present the possible coupling to the reactor, and it was found that between “nuclear' oscillation and 1f the reactor suarvived the first ) L ANP PROJECT QUARTERLY PROGRESS REPORT short maximums of power and temperature the following power and temperature maximums would be rather moderate. The general properties of the reflector-moderated circulating-fuel reactor were discussed in the last report. (') A number of multigroup IPM calculations has since been performed. However, this 1s the first time that the method has been applied to reactors with very different properties 1n core and reflector and with absorptions that are fairly large and also vary within a lethargy interval. Hence, con- firmation of the results by critical experiment 1s necessary, A calculation was performed regard- ing the critical mass, power distri- bution, and neutron spectrum of a mockup of the ABRE for use in critical experiments. So far only the critical mass has been determined experimentally, The agreement with the calculation was very good (cf., sec. 4, "“"Critical Experiments’’). A previously known method of com- puting the effects of gaps on reactivity was refined and has yielded satis- factory agreement with experiment. In addition, several slowing-dcwn kernels have been expressed of kvnown functions, in terms OSCILLATIONS IN THE CIRCULATING-YUEL AIRCRAFT REACTGR S. Tamor, ANP Division A summary report{?’ was issued re- garding the work on the kinetic equations(s) of a somewhat idealized (1)Aircraft Nuclear Propulsion Project Quarteriy Progress Report for Period Ending June 106, 1952, OBRNL-129%4, p. 6 and 31. S. Tamor, Note on the Non-Lineaer Kinetics of Circulating-Fuel Reactors, Y-F10-109 (Aug. 15, 195%)3 3 W. K. Ergen, Aircrafit Nuelear Propulsion Project Quarterly Progress Report for Period Ending March 10, 1952, ORML-1227, p. 41, 36 In addition to the results gquoted in the last report, (%) it was found that rather satisfactory limits can be set on the second power and temperature maximum following a disturbance of the resactor, circulating-fuel reactor, A large disturbance, such as intro- duction of excess reactivity, will be followed by large power and tempera- ture maximums. However, these maximums are of short duration and can, withan limits, be tolerated, 1f it can be ascertained that they will not be followed by maximums of similar height. Hence, it was regarded as important to investigate the second power and temperature maximums, The third and following maximums are smaller than the second because of the damping of the reactor oscillations, No general relationship has been set up between the upper limit for the second waximums and the other parameters of the reactor. However, two specific examples have been investigated by numerical integration. Both inte- grations refer to a reactor with the following parameters, which fall an the general range of contemplated ANP design: temperatuvre coefficient of reactivity, a, 10°*/°C; prompt gener- ation time, 7, 10°% sec:; reciprocal heat capacity of total fuel in the reactor, 2.67 X 10°°%°C/watt sec; reactor power, P, , 3 X 103 watts; fuel transit time, &, through reactor, 1/8 sSsecC., One integration referred to the case 1n which the power 1s four times the average power, P,, for all times t <0, At t > 0, the kinetic equations of ORNL-1227¢2? apply. The other case involved normal power, P , at all times up to ome-tenth of a transit time before t = 0, For a period of (4)5. Tamor, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending June 10, 1952, ORNL-1294, p. 31. FOR PERIOD ENDING SEPTEMBER 160, one~teth transit time duration before t = 0, the power is 10 P,. As before, at t >0 the kinetic equations of ORNL- 1227¢(2) valid. The power and temperature as a function of time were calculated through the second overswing. The results are shown in Figs. 12 and 13, It can be seen that the extreme values of P and T were nearly exactly the same for the two cases, indicating that they are close to the most extreme values obtainable. | are For the reactor constants given, it 1s found that after the first over- swing has passed, P 1s bounded by 0.013 < P/P, < 2.1, and the bounds on T are -50°C < T < +26°C. P4 740 dat POWER ( units of average power 1952 The similar shape and limiting values of the curves in the two above cases are easy to understand. After the large initial disturbance, the reactor almost shuts i1tself off. The period of very low power is essentially the same in both cases, about one transit time, and 1t determines the future time behavior of the reactor. the following fic- is plotted 1in For this reason, titious limiting case Fig. 14. The power is zero for t < 0. At t = 0, the power 1s suddenly brought to P,, and thereafter the kinetic equations of ORNL-1227(?) apply. Even in this case, the power and temperature oscillation is not much more violent than in Figs. 12 and 13. DWG .1€344 { lrjlog F=-12.5 fl_P(;“M]-—U —-1] 7 &7 O TEMPERATURE {units of r/af) O 0.2 04 06 08 1.0 1.2 1.4 16 1.8 2.0 2.2 2.4 TIME (units of transit time) Fig. 12. Power and Temperature vs. Time for P = 4 P,, t < 0. - 37 ANP PROJECT QUARTERLY PROGRESS REPORT DWG. 16345 £l 7 log P =425 F<~0 g 097 =-le A =10 0 —01 <0 3 8 — v o 5 S 5 2 2 E £ 2 o S # if i 1 & G o. = L) '—... 0 0z 0.4 0.6 08 1.0 t.2 14 16 1.8 20 2.2 TIME (units of transit time) Fig. 13. Poweyr and Temperature vs. Time for P = Py, t < =0.1; P = 10 P, -B,1 < t < 0, Of course, these calculations have stationary-fuel reactors as well and to be gualified. It is necessary for are not uniguely problems of a circu- the survival of the reactor that the lating-fuel reactor. first temperature maximum stay within tolerable limits. Furthermore, 1if sufficient excess reactivity were EFFECT OF GAPS ON REACTIVITY added for a prolonged period, the 5. T ] o average temperature of the reactor 5. Tamor, ANP Division zould 1ncrease_beyonqullowabfie lirjts As mentioned in the last report, (5) cor A ]”N?g time, so, the a'zve calculations regarding the effect of a calculations neglect any possible transverse gap through a reactor have coupling between the “nuclear’ oscil- lati1ons and mechanical vibrations. et e All these qualifications apply to ) reida., p. 39, 38 FOR PERIOD ENDING SEPTEMBER 10, 3.0 1952 S e ;:;_y,fi‘-?f:_ T DWG.1634¢6 1 g 2.6 77 109 P=~—12.5'/ [P(f—l-n—fl-—l]-q dn 0 2.2 POWER (units of average power) 0.6 0.2 TEMPERATURE (units of r/28) O 0.2 0.4 0.6 08 TIME (units of tronsit time) Fig., 14, Power and Temperature vs. These calcu- large reactor been carried out.¢%’ lations refer to a bare, and a gap that ts small compared with the transverse reactor dimension, They do not neglect, as in earlier calculations(?? on the same problems, the fact that neutrons preferentially stream toward points of low neutron density. The new calculations con- stitute an improvement over the earlier ones in cases where the gap is small. The Effect of Gaps on Pile Re- (6)5. Tamor, 1952). activity, ORNL-1320 (July 14, —4 ~—8 1.0 1.2 14 1.6 18 Time for P = 0, ¢t < 0; P =P, t = 0, comparison be- based on the Figure 15 shows a tween calculations earlier method, the calculation according to the new method, and the experiments on a 130- by 112- by 112-cm, rectangular, graphite-moderated reactor with the gap perpendicular in the long dimension and slightly off center. The abscissa is essentially the gap width, (7) J. E. Beactivity, M. G. Goldberger, M. L. Goldberger, and Wilkins, Jr., The Effect of Gaps on Pile CP-3443 (Feb. 20. 1946), 39 ANP PROJECT QUARTERLY PROGRESS REPORT e DWG. 16347 100 | NS/ / REACTIVITY LOSS {cents} 03 Reactivity Loss vs. Gap Fig. 15. Width. SLOWING-DOWN KERNELS R. D. Worley, USAF The following kernels, of interest to reactor theory, have been expressed in terms of known and tabulated functions: (&) (1) the Fermi age kernel in the case of weak absorp- tion, (2) the kernel describing, for an infinite medium, the slowing down of a neutron according to age theory and the subsequent diffusion at thermal energy, and (3) the kernel describing the slowing down of a neutron in an infinite space filled with a medium of slowing down, and except that a uniform scattering, absorption properties, finite part of the space also con- tains an additional absorber with an absorption line at a definite energy. (B)R. D. Worley, Slowing-Down Kernel with Absorption and the Convolution of Various Kernels, Y-F10-105 (to be issued). 4. CRITICAL EXPERIMENTS A. D. Callihan, Physics Division Measurements have continued on the mockup of the G-E direct-cycle, air- cooled, water-moderated reactor. Some berylliumin the reflector was replaced by a plastic and steel composite reflector to compare the two arrange- ments with respect toreflector savings and reactivity., Loss 1in reactivity incurred by the insertion of baron carbide control rods was determined, The preliminary assembly of the fluoride-salt circulating-fuel air- craft reactor experiment was completed, and criticality was attained with 5.8 kg of U2?%% in the core and a total investment of 6.5 kg of U??*%, An analysis of the results indicated almost perfect agreement with theoreti- cal caleculations. 41) DIRECT-CYCLE REACTOR R. C. Keen D. V. P. Willaiams Physics Division D. Scott, ANP Division The study of a mockup of the G-E direct-cycle reactor has continued at ORNL. The critical assembly was fully described 1n earlier reports,(1> and the data summarized here have appeared in more detail elsewhere.(?) (I)E. V. Haake, D. V. P. Williams, ¥. G. Kennedy, and D. Scott, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period EndingMaerch 10, 1952, ORNL-1227, p. 5§9; V. Haake, D. V. P. Williams, B, C. Keen, W. G. Kennedy, and D. Scott, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending June 10, 18352, ORNL- 1294, p. 34. (2)A. D, Reactor Assembly, 1952}, Callihan, Preliminary Direct Cycle Part III, Y-B23-5 (June 18, and Part IV, Y-B23-7 (June 30, 1852). FOR Réflector Studies. Some measure- ments have recently been made with the bervllium 1ina section of the reflector replaced by a composite of stainless steel and a hydrogenous plastic, Boron, when placed between the com- ponents of this composite reflector, will reduce the neutron-induced gamma radiation from the iron. The purpose of this study was to compare the reflector savings of the plastic and steel composite reflector and the neutron distribution in 1t with and without the boron, The test section was 5 3/4 in., thick, 14 3/8 in. wide and 36 in. long. The beryllium re- flector in these cells was replaced by type 310 stainless steel except for a 5/16-in.-thick layer adjacent to the cere, which was filled with Boral strips 36 in. long. The Boral strips were prepared from a boron carbide and aluminum mixture that contained 35wt % boren carbide. The mixture was sandwiched between two layers of aluminum sheets about 0.04 in. thick and wrapped with masking tape. Thesys- tem was made critical by the addition of beryllium at locations remote from the test section and by control rod adjustment, and the loss in reactivity was measured, By keeping the over-all dimensions of the test section con- stant, the thickness of the steel was reduced and plastic was added between the core and the Boral in a step-wise manner until all the steel had been removed, The changes in reactivity that were incurred by these reflector alterations, referred to the all- beryllium reflector, Fig. 16, An additional experimental point on Fig. 16, designated as “2 9/16-in. Plexiglas and 2 7/8-in. stainless steel,”” shows the result of removing the Boral from one plastic and steel composite reflector section. Another, measured with the Boral re- placed by 5/16 in., of plastic, shows that the effect of a 5/16-in. void between the plastic and the steel is are recorded 1in PERIOD ENDING SEPTEMBER 10, 1952 small. The change in reactivity upon removal of all material from the test cell 1s also shown on the graph. It is noted that with the hydrogenous component 3 in, or more thick, the presence of the boron does not greatly reduce the reflector savings. Com- parison of these data with earlier results show, however, that the beoron reduces the reactivity up to 30% for plastic thicknesses approaching zero. 8&6. 16348 i | REFLECTOR COMPOSED OF ~2%g~in. PLEXIGLAS AND 2% -in. STAINLESS STEEL B0 e [ 70} - REACTIVITY LOSS {cents) W o < o & 3 VOID 1IN TEST CELL 1300 ) . S RURNUORY RSPUOUUN: AUOSPUUES HEESOUOOIO NSO 0 f ? 3 4 5 6 PLEXIGLAS THICKNESS (in) Fig. 16. Reactivity Losses vs. Plexiglas Thickness for Stainless Steel~Plexiglas~-Boral Composite Re- flector. Bare-indium and cadmium-covered- indium traverses were made through one of the above stainless steel, Plexiglas, and Boral composite re- flectors starting at a point 9 in. inside the core periphery and termi- nating at the edge of the reflector. The reflector consisted of an outside layer of 2 7/8 in. of stainless steel separated from a 2 9/16~in, Plexiglas 41 ANP PROJECT QUARTERLY PROGRESS REPORT inner layer by the 5/16-in. Boral sheet, The data for the traverses through the composite reflector when the Boral strips were present in the test cells are given in Fig, 17. Curve 1l is the bare-indium activation, curve 2 shows that obtained for the cadmium covered indium, and curve 3 shows the difference between curves 1 and 2, All data were plotted as a function of the distance from the outside of the test sec&ion. Similar traverses were made with the Boral strips omitted and the space left empty. Figure 18 shows the bare-indium and cadmium-covered-indium activations and their difference plotted also as a function of the dis- tance from the ovtside of the test section. DWG. 16349 ACTIVITY {counis/min X 10%) STAINLESS | /' STEEL 0_5 R “\ CoTTT T o '_‘“ """ N ;_’1 """"""" T i O . } ‘ &‘::‘.fi—-f/’:'/:? TS 2 9 6 3 o DISTANGE FROM OUTSIDE OF REFLECTOR {in) Fig. 17. Bare and Cadmium-Covered Indium Traverses Through Stainless Steel-Plexiglas~Boral Composite Re- flector. 42 e O] AT S 30 . 25— ACTIVITY {counts/min X ’03) — A R \ 3 \\\\\\ | e N _ N uf e 2.0 < [ | i \\ / } J i VoID ! \\ 2 t | i 15 e e S i s - | | )E li _ 1Ok el o h STAINLESS_ ‘ | STEEL | ‘ PLEXIGLAS 05 b \ \ ' CORE-__ ¢ | | \ okf \ L = 15 12 9 6 3 0 DISTANCE FROM QUTSIDE OF REFLECTOR (in.) Fig. 18. Bare and Cadmium-Covered Indipm Traverses Through Stainless Steel~Plexiglas Composite Reilector. Poison Rod Calibrations. A measure- ment was made of the reactivity depression by each of three boron carbide rods furnished by the General Electric Co, for test as poison control rods. The rods were 1/2 in. in diameter and about 36 in. long and differed slightly 1n 1internal structure; one contained a plastic insert within the boron carbide and the other rontained aluminum. The loss in reactivity 1in- curred by substituting the poison for plastic within the reactor was found to be about 20 cents‘?? at the center of the core and about 11 cents at the (S)One hundred cents is a rcactivity change equivalent to the effective fraction of delayed neutrons. FOR PERIOD ENDING SEPTEMBER 10, 1952 edge. All the rods were of comparable value. ARE CRITICAL ASSEMBLY D. Scott C, B. Mills ANP Division The preliminary assembly of the ARE was described in the preceding re- port.(4) The moderator and the re- flector are beryllium oxide and the fuel is enriched UF,. Hexagonal beryllium oxide blocks with vertical axial holes are stacked 1n a cylinder 36 in, high and 47 in. in diameter. In the central blocks, a. section about 33 in. in diameter, are placed stain- less stee)l tubes 1 1/4 in. in diameter and 40 in. long containing a mixture of 66 wt % Zr0,, 24 wt % NaF, and 10 wt % graphite to which was added sufficient UF, to give a U?3% design density of 0,16 g/cc. These com- ponents, as dry powders, have a density of about 1.9 g/cc. The same mixture, without the uranium, has been packed in 1/2-in.-dia tubes and placed in the in the peripheral beryllium The assembly holes oxide reflector blocks. was made critical on August 21 with a loading equivalent to slightly more than 61 fuel tubes containing approxi- mately 5.8 kg of U?3% 4in the core. The total uranium investment, 1n- cluding the sections of the tubes 1 i, ¢ )D. Scott, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period fnding June 10, 1952, ORNL-1294, p. 38, extending beyond the beryllium oxide, is about 6.5 kg of U23%, Before the critical experiment was performed, a calculation of the critical mass, power distribution, and neutron spectrum had been carried out.¢%) According to this calcu- lation, the assembly should have gone critical with about 58 tubes. However, the following three effects, which had not been included in the computation, each add about one tube to the com- puted critical mass so that there is almost perfect agreement between ex- periment and theoretical prediction. 1. A chemical analysis of the Inconel actually used was received after the computation was finished. The analysis showed the Inconel to be more of a poison than anticipated. 2. The computation did not consider the loss of reactivity caused by voids into which additional fuel tubes would have been inserted 1f necessary. 3. Over and above the poison con- sidered in the computation, some polson was introduced by a stainless steel piece in the control assembly. No experimental data are available, as yet, regarding power distribution or neutron spectrum, (S)C. B. Mills and D. Scott, The ARE Critical Experiment, Y-F10-108 (Aug. 8, 1952). 43 SUMMARY AND E. P. Blizard Physics The 1id tank facility was used primarily for tests of the reactor shield for the G-E direct-cycle design (sec. 5). Extensive surveys have beén made on the flux patterns in and around large annular air ducts. In addition, a study has been carried out on the production of secondary gamma rays 1in a metal layer near the reactor., The attempted correlation between the water-neutron data from the 1id tank and the bulk shielding facilities indicated that the data from the lid tank facility give an attenuation that 1 low by a factor of about 2. Thais is of particular importance because of the number of shield designs based on li1d tank data. | The gamma~ray spectral measurements on the divided shield mockup have been completed 1n the bulk shielding facility {sec. 6). The last measurements made gave the energy and angular distri- bution of the gamma rays streaming around the edge of the lead shadow shield, Approval has been obtained to operate the bulk shielding reactor at 100 kw instead of 10 kw and experi- ments have been started to remeasure the air scattering from the divided INTROBUCTION J. L. Meem Division shield., An experiment is being carried out to determine the amount of energy released per fission, and a special fuel element has been constructed for this purpose. JTrradiation of mice, rats, and rabbits by neutrons from the Cockeroft-Walton accelerator was con- tinued, and plans are being made to expose monkeys to the radiation from the bulk shielding reactor. A graphite thermal column has been installed and will be used with the gamma-ray spec- trometer for measuring capture gamma rays. Construction of the tower shielding facility has recently been approved, and the design criteria have been completed (sec. 7)., Preliminary design work onmechanical and electronic components 1s under way, and orders have been placed for some of the commercially available components. Fission cross sections of U?%* and U23% up to 4 Mev and the total cross section of N** have been measured by using the 5-Mev Van de Graaff neutron accelerator (sec. 8). Some measure- ments with the time-of-~flight spec- trometer have been made on U?%® and Th. 47 ANP PROJECT QUARTERLY PROGRESS REPORT 3. LID TANK EXPERIMENTS C. E, Clifford L. S. Abbott J. D. Flynn G. T. Chapman F. N. Watson J. M. Miller M. K. Hullings Physics Division C‘ [Jl Experimental efforts inthe lid tank facility have centered around the de- sign of the shield for the G-E direct- cycle reactor., Studies made on a mockup of the outlet air ducts in the GE-ANP initial engine have provided estimates of the radiation that will penetrate the G-E aircraft-reactor shield., These estimates agree with the approximate values obtained theoretically by using assumptions of isotropic scattering at the bends and geometrical attenuation down the ducts, The gamma shield for the side of the G-E reactor was also mocked up in an attempt to optimize the arrangement of the components. In addition, the contribution of a metal layer near the reactor 1s being evaluated both in terms of the 1ncrease in shield weight and the reduction in radiation level, Calculations have been made in an attempt to correlate the neutron attenuation data obtained in the lid tank facility with those obtained in the bulk shielding reactor. The lid tank measurements seem to be lower by a factor of about 2, depending on the distance from the source., Since many shield designs are based on lid tank data, an attempt 1s being made to improve this situation by altering the lid tank geometry source, AIR DUCTS A mockup of the air outlet section of the GE-ANP initial engine, com- prising a transition section and duct 48 Storrs, General Electric Co. (Fig. 19), has been examined in the lid tank facility. Measurements were made of the thermal-neutron intensity, the fast-neutron dose, and the gamma dose. Figures 20 and 21 give the results plotted as isodose curves, and Figs., 22 and 23 give fast-neutron traverses as taken, These measurements are primarily of interest in estimating the radiation that will penetrate the shield of the G-E aircraft reactor., The geometry 1is so complex that calculations of the effect of the duct are difficult and (with the present knowledge) not very reltable. To a very rough first approximation, however, the measured transmission of neutrons agrees with calculations based on the assumptions of i1sotropic scattering at the bends and geometrical attenuation down the ducte. (1) At the sides of the tank, where the neutron flux varied laterally across the counters, discrepancies appeared as a result of different counter diameters., This 1s attributed to the fact that the effective center of counting 1s not the geometric center of the counters. Careful corrections for this effect were not made, and as a consequence the dosages given may be in error by an amount corresponding to a 1- or 2-cm displacement in the water, The relative spacing of the isodoses, however, is considerably more accurate. (l)A. Simon and C. E. Clifferd, Simplified Duct Theory (to be issued). FOR PERIOD ENDING SEPTEMBER 10, 1952 L OWG. 15405 42 in. —% ; _ ! = | . ‘ S i ! Hj L~ o P 02" . TRANSITION \< SECTION Fig. 19. G-E Qutlet Air Duct. 49 ANP PROJECT QUARTERLY PROGRESS REPORT s THERMAL FLUX NORMALIZED TO NEUTRCN DOSIMETER 18Q 160 FIGURES ON CURVES ARE DOSE LEVEL IN mirep/hr 140 ; ST et 120 o @ & o o Q £, DISTAMCE FROM SQURCS (em) o o 2¢ '3 AIR-FILED TRANSITICN SECTIO! canlen SOURCE /7, 1¢ 120 cOo 80 60 49 20 o 20 40 60 80 oo 120 Y, HORIZONTAL DISTANCE FROM SOURCE AXIS (cm) Fig. 20. Neutron Isodose Measurementis with G-E Outlet Air Duct Mockup. {cm} ul Z, DISTANCE FROM SOURC 50 DWG 154424 180 160 140 120 100 80 60 40 -~ TRANSITION SECTION 20 i SOURCE Yl 140 120 100 80 &0 40 20 0 20 40 60 B8O 100 120 140 Y, HORIZONTAL DISTANCE FRCM SOURCE AXIS {cm) Fig. 21. Gamma Isodose Measurements on G-E Outlet Air Duct. FOR PERIOD ENDING SEPTEMBER 10, 1952 2 - DWG. 154644 . 2X 10 {mrep/hr) y=t047cm 1] U3 e NEUTRON DO =75¢m 80 52 Z, DISTANCE FROM SOURCE (cm) Fig. 22, Fast-Neutron Dosimeter Measurement on G-E Outlei Air Duct Showing z Traverses for Various Values of vy. 51 ANP PROJECT QUARTERLY PROGRESS REPORT DWI! 154654 2% 107 NEUTRON DOSE (mrep/hr) =57cm 20 32 a4 68 80 g2 140 Z, DISTANCE FROM SOURCE (cm) Fig. 23. Fast-Neutron Dosimeter Measurement om G-E Outlet Air Duct Showing z Traverses for various Values of y. 52 FOR %F1 the neutron data given were obtained with the shutter between the reactor and the source plate open. The counting rate with the shutter closed, that is, with only the back- ground neutrons from the reactor, was between 2 and 10% of the rate with the shutter open. ‘ Most of the thermal-neutron data were obtained by using an automatic counting-rate plotter connected to a counter that was moved at a constant rate across the tank. Some isodoses subsequently obtained directly an automatic 1sodose plotter. The obtained by the two methods are in agreement. were with data good The fast~-neutron dose was obtained at several fixed positions with a poly- ethylene-lined dosimeter,(%2) The thermal-neutron fluxes can thus be normalized to fast-neutron dose, to which they should be directly pro- portional, provided there are no major perturbations in the neutron spectra, The gamma background dose with the shutter closed was of the order of 50% of that with the shutter open and therefore not negligible, Since, furthermore, this ratio varies with position, the difference was perforce measured for all points to obtain the attributable to the fission source., dose The distribution of neutron and gamma dose around this duct was also measured for various lengths of tran- sition section. The gamma-dose distri- butions for transition section lengths of 0, 6, 12, and 18 in. have been completed., For each geometry a center- line traverse was made (Fig, 24) and an isodose curve plotted (Figs. 25 and 26)., Figure 27 shows several isodose (Q)F. M. Glass and &. 3. Instruments 23, 67 (1952). Hurst, Revw. Sci. PERIOD ENDING SEPTEMBER 10, 1952 curves made with only water in the tank. The center-line measurement given in Fig. 23 shows the difference be- tween the readings with the shutter open and closed. The 1sodose plots have not been subtracted, since 1t 1s apparent that the transition section has no marked effect upon the dose. The gamma isodoses are all auto- matically normalized to alecal reactor power monitor. Since this monaitor reads less by a factor of 0.777 when the shutter i1s closed, the corres- ponding 1sodoses represent a dose higher by 1/0.777 than the shutter- open isodoses, as 18 noted on the figures. In the case of the pure- water isodoses, the resunlting curves very nearly coincide, whereas with the duct mockup in place there i1s a sepa- ration of about 10 cm. This can be attributed to the presence of gammas from thermal-neutron capture 1in the iron of the duct. The thermal-neutron flux, of course, is considerably affected by the shutter. R-1 REACTOR(3) GAMMA SHIFLDING The gamma shielding on the side of the G-E aircraft reactor was mocked up in the lid tank facility, and measures were taken to determine its optimum position in relation to the pressure shell. The shield design under con- sideration consisted of 4 in, of pure- water reflector next to the reactor, a l-in, steel pressure shell, and additional iron and lead in borated water, The guestionunder consideration here was whether the additional iron and lead could be placed next to the pressure shell or whether 1t was (B)The General Electric Co. direct~cycle air- craft reactor, see Airecraft Nuclear Propulsion Project Enginecering Progress Report No. 4, for April 1, 1852 - June 30, 1552, APEX-~4. 53 ANP PROJECT QUARTERLY PROGRESS REPORT GAMMA DOSE {(mr/hr} DWG. 156334 12-in. TRANSITION SECTION | 156 G-£ Outlet Air Duct. 10 i o &0 72 84 96 108 120 132 Z, DISTANCE FROM SOURCE {cm) Fig. 24. Gamma Center Line Measurements with necessary to locate it farther out 1in the borated water to reduce secondary gamma production, In the mockup used in the lid tank facility, one 2.2-~cm iron slab was placed 4 in. from the source to represent the pressure shell., This was followed by a 38B-in.-wide steel tank, the inner wall of which (0.32 ¢cm) is considered part of the pressure shell., This tank was filled with 54 borated (0.6 wt % boron) water in which the gamma shield wasmoved with respect to the source. The gamma shaield, sisting of one layer of i1ron and two unequal layers of lead with each layer separated by small thicknesses of borated water, is shown 1n Fig. 28. The gamma rays were measured 1in the water behind the tank for six positions of the gamma shield with both an 1o0n chamber and an anthracene scintillation counter. A seventh configuration was con- Fi it 180 160 140 120 1C0 80 60 Z, DISTANCE FROM SOURCE (cm) 40 20 Fig. 140 25. 120 100 80 60 40 FOR PERIOD ENDING SEPTEMBER 10, 1952 D\!! l !634.5 ———NQ TRANSITION SECTION — 18~in. TRANSITION SECTION SHUTTER OPEN, 102 r/hr SHUTTER CLOSED 1 1/0777 X152 r/hr AIR-FILLED oucT TRANSITION SECTION AiR-FILLED DUCT 7 20 0 20 40 80 20 100 120 140 ¥, HORIZONTAL DISTANCE FROM SOQURCE AXIS (¢m) Gamma Isodose Curves with G-E Outlet Air puct Showing Effect of Tramsition Section. 160 140 120 100 80 L LT Ao , A_JWW =Y o SHUTTER OPEN,IO # r/hr N p \i R e PR 60 Z, DISTANCE FROM SOURCE (cm) === i2~in. TRANSITION SEC s B-in. TRANSITION SECTION TION 1ol e w! = - LIOTTT MO E pl b 40 4 — e mimabrnmmn e seds e il e e . _.,,_.;,:__,,_,_._L,__. oo bl Ll L } | L L ) RN | 140 120 100 80 & 40 ¥, HORIZONTAL DISTANCE FROM SCURCE AXIS (cm) Fig. 26. Gamma Isodose Curves with G-E OGutlet Air Duct Showing Effect of Transition Section. 55 ANP PROJECT QUARTERLY PROGRESS REPORT 180 ” - SHUTTER OPEN 160 e o 140 f—1 - - SHUTTER OPEN = bt DA , .- =120 | L UTTER CLOSED e L L L S I S U 2 SHUTTER CPEN 100 - = o T 5 SHUTTER CLOSED o 80 SHUTTER GPEN 2 . < v = 60 N TTER CLOSED 40 20 140 120 100 80 60 40 20 DWG.15636A 10_11 r/hr P sefenees : 1 10”2 r/hr 10°% r/br ] 20 40 60 80 100 120 140 Y, MORIZONTAL DISTANCE FROM SOURCE AXIS (cm) Fig. also measured in which the gamma shield was at the outside of the tank (farthest from the source) and the 2.2-cm iron slab for the pressure shell was removed. These data are shown graphically in Figs. 29, 30, and 31. Figure 32 shows the variation in gamma dose at 70 cm from the source as a function of gamma shield position, Moving the gamma shield back 28 cm from the pressure shell reduced the gamma intensity by only 20% although the weight of this component was in- creased by 5000 pounds. This intensity reduction can be achieved, on the other hand, by adding only 1 cm of iron 28 cm back from the main gamma shield with a corresponding weight increase of only 2000 pounds. The iron slab 10,16 ¢cm from the source reduced the radiation by about 20%, whereas 1t would have reduced the 56 27, Gamma Isodose Curves with Pure Water. intensity by about 33% if neutron-capture gammas in the 1ron itself had been eliminated. g a mma OTHER LID TANK EXPERIMENTS Other work that has just been com- pleted in the lid tank facility and has not been published consists of measurements on a mockup of the inlet air duct¢*? of the G-E aircraft reactor and a mockup of an array of round, wavy ducts (Fig. 33), which arestill considered to be a possible means of carrying the cooling air to and from the reactor., Work 1s now under way to determine the effect on thermal-neutron dosage of borating the water around the exit air duct. o ( )AircraftfluclearPropulsion Project Quarterly Progress Report for Period Ending June 10, 1952, ORNL-1294, p. 753, Fig. 41l. SOURCE FOR PERIOD ENDING SEPTEMBER 10, 1952 5 £ §5 EE © BE an 2F S ool ¥ o Ad . I;M DWG. 161554 31 35¢em 5t67cm 9 >~ 3 ~ - X 2 WATER —— o < 2 3 m —uQ = 5 m 3 el el Ll L : M- Tl 3 < ! W x m o 3 = ~ . - F . = e = S Y b T XL N "oieg D 250 GO .’ oo Lesipattiy SR ) ] ,l R & o0 t s BN HIARL XSRS Qo0 (LR & . ERRXS RS X LI, -:‘p LIRS i I I IIPITTT 77 &\\ s % .‘ - 3 2 L A e ¥ i Ll L, = N =Wz 5 cm— . : CONFIGURATION 2 KEY: CONFIGURATION 3 SAME AS ABOVE EXCEPT W=1i0 cm ‘CONFIGURATION 4 SAME AS ABOVE EXCEPT: W =15 em CONFIGURATION 5 SAME AS ABOVE EXCEPT W=20cm PURE WATER 74 IRON BORATED WATER B LeaD 4 o SEAUT X, K AR ) B IO DLICRI e '0 £ (RPN S :"::;:‘..‘ ) 205 TR, LI & SRR ey 3 e 3 £ Q‘fl ] o ML) R I SR LA n LA L e i SRR NI RS R A M Pt n i rte i e > LK £ o 2 el & ‘O brormacicn B Pt 2 e, o - T T (i Yl : I ; oo \i’ Ly | TN | gy e U Vh\\t | ; t i E W=288 cm CONFIGURATION & Faretiatet XL oo SIS o S0 RO RIS I IS L AR R MR AT Xl X | AL 114 CONFIGURATION 7 Fig. 28. Mockup of Gamma Shield for Side of GE-ANP Reactor. ANP PROJECT QUARTERLY PROGRESS REPORT DWG. 1633 2 X 10° I s CONFIGURATION 1 v : o _ 107~ CONFIGURATION 3 | = = ] - . CONFIGURATION § e e — ] 5 ,,,,,,,,,,,,,,,,,,,,, E E i ) O 0 3 Z w0 - q e ——— el W 5 — — I 10'® 1ION CHAMBER Lo 1, _ A Yp-in ANTHRACENE-CRYSTAL N SCINTILLATION COUNTER & ] ?- ....... —_ 10’ 30 50 70 90 10 130 150 (70 Z, DISTANCE FROM SOURCE {cm) Fig. 29. Gamma Measnrements Beyond Mockup of Side shield for GE-ANP Reactor. Configurations 1, 3, and 5. 08 FOR PERIOD ENDING SEPTEMBER 10, 1952 o, DWG. 16352 2X10 & _CONFIGURATION 2 Tok & N AN CONFIGURATION 4 - 5 \%\ = - ‘ X £ ‘ E 3 wi 2 u O [ ] <{ = s > 10 \\ o 0 AN & 10 10N CHAMBER LN |4 Y-in ANTHRACENE - CRYSTAL. N\ SCINTILL.ATION COUNTER | 5 : : \{ \ N R 2 A O 10° 30 50 70 90 110 130 150 170 Z, DISTANCE FROM SOURCE (cm) Fig. 30. Gamma Measurements Beyond Mockup of Side Shield for GE-ANP Reactor. Configurations 2 and 4. ANP PROJECT QUARTERLY PROGRESS REPORT ) 150 2 DWG 16353 2 X105 CONFIGURATION 7 102 N N S _+.. _ _ ””” - . CONFIGURATION 6 B = = £ e 2 [ [ —_ el w 2 8 Y,~in. ANTHRACENE - CRYSTAL 3 SCINTILLATION COUNTER = <1 v 10 T - 5 L i [ e 2 1 - 10° | | 30 50 70 90 Ho 130 Z, DISTANCE FROM SOURCE {cm) Fig. 31. Gamma Measurements Beyond Mockup of Side Shield for GE-ANP Reactor. Configurations 6 and 7. 60 FOR PERIOD ENDING SEPTEMBER 10, 1952 o e DWG. 16354 2 X 10° 1 02 h e ——— — G, e e . O S _ L ; _ 3 . O - e . =233 T . 5 Ll D & _ . o O @B = . - O o R E o] 1 . O [ e «t z E 10 LLJ e o s -— e et 2 | O - o o < e e = ; 3 o B e —T L e - e ] 2 - ettt mt e e e e e ———————— e Tox O 5 10 15 20 25 30 BORATED WATER BETWEEN PRESSURE SHELL AND SHIELD {(cm) Fig. 32. Gamma Dose as a Function ;of Borated Water Thickness in Side Shield Mockup. ' 61 ¢9 Fig. 33‘ P Wavy-Duct GE-ANP Mockup in Lid Tank. e PHOTO 10289 LHOdIY SSTY90Ud ATHALYVAO LJAl0¥d dNY FOR CORRELATION OF NEUTRON ATTENUATION DATA E. P, Blizard Physics Division By the use of somewhat elaborate transformations, the water-neutron attenuation data from the lid tank facility and the bulk shielding reactor have been compared.(s) The 1id tank data seem to be lower by a factor of somewhat less than 2 at large distances but more than 2 close to the sources., The reason for this discrepancy 1is (S)E. P. Blizard and T. A. Welton, The Shield- ing of Mobile Reactors - II., ORNL-1133 (to be published in Reactor Science and Technology). PERIOD ENDING SEPTEMBER 10, 1952 not understood, but several possible explanations have been suggested. Primary among these is the possibility that the lid tank source plate 1is actually operating at a somewhat lower power than was indicated by early measurements., The calculation of 1lid tank source-plate leakage is compli- cated by the rather difficult geometry of the cylindrical, natural-uranium slugs of which the source is made, Installation of new, thin-plate U238 is being studied to determine how much this might be expected to improve the situation. Indications are that it would be very desirable since so many shield designs are based on lid tank data, 6. J. L. Meem R. G. Cochram M. P. Haydon K. M. Henry L. B, Holland - F. G. M. McCammon BULK SHIEFLDING REACTOR Hungerford . Johnson Leslie . Love . Maienschein 0P xS m Physics Division The first series of studies of gamma-ray spectra and angular distri- butions with the divided shield mockups have been completed. The final measurements incorporated a lead shadow shield and data obtained with this arrangement will be used to calculate the gamma-ray dose at the crew compartment. Plans are being made for using the recoi1l-proton spectrometer to measure spectra and angular distributions of fast neutrons leaving the reactor part of the divided shield mockup. In order to resolve the uncertainty of the power of the bulk shielding reactor, an attempt 1s being made to obtain a more accurate value of the energy released per fission. A special fuel element equipped with thermocouples and water-flow tubes has been con- structed so that the energy per fission can be determined as a measure of temperature rise and water flow. Preparations are under way to expose monkeys to radiation from the bulk shielding reactor, and a tentative program has been arranged. Animals previously irradiated by the Cockcroft-Walton accelerator are now under observation. A graphite thermal column has been installed in the bulk shielding facility to provide an intense source of thermal neutrons. With this source, the gamma rays produced by neutron capture can be studied. 63 ANP PROJECT QUARTERLY PROGRESS REPORT MOCKUP OF THE DIVIDED SHIELD Gamma-ray spectral measurements during the past quarter were made with a lead shadow shield added to the mockup of the aircraft divided shield.(!) These measurements complete the first series of studies of gamma-ray spectra and angular distributions with the divided shield mockup. The data obtained will be used for a preliminary calculation of the gamma-ray dose received at the crew-compartment location, which will then be compared with the estimates of the Shielding Board.(?) The results obtained with the lead shields are shown in a series of five figures. Figure 34 shows a gamma-ray 1sodose curve obtained by using an 1ion chamber without the lead disks. This curve 1s very nearly a segment of a circle with a radius of 168.2 cm and a center near the reactor center. The other curve in Fig. 34 shows the dose along this same circle after addition of the leak disks. From these data it was decided to examine the energy and angular distri- butions at the angles ¥ = 0 deg and Yy = 50 deg, where Y is the angle between the aircraft axis and a line connecting the pseudo reactor center and the nose of the spectrometer collimator. The energy spectra ob- tained for various values of the angle & are shown in Figs. 35 and 36 for Y = 0 and 50, respectively, where fis the angle between the spectrometer collimator and the line joining the pseudo reactor center and the nose of the spectrometer collimator. The corresponding angular distributions are shown 1n Figs. 37 and 38. (I)Aircraft Nuclear Propulsion Project Quarterly Progress Reports, ORNL-1227, p. 73 and ORNL-1204, p. 44. (2 peport of the ANP Shielding Board, NEPA- ORNL, ANP-53, Appendix C (Oct. 16, 1950). 64 Development of the recoil-proton spectrometer 1s continuing. Plans are being made to use the instrument for measuring spectra and angular distributions of fast neutrons leaving the reactor part of a divided shield mockup when the air-scattering ex- periments are completed. A report on the theory of this instrument has recently been published.(3) AIR-SCATTERING EXPERIMENTS A memorandum(?%) giving the details of the air-scattering experiments described in the previous report has been issued. Similar experiments at 100-kw reactor power are now getting under way. REACTOR POWER DETERMINATION One of the largest uncertaintiles in the determination of the power of the bulk shielding reactor rests in the lack of an accurate value for the energy released per fission., A special fuel element has been constructed to be used in measuring this quantity (Fig. 39). The center fuel plate of the element i1s removable and disks are punched out of the plate. The gamma activities of these uranium- bearing disks are counted after an exposure to determine the number of fissions that occurred in the fuel element and then compared with the gamma activity in a similar disk previously calibrated in the standard reactor. The special fuel element is equipped with thermocouples and water-flow tubes. Special precautions against (3)8. R. Gossick, General Principles of Proton-Recoil-Fast-Neutron Spectrometer, ORNL-1283 (July 14, 1952). )J. L. Meenm and H. E. Hungerford, Air Scattering Experiments at the Bulk Shielding Fecility, ORNL CF-52-7-37 (July 8, 1952). 89 —GANMMA DOSE ALONG 1SOT0SE CURVE ~ WITHOUT LEAD SHADOW SHIELDS LEFT SiDE B.49 X107 r/he/wett SCALE DIAGRAM SHOWING LOCATION OF ISODOSE CURVE : roheswe WTH RESPECT TO REACTOR SHIELD = RIGHT SIDE: _ S POLAR PLOT OF GAMMA DOSE KEY TO SYMBOLS ) - 7R ALONS 1SODOSE CURVE WITH & POSITIONS OF DETECTOR AT WHICK EXPERIMENTAL ARG SHADOW SHIELDS iN POSITION DATA WAS O3TAINED - SN = :SODOSE CURVE VALUES OBTAINED BY INTERPOLATION S "‘\\ o \\ \ © EXPERIMENTAL PCINTS . ., DA . ™ \\\ ~ \ . . e \ \\\ \\\\\ ~ N \\\ “, N (em) N OIIR 80139 160 140 120 100 80 ) 40 20 3 LR 180; ‘ T ', I ‘ : LN N X ! ! | : ’ ; : .‘3 . : : \ N : ‘[ b | DOSE: 849 X107 vhrwoltl—dee L > ; ; oo ;‘ : | N N N 160} — | \¥~; T SN o | e el | 2 \ ] . N : ch/ P S a0l | GAMMA SODOSE CURVE | i pe— LEXPERIMENTAL PATH [ 1 ¢ : i {WITHOUT Pb SHADOW SHIELD) LA~ —h FOLLOWED BY GAMMA CHAMSER N\ L L_Lu+_i“flt£ oo b A T ST b TS . > N\ 120 T s0uD Ling 5 \ r g A L || 1SODOSE CURVE‘_-?/I L N\Je168 2 m RADIUS \OGAMMA DOSE BEHND | ’: ; N Do P O\ LEAD SHADOW SHELDA\Y 100— f } Sl —— — : ON ISODOSE CURVE \} . o N A g 1 AN + . " ’ )i = : /—A ; : \ N \ \ O T T \ g.._—il_;__Ltf}Di\SHED LNE A X( ‘ S \ P 1/ RADALCURvE | [ \ | VL 50 ; N B m e b St &X \ e F : oo : - AT e b L \ P g | LEAD SHADOW SHELDS ‘ | | N . , b | | - VoL 20! it - - vl ; T ‘ 7 REACTOR ,wlfq——rf GAMMA DOSE (r/hr/watt) Fig. 34. Gamma Isodose Curve Around Reactor Shield Without Shadow Disks and Dose Alomg Isodose Curve with Shadow Disks im Position. ‘01 HAGWIALAAS ONIOANA GOTHAd HOod cS6l ANP PROJECT QUARTERLY PROGRESS REPORT =0 8 = VARIABLE T { gammas /cm2/sec/Mev/watt /steradian’ 0 2 4 6 8 10 GAMMA-RAY ENERGY (Mev) Fig. 35. Lead Disks. Gamma-BRay Flux Behind spurious loss of heat are taken so that the heat generated in the fuel element 1s measured directly from water flow and temperature rise. Making allowance for the net leakage of gamma radiation into or out of the fuel element, the energy per fission can then be calculated. The experiment is well under way. IRRADIATION OF ANIMALS In cooperation with the Health Physics Division and the USAF School of Aviation Medicine, preparations are being made to expose monkeys to radiation from the bulk shielding reactor. Special water-tight cages 66 r (gammas/cma/sec/Mev/wuH/ steradian) 0 2 4 6 8 10 GAMMA-RAY ENERGY (Mev) Fig. 36. Gamma-Ray Flux Behind Lead Disks. are being constructed in which the monkeys will be placed under water near the reactor. The tentative program for the irradiation of the monkeys is presented in Table 4. In addition to the animals ir- radiated, 12 animals will be used for control immersions in the bulk shielding facility, After completion of the exposures the monkevys will be returned to the Primate Laboratory, USAF-SAM, at Austin, Texas. Particular attention is to be devoted to examination for eye cataract, L9 DWG. 45757 \0 HYPOTHETIGAL SHIELD ¢5§¢ / BOUNDARY Zé\% @ \ so0 . \(J L AN DIVIDED SHIELD ™~ s \\\ ‘ \ e \ S §§\\ \ 3 ’Q ¥ "‘b TR \ OB\ s DIVIDED SHIELD AND REACTOR & 2 ___________ Ao L) Fig. 37. Gamma-Ray Flux as & Function of Angle for Energies as Shown Be- hind Lead Disks (DSML, Y = 0°), ‘0T HAEWALJIS ONIONT qOoIddd HOA csol ANP PROJECT QUARTERLY PROGRESS REPORT Wt e e Snanin —mmmn mmaan SRl e—— —— ma —rm—— ——— ~ YPOTHETICA HI DIVIDED SHIELD \\\///H POTHE L SHIELD MOCKUP DIVIDED SHIELD AND REACTOR & T st — ——— i A R, e s et -30° Fig. 38. Gamma-Ray Flux as a Function hind Lead Disks (PSML, ¢ = 50°), 68 BOUNDARY LEAD SHADOW DISKS of Angle for Energies as 60° 50° 40° 20° ANGLE O Shown Be- FOR PERIOD ENDING SEPTEMBER 10, 1952 Fig. 39. sion (Partly Disassembled). Special Fuel Element for Measurement of the Energy Released per Table 4 TENTATIVE PROGRAM FOR IRRADIATION OF MONKEYS IN THE BULK SHIELDING FACILITY SERIES 1 SERIES 2 Exposure rate* (rem/hr) 1 0.25 Number of exposures 8 16 Time per exposure (hr) 16 8 Time between exposures (days) 7 7 Total exposure (rem) 128 32 Animals used 12 12 *One~-half of dose will be in ueutrons and one«half in gamma rays to the animals irradiated, The initial irradiations of mice, rats, and rabbits on the Cockcroft- Walton accelerator have been completed. The animals were subjected to total fast-neutron doses of 14-Mev neutrons ranging from 107 to 3 X 10!° neutrons per e¢m?. Tt is expected that the threshold for eye cataract will lie in this interval and the animals are now under observation. The results will be published by members of the Biology Division. IRRADIATION OF ELECTRONIC EQUIPMENT Additional electronic equipment from the Air Force was exposed to radiation frow the bulk shielding reactor early in the quarter. No appreciable damage was observed. The exposures have been discontinued temporarily because of conflicting higher priority work. 69 ANP PROJECT QUARTERLY PROGRESS REPORT CAPTURE GAMMA-RAY MEASUBEMENTS Of considerable importance 1in shielding calculation 1s a knowledge of the gamma rays produced by neutron capture in the various materials of the shield. Upon completion of the gamma~ray spectral measurements with the divided shield mockup, the spec- trometer(53) will be used for measuring capture gamma rays from many materials of interest. 7To provide an intense source of thermal neutrons for the production of such gamma Tays, a graphite thermal column has been installed in one corner of the pool. A 4- by 6- by 7-ft stack of AGOT-grade graphite has been installed, and the spectrometer is being adapted to the new program, (S)F. C. Matenschein, Multiple-Crystal Gamma-Ray Spectrometer, ORNL-1142 (July 3, 1952). 7. TOWER C. E. Clifford SRIELDING FACILITY T. V. Blosser Physics Division The tower shielding facility, as described in the recent proposal,(l) has been approved for construction, Calculations of structure and ground scattering to be expected in the tower shielding facility have been completed and published in an ORNL report.(?) Design criteria for the tower and bui1ldings have been prepared for sub- mission to an architect engineer who will complete the detailed design. Completion of the design contracts should require three to four months, and construction should be completed in approximately nine months. (Vg p. Blizard, Proposal for a Divided Shield Testing Facility, ORNL CF-52-4-85 (Apr. 17, 1952). (2)A. Simon and R. H., Ritchie, Background Calculations for the Proposed Tower Shielding Facility, ORKL-1273 (to be issued). 70 Orders have been placed for a majority of the reactor control com- ponents and instrumentation, and an investigation 1s being made to deter- mine the best solution to the rather difficult problem of rigging the cables for this configuration. Orders for components of the experimental instrumentation that are commercially available are being placed, and pre- liminary design of both the mechanical and electronic components of the remote instrumentation is in progress. A supersensitive fast-neutron dosimeter 1s being developed with the assistance of G, S. Hurst of the Health Physics Division, who developed the original instrument. The first approach will be to try to make use of multiple sensitive chambers operating a single electronic recording circuit, FOR PERIOD ENDING SEPTEMBER 10, 8. NUCLEAR 1952 MEASUREMENTS A. H. Snell, Physics Division Fission cross-section curves of %34 and U23% relative to U??® for incident neutron energies up to 4 Mev have been determined by using the 5-Mev Van de Graaff neutron accelerator. The N!* total cross section has been measured from 1.6 to 4 Mev on this accelerator. Some energy levels in thorium and resonance values in U?3% were observed with the time-of-flight spectrometer. FISSION CROSS SECTION OF U?3* anp y23é H. B. Willard, Physies Division The shapes of the fission cross- section curves for U2?34 and y23é relative to U23%% have been determined by using the 5-Mev Van de Graaff accelerator.(!) The statistical error was 2 1/2% when taking readings about (1) R. W. Lamphere, The Fission Cross Sections of Urenium-234 and Uranium-236 for Incident Neutron Energies up to 4 Mev, ORNL-1312 (July 15, 19§2). 3.0 e ettt e et W) 2.0 TroTAL {barns} 50 kev apart up to 4 Mev, A tritium- gas target that gave an rms energy spread varyving from 60 to 100 kev was used. Owing to foil uncertainties, however, the ordinate scale factors are in doubt to a probable error of 14% for U?** and 9% for U?3®, Thres- holds were found to be 0.37 and 0.69 Mev for U??* and u??$, respectively, Two minimums were found im each curve at different neutron energies for the two isotopes, which indicates that they are true minimums for U%3* and U236 rather than maximums in U235, TOTAL CROSS SECTION oF N!? H. B, Willard, Physics Division The total cross section of N'? has also been measured with the 5-Mev Van de Graaff accelerator. Resolution of 35 kev was obtained between 1.6 and 4 Mev., The graph of these results 1is shown in Fig. 40. UNCLASSIFIED BWG. 16355 15 20 25 3.0 35 40 NEUTRON ENERGY (Mev) Fig. 40, Total Cross Section of N4 (35-kev Resolution). 71 These measurements corroborate the mean free path values used in air- scattering calculations for the divided shield.(?) Eventual extension of these data up to 8 Mev will also be of interest, although it 1s unlikely that the extended data will indicate any gross differences from the extrapolated values, The data are sufficiently detailed for IBM calculations of air scattering as far as incident neutron energy is concerned, but there 1s still considerable uncertainty regarding the energy and angular distribution of scattered neutrons. When the latter data are available, and when the energy and angular distribution of neutrons leaving the reactor shield have been measured in the bulk shielding facility, it will then be possible to make machine calculations of neutron air (2) heportof the ANP Shielding Board, NEPA-ORNL, ANP-53 (Dct. 16, 1950). 72 These calculations should provide valuable information prior to operation of the tower shielding facility. scattering. TIME-OF-FLIGHT SPECTROMETER G, S, Pawlicki E. C. Smith P. E. F. Thurlow Physics Division The time-of-flight spectrometer has been used to measure the transmissions of thick samples of thorium oxide and depleted uranium oxide (U,0, with less than 7 ppm U??%). Energy levels were observed in thorium at 23.5, 35, 71, 127, 260, and 870 electron volts. Resonances were also observed in U?38 at 6.8, 21, 39, 62, 110, 200, and 1700 electron volts., Further details will be found in the next Physics Division quarterly report, SUMMARY AND INTRODUCTION The research on high-temperature ligquids has been directed primarily toward the production of a fuel for the ARE. The longer range work 1is principally concerned with studies of fluoborate systems and the purification of hydroxides (sec. 9). The ARE fuel is in the system NaF-ZrF,-UF,, and the most probable composition is 46 mole % Na¥F, 50 mole % ZrF,, and 4 mole % UF,, The melting point of this composition is 510°C and its vapor pressure and viscosity are tolerable. A satis- factory loading technigue involving the addition of ZrF,-UF, toan NaF-ZrF, base has been proposed. Numerous tluoride systems, with and without UF,, have been examined, and systems con- taining UF, have been prepared. The study of the reaction of fluorides with alkali metals has indicated that large quantities of NaK may be added to the system NaVF-ZrF, -UF, before producing any free UF,. Most of the corrosion research during the past guarter has centered around the determination of corrosion characteristics of fluoride mixtures containing ZrF, (sec. 10). Both static and dynamic tests 1indicate that this class of fuels and coolants 1is less corrosive than the previously tested fluoride fuel systems, an improvement which may result from better material production and testing technigues, Static tests and dynamic tests have been run with fluorides and hydroxides to determine the effects of such time, temperature, and environment, variables as additives, stress, A curious result was the apparent de- crease of fluoride attack with in- creasing temperatuve. This may have been the result of the formation of a protective oxide coating., Aside from the reduced attack that may have resulted from improved preparation techniques, the most significant advance during the last guarter came from the addition of ZrH, to the NaF-KF-LiF-UF, mixture. To supplement corrosion tests, considerable effort has been applied to fundamental studies of the corrosion mechanism, including synthesis and identification of cor- rosion products, The hypothesis that fluoride corrosion of Inconel depletes the chromium from the metal lattice and the resulting voids precipitate has been supported by further tests and observations. The metallurgical methods for the construction and assembly of the Air- craft Reactor Experiment, including welding and brazing, fabrication of control rods, and ceramic coating of radiator fins, have been developed (sec. 11). Specifications have been established on procedure and qualifi- cations for inert-arc welding of Inconel pipe and fittings for highly corrosive applications., The B,C-Fe and the Al,0,-B,C inserts for the ARE control and regulating rods have been pressed, An apparently satisfactory high-temperature ceramic coating has been applied to nickel sheet for use in a liquid metal-to-air radiator. In addition, the creep and stress of Inconel are being determined in air and argon, and the mechanical and corrosion-resistance properties of brazed joints are being investigated. Tensile tests have shown brazed-joint efficiencies as high as 92%. Heat transfer and physical proper- ties measurements on various fluorides have resulted in lessening the effort on liquid metal and hydroxide measure- ments (sec. 12). The viscosity of the ARE fuel, NaF-ZrF,-UF, (46-50-4mole %}, ranges from 20 to 7 centipoises he- tween 580 and 830°C. The vapor pressure of this fuel increases from 12 to 84 mm Hg between 807 and 940°C. 75 ANY PROJECT QUARTERLY PROGRESS REPORT Thermal conductivity, heat capacity, and density measurements have been made on several high-temperature liquids. The experimental heat transfer data for sodium hydroxide may be represented by an equation that can be evaluated to within 9% of the values normally used for ordinary fluids. A heat-momentum-transfer analysis of a thermal convection loop indicates a circulation velocity of about 0.1 ft/sec. The radiation damage program in- cludes irradiation of fluoride fuel mixtures, measurements of the effect of radiation on creep and thermal conductivity, and operation of in- reactor loops (sec. 13). Reactor irradiations of a mixture containing ZrF, indicate that no significant radiation-induced corrosion will occur at ARE intensities., An experiment to determine the rate of diffusion of Xe!®® from the irradiated static fuel mixture NaF-BefF,-UF, indicated that almost all the xenon will remain in the fuel unless flushed out. The 1700°F annealing temperature proposed for the ARE fuel tubes appears to be of some consequence in minimizing creep and thermal conductivity changes under irradiation. No radiation-induced corrosion was observed in the sodium in-reactor loop. 9. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS W. B. Grimes, Materials Chemistry Division Research in the ANP chemistry group has been concerned almost entirely with studies of fluoride mixtures for use as fuels and coolants for an aircraft reactor. Although the major effort has necessarily been directed to problems of immediate concern to the ARFE, a number of longer range studies is being carried out. The program in research and pilot- scale production of liquid fuels of high purity isstill actively followed, although responsibility for production of the large quantities of such ma- terials needed for engineering evalu- ation 1s now shared with the experi- mental engineering group, Efforts continue on definition of the optimum composition of the ARE fuel and on studies of correosion of metals by the various high-temperature liguids being considered. The previous attempts to identify chemical species in the cooled melts before and after corrosion testing have continued. 76 A study of the possible reactions of the fuel mixture with alkali metals at elevated temperature has been initiated to ascertain in quantitative fashion the effect of a leak between the fuel and moderator circuits. Detailed study of the system after reduction of UF, and/or Zr¥, has indicated that the mixture of products is quite complex even if chemical equilibrium is obtained. Some indi- cations of equilibrium among U¥F,, UF,, ZrF,, and lower valence fluorides of zirconium suggest, however, that addition of small amounts of alkali metals may be beneficial as far as corrosion 1s concerned. Detailed explanation of the systems obtained by addition of NaK to the various fluoride fuels has necessitated reactivation of the study of the preparation and the properties of UF,. Preparation of the pure material has been accomplished on a moderate scale, and studies of its chemical properties and phase eguilibriums in systems of which i1t is a component are under way. I'OR The preparation of a simulated fuel for the ARE critical experiment has been concluded, and the tubes have heen filled with powder by the Y-12 Production Division. Satisfactory progress 1is being made on the only remaining problem ~ filling one tube with a solid rod of fuel. A small program dealing with the purification of alkali hydroxides and the determination of the high-tempera- ture properties of these materials has been maintained. FUEL MIXTURES CONTAINING UF, R. E., Traber, Jr. Barton L:NM? Bratcher C. J. Materials Chemistyry Division Several binary systems have been studied because of the need for thas information 1in construction on the three-component systems containing them. The binary and termary mixtures containing aluminum fluoride that have been examined show little promise of successful application. The four-component system NaF-KF- Zr¥,-UF, appears to yield no melting points below 510°C at the 4 mole % UF, level. Since this temperature can be achieved in the NaF-ZrF, -UF, system at this uranium level, more attention has been paid to the simpler three- componenti system, The mixture containing 50 mole % ZzF,, 46 mole % Nal’, and 4 mole % UF, appears at present to be the most promising of the fuel mixtures, This material, which melts at 510°C, shows a partial pressure of ZrF, of 13 mm at 1500°F, and its viscosity seems satis- factory for the ARE (c¢f., “Heat Transfer and Physical Properties Research,” sec. 12). Slightly lower (8 mm) vapor pressure and lower viscosity values are available with PERIOD ENDING SEPTEMBER 10, 1952 the similar mixture containing 46 mole % ZrF,, 50 mole % NaF, and 4 mole % UF,. This mixture, which appears to freeze to a single solid solution of NaUF; 1in NaZrF{T melts sharply at 520°C., The final choice between these mixtures has not been made. Their preparation, handling, and corrosion problems appear very similar, Start-up of the reactor by filling the machine with Na¥F-ZrF, (corres- ponding to either of these fuels) and adding the NaF-UF, eutectic (26 mole % UF,) to bring the fuel to criticality would not be an easy operation, since a mixture melting as high as 700°C could be produced by local concen- trations and inadequate mixing. However, on the line representing 50 mole % NaF the melting point de- creases gradually from NaUF, (m.p., 710°C) to NaZrFg (wm.p., 510°C). All intermediate mixtures appear to be solid solutions of the two compounds. The mixture containing 25 mole % UF and 75 mole % ZrF, melts at 610°C and should contain about 105 lb of uranium per cubic foot at 800°C., This mixture could, presumably, be used to bring the reactor to criticality at tempera- tures above 625°C, NaF-KF-ZrF,-UF,. Data presented previously({1) showed that at the 4 mole % UF, level two compositions in the NaF-KF-ZrF,-UF, system yielded melting points below 500°C. Re- examination of the system 1ndicates that these determinations were 1n error. The lowest melting point verified in this system is at 505°C and corresponds to 38.4 wole % ZrF,, 18.7 mole % NaF, 38.9 mole % KF, and 4.0 m(}le % UF4 . (I)L. 8. Bratcher, R, %, Traber, Jr., and C. IJ. Barton, Aircreft Nuclear Propulsion Project Quarterly Progress Report for Pertiod Ending June 10, 1952, ORNL-1294, p. 84. 77 ANP PROJECT QUARTERLY PROGRESS REPORT Material containing 53 mole % KF, 5> mole % NaF, 40 mole % ZrF,, and 2 mole % UF, melts at 410°C, and it would prove to be a very satisfactory fuel for a future model with sufficient volume to contain a critical mass of this composition, NaF-ZrF,-UF,. Tentative contours for the NaF-ZrF,-UF, system are shown in Fig. 41, Although some difficulty was experienced in reproduciang data in this system, possibly because of the ease of oxidation and hydrolysis of the guadrivalent fluorides, repeated measurements have indicated that the contours are not 1in serious error, especially in the low-melting-point region, It should be possible to use up to 8 mole % UF, should this be necessary with melting points below 550°C, NaF-ZrF, -BeF,-UF,. Systems con- taining BeF, have previously been shown to possess high viscosities in the 600 to 700°C range, but a number of such systems have usefully low melting points, Some attempts have been made to decrease the melting point of the NaF-ZrF,-UF, system by addition of BeF,. Additions of BeF, to a mixture containing 50 mole % NaF, 46 mole % ZrF,, and 4 mole % UF, DWG. 16356 Ko / | Q Na3ZrF7 NaoF Fig. 41. 18 INDETERMINATE ZF‘Fq, The System NaF-ZrF4-UF4. FOR redwcé&dthe melting point to 487°C, Mixtures with 10 and 15 mole % BeF, apparently melted at this temperature. Addition of ZrF, to a ternary mix- ture of NaF-BeF,-UF, (76-12-12 mole %, respectively; m.p., 480°C) caused an initial rise in melting point. At 10 mole % ZrF,, the mixture melted at 670°C; at 20 mole % ZrF,, the melting point was 570°C, mixtures study of will be I1f the viscosities of these are usefully low, additional the four-component system attempted. ZnF,-UF,. Thermal data were obtained with a number of mixtures in the ZnF, - UF, system., The data indicate compound formation but do not show definitely the composition or melting point of the compound. The lowest melting point observed was 730 + 10°C. Since none of the alkali fluoride-zinc fluoride systems showed low melting points, further work on this system 1s not presently contemplated. ' FUEL MIXTURES CONTAINING UF, V. 8. Coleman W. C. Whitley Materials Chemistry Division Study of the phase relationships of uranium trifluoride with other fluorides is of interest for several reasons., This compound may be a product of radiation damage to fuels containing uranium tetrafluoride. Tt has a lower vapor pressure than uranium tetra- fluoride and might produce fuels with ‘more suitable liquid-temperature ranges than uranium tetrafluoride. Since i1t 1s a much less powerful oxidant, uranium trifluoride may be less corrosive to container material than uranium tetrafluoride. Uranium trifluoride has been observed 1in the product after NaK has been added to fuels containinguranium tetrafluoride, PERTOD ENDING SEPTEMBER 16, 1952 About 3000 g of UF, has been prepared by the method described in a previous report.(?’) This material, the analysis of which is higher than 99% UF,, has been used in all the studies reported below, Uranium trifluoride is quite stable at room temperature but oxidizes rapidly at elevated temperatures in the presence of traces of water vapeor or other oxidizing agents. It is therefore necessary to protect UF, from all oxidizing agents when it 1is involved in high-temperature studies. This 1s accomplished in these experi- ments by use of the apparatus dia- grammed in Fig. 42, This apparatus contains a stuffing-box seal(?) packed with an oil-free graphite- asbestos product, This seal permits the removal of air and water vapor by means of a vacuum pump; an 1inert atmosphere is then maintained while the sample i1s studied at high tempera- ture, NaF-KF-LiF-UF;. Some preliminary experiments were performed with the eutectic mixture of NaF, KF, LiF, and varying amounts of UF, up to 10 mole %. Jt was observed that the freezing point of the eutectic mixture was depressed by an amount up to 13°C. However, the molten material diffused through the graphite crucible and stirrer and cemented them quite fairmly to the stainless steel container, Examination of the product showed that some of the UF; had been oxidized, some had been converted into alkali fluoride complexes, and some remained as unreacted UF,. ' The reaction of NaF-LiF-K¥ eutectic plus UF,; with the graphite crucibles (2)W. C. Whitley and C. J. Barton, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending Septermber 10, 1952, ORNL-1154, p. 159, ' (3)Deuigned by W, C. Pivision. Tunnell of the ANP 79 ANP PROJECT QUARTERLY PROGRESS REPORT SHAFT (CONTAINING THERMOCOUPLE WEL.L)'“*-—-\“k < - THERMOGOUPLE DWG. 16357 _——PULLEY < 1 ] PACKING GLAND ——- COOLING FING————_ SN CrA e RRABT N NN e e L . ——GRAPHITE-ASBESTOS SEAL V A TUBE FOR ADDING SAMPLE—— COPPER GASKET = | | S X s e-—""UPPER FLANGE SIS RS - L OWER FLANGE VACUUM HELIUM _ér,w- -------------- —THERMOGOUPLE WELL SAMPLE——~w~nH%¥ - GRAPHITE OR STAINLESS STEEL STIRRER FEENTT - GRAPHITE OR STAINLESS Vo STEEL CRUGIBLE " fFig., 42. 80 Apparatus for Phase Studies of Fluoride Systems Containing UF,. FOR PERIOD ENDING SEPTEMBER 10, a is a matter of interest This reaction does not occur with the eutectic when UF, is used instead of UF;., Three pieces of graphite exposed to 5 mole % UF, in the NaF-LiF-KF eutectic at temperatures up to 900°C were found to contain 11.8 wt % uranium. The pronounced odor of acetylene when the mixture 1is exposed to moisture suggests that a uranium carbide 1s produced. Since the lowest recorded temperature for the reaction of uranium metal and carbon was 1200°C and most workers reported 1700°C or higher, the uranium carbide is being formed at a in 1tself. much lower temperature than would be expected. -‘Examination of the broken graphite with binoculars magnifying 30X shows the presence of red-orange, green, and white crystals. Inter- granular penetration of the carbon 1s obvious. Since the reacted mixtures smell strongly of acetylene when exposed, 1t appears that the carbide compound diffused throughout the melt. KF-UF;. Attempts to study the simpler binary system, KF-UF;, showed that the mixture containing 5 mole % UF; yvielded breaks in the cooling curve at 824 and at 711°C, However, the molten material again diffused through and reacted with the graphite crucible to such an extent that further work on this system could not be attempted in graphite containers. NaF-UF,;., Studies are now being made on the NaF-UF, system. The molten material in this system does not diffuse through a graphite crucible. Cooling curves have been taken on two mixtures, and in each case two breaks were observed: UF3 FIRST BREAK SECOND BREAK {mole %) (°c) {(°c) 5 agy 588 10 922 588 1952 These few studies suggest that a eutectic exists at 588°C, but the location is as yet indefinite., These studies will be continued. Preparation of Compounds of UF;. Attempts have been made to prepare compounds of UF; of known composition that would aid in i1dentification of unknown compounds in phase equilibrium studies. For this purpose samples corresponding to NakF-UF;, 2NaF-UF,, KF-UF,, and ZrF -UF; have been mixed and added to reaction tubes that were then evacuated, flushed, filled with helium, sealed, and heated to 900°C for about 1 hour. The materials pro- duced were examined by the technigues of x-ray diffraction and chemical Microscopy. Although the products are not all completely characterized, it appears that NaUF, is prepared in reasonable yvield by heating the first three mixtures. The samples corresponding to NaFfiUF3 show some unreacted UFS; a suggestion of a second compound, but no NaF. The 2ZNaF-UF, samples show material that 1is presumably NaUF, with unreacted NaF, no UF,, and some poorly de fined crystals that may be an additional compound. The KF-UF, sample shows a dark-red cubic compound with a refractive index of 1.556 and essentially complete reaction, as revealed by failure to detect the starting materials. It 1s likely that the red material is KUF, . The UF;-ZrF, sample appeared to be mossy-green when in large crystals and mustard~yellow when finely ground, A new phase is evident, with 15 to 20% of the UF; remaining and with some ZrF, and UF,. The refractive index of the material is in the range 1,56 to 1.58, Incomplete examination of the material 81 ANP PROJECT QUARTERLY PROGRESS REPORT does not permit evaluation of the reaction UF3 + ZrF4 ey UF4 + ZrFS or additional these materials. complex compounds of ALKALT FLUOQBORATE SYSTEMS J. G. Surak R. E. C. J. Barton Materials Chemistry Division Moore interest in fuel with melting points and viscosities lower than those under consideration for the ABE has prompted examination of the alkali fluoborate systems. It is obvious that separation of boron isotopes would be required to make such a fuel feasible and that appreciable pressures of BF; would exlist over such a fuel in the tempera- ture range proposed for aircraft reactors, Since neither of these obstacles would appear to be 1nsur- mountable, the study of phase relation- The long-range mixtures ships and decomposition pressures of some alkali fluoborate systems was started. To date, only the commercially available fluoborates of sodium and potassium have been used in this study. Figure 43 shows a schematic diagram of the apparatus employed for measure- ment of decomposition pressures of these materials. Graphite liners were tried 1nitially as containers for the fluoborates, but they were discarded 1n favor of stainless steel liners because of the adsorption of gases by the graphite, Nickel and Monel thermocouple wells proved unsatis- factory because of corrosion and embrittlement i1n a few hours of ex- posure., Stainless steel wells are, apparently, quite satisfactory. It should be noted that the corrosion described resulted from contact with commercial NaBF, and KBF,, which are 82 known to contain appreciable amounts of i1mpurities. The apparatus used for thermal analysis of fluoborate systems is the same as that used for phase studies with UF,. The preliminary studies have shown that the techniques and apparatus are satisfactory., Extension of this research with pure NaBF, and KBF, and with mixtures of these materials with alkali fluorides will be attempted. Vapor Pressures of Commercial Fluoborates. The vapor pressures of the commercial NaBF4<4) and KBF4(S) and one mixture of these two materials have been measured, The data may be expressed by the equation: log P (mm Hg) = "w5—+ B °K Table 5 gives the values for the con- stants A and B together with the values reported in the literature, which were presumably obtained with purer ma- terials. The vapor pressures observed with the commercial NaBF, are slightly lower than the values calculated from the equation in the literature, The pressures observed with the commercial KBF, are much higher than the literature values. Since the NaBF, contained about twice as much silicon as the KBF,, the discrepancy must be due to other volatile impurities in the KBF,. Thermal Data for Fluoborate Systems. Data obtained from cooling curves of the commercial NaB¥, and some mixtures are given in Table 6. The cooling curves obhtained on NaBF, -KBF, mixtures give no indication of a eutectic in this system; the melting points deviate only slightly — T General Chemical Co. Code 2240, Tech, Lot No. G291, (S)General Chemical Co. Code 2136, Tech. Lot No. GG63, FOR PERIOD ENDING SEPTEMBER 10, 1952 TO UNCLASSIFIED ARGON DWG, 163568 TO VACLIUM TO PUMP VACUUM PUMP % FULTON 10 ! INSTRUMENT fl;fiifi? " & U CVALVE ~, 1 RUBBER COPPER po— ‘ggumme //ruTusmm [‘ \____ . - THERMOCOUPLE FULTON TO RECORDER =>i/TEéTRUMENT — (= VALVE i sTaINLESS [} (" COPPER KOVAR STEEL TU&NG { SEAL TRAP— .‘\i — (LIQUID Np) N/ T0 w/ COOLING GLASS TER WATE STAINLESS fflmm STEEL J . REACTOR ————STAINLESS m— ="y " gTEEL ] LINER — ! | v . HOSKINS ~4 POT | | | S~ GRAPHITE FURNACE BLOCK Fig. 43. from a straight line connecting the melting points of NaBF, and KBF,. The thermal data for the commercial NaBF, check with the literature values gquite well, but the melting point of the commercial KBF, is considerably higher than the literature value. Thermal data on the two mixtures of NaBF4 and NaUF, that have been examined Dissociation Pressure Apparatus. are rather inconclusive. It appears that a mixture of 10 mole % NaUF¢ in NaBF, has a melting point apprec1ab1y greater than that of the NaBF, A 95 mole % NaBF4—5 mole % NaUF, ‘mixture was examined with the petro- In addition to green, crystal- amounts than graphic microscope. NaBF,, an unidentified, line phase in larger 83 ANP PROJECT QUARTERLY PROGRESS REPORT Tablie 5 VAPOR-PRESSURE CONSTANTS FOR ALKALI FLUOBORATES VALUES FOR CONSTANTS MATERTAL This lLaboratory Literature A B A B NaBF, 3820 6.68 3650 6.63 KBF, 6660 9.065 6317 8.15 90 mole % NaBF,-10 mole % KBF, 3870 6.64 80 mole % NaBF,-20 mole % KBF, 3895 6.55 10 mole % NaBF,-90 mole % KBIg 6330 8.99 Table 6 THERMAL DATA FOR FLUOBORATE SYSTEMS THERMAL EFFECTS MATERTAL " Melting Point (°C) Solid Tramsition (°C) General Chemical Co. NaBF, General Chemical Co. KBF, 90 mole % KBF4—10 mole % NaBF4 80 mole % KBF,-20 mole % NaBF, 70 mole % KBF,-30 mole % NaBF, 60 mole % KBF,-40 mole % NaBF, 50 mole % KBF4—50 male % NaBF4 381 563 535 515 498 470 450 238 277 242 233 232 215 205 would be expected from the amount of NaUF, added and an unidentified, colorless, crystalline phase were present., On the basis of this infor- mation, the existence of a complex fluoborate compound containing uranium may be postulated. 84 Rl J. DIFFERENTIAL THERMAL ANALYSIS J. Sheil R. E. Traber, Jr. Barton Materials Chemistry Division The need for a more sensitive thermal analysis procedure, especially FOR PERIOD ENDING SEPTEMBER 10, for careful examination of important samples, has been evident for some time, Two different experimental arrangements have recently been used for differential thermal analysis, with promising results. It is expected that increasing use of equipment of this type will be made., FEquipment for accurate control of heating and cooling rates will be incorporated as soon as possible; manual adjustment of these rates, however, seems to lead to use- ful information. For the smaller apparatus, a nickel block was drilled to accommodate the 1- to 2-g samples, the Al,0, reference material, and a thermocouple. The block, the differential thermocouple, the temperature recording couple, and the metal tube for introduction of inert gas are mounted together and inserted into a ceramic tube within a vertical l-in. tube furmnace. The output of the platinum-platinum-rhodium couples is fed to a preamplifier and to a Brown recorder. Tracings were obtained with two NaF-ZrF, -UF, fuels of the compositions 50-46-4 mole % and 46~50-4 mole %. The latter curve is shown in Fig., 44, Neither fuel exhibited thermal effects in these tracings above the stated melting points of the fuels. However, both materials showed thermal effects in at least two regions at lower temperatures, | An apparatus that handles larger samples and in which the sample is stirred has also been devised that has the sensitivity of the differential method. Although only a few measure- ments with this apparatus have been performed, i1t appears that use of iron-constantan couples fed directly to a Micromax recorder will be satis- factory., This apparatus will be used in the study of difficult systems such as those containing BeF,. 1952 A !!! !6359 ™ DIFFERENTIAL TEMPERATURE {mv) Q — 100 2C0 300 400 5C0 B8CO TEMPERATURE (°C) Fig. 44. Differential Temperature Curve of NaF-ZrF,-UF, (46-50-4 mole %) vs. Al,0, in Nickel-Cup, Pt-Pt, Rh Thermocouples. X-RAY STUDIES OF COMPLEX FLUODRIDE SYSTEMS P. Agron, Chemistry Division The binary NaF-ZrF,, NaZrF,-NalUF, systems, and several systems 1nvolving the trivalent uranium fluoride and possibly a reduced form of ZrF, have been examined on the x-ray spectrometer. The trivalent uranium fluoride systems are at an early stage of development, and their comprehensive treatment will require further experiments and study. Fused mixtures analogous in compo- sition to compounds found in the NaF- UF, system were examined. Interpre- tation of their x-ray patterns gave the crystal structures listed in Table 7. Na,ZrF,-Na ,UF,. The structure of NayZrF, is isomorphous with Na,UF, (tetragonal; a = 5.45, ¢ = 10.90). It may safely be inferred from the similarity of the twocrystal structures that their solid solutions should exist. This is confirmed by the observation that the Na,UF, goes 1into the lattice of the Na,ZrF, with the expected expansion of the crystal structure of the latter. ' 85 ANP PROJECT QUARTERLY PROGRESS REPORT ‘Na,Zr¥,-Na ,UF,. The results from the x-ray study of melts approximating the formulation of Na,ZrFg, shown in Table 7, have been grouped together since detailed diffraction analysis is lacking at present, Compositions ranging from 28.5 to 33 mole % ZrF, show some similarities in their dif- fraction patterns. At 33 mole % ZrF,, certain differences 1in the x-ray pattern appear when samples are prepared either vnder an inert atmosphere, in a sealed tube, or in large batches 1n the hydrofluorinator. At 28 to 29 mole % Zrk,, additional dissimilarities are evident when the sample is formed under an inert atmosphere or obtained commercially. It seems probable that solid solutions of NayZrF, and NaZrky in the Na,ZrF, structure are to some extent responsible for the variations observed. NaZr¥,-nNaUF.. The crystal structure of NaZrFg is also isomorphous with that of NaUF, (hexagonal; a = 14,72, c = 9,76), These compounds should also form solid solutions with each other, Detailed inforwmation as to the extent of the mutual solubilities in the NaZrky-NallF, system 1s of interest both for the i1nitial loading and the operation of a reactor using this type of fuel, A seriles of compositions corres- ponding to 4, 8, and 12 mole % NalUFy was made by melting NaF, ZrF,, and UF, in their proper concentrations. In addition, fused mixtures were prepared from NaZrkFy and NaUF, cor- responding to 3NaZrF,-NalUF,, NaZrF,- NaUF,, and NaZr¥;-3NaUF,. The x-ray diffraction patterns of the above materials indicate that the distri- bution of NalUF, in the lattice of NaZrF, causes a uniform increase with concentration in its lattice up to a region lying between 12 and 25 mole % NaUF,. The diffraction pattern also indicates that the major phase present at 25 mole % is the solid solution of NaUF,; 1in the NaZrF,; crystal. At the 50 mole % composition both the expanded lattice of NaZrF; and the contracted lattice of NaUFg are observed in major concentrations, demonstrating that two solid solutions are present in the solidified melt., The pattern for the NaZrFg-3NalUF, melt shows that the Table 7 X-RAY ANALYSIS OF NaF-ZrF, COMPOSITIONS COMPOSITION CRYSTAL LATTICE Q}MENSIONS CHEMICAL ' T STRUCTURE () FORMULA* 25 mole % ZrF,-15 mole % NaF Tetragonal a = 5.3, ¢ = 10.7 Naj,ZrF, 28.5 mole % ZrF4w71.5 mole % NakF** 33 mole % ZrF,-67 mole % NaF Unknown (Nazszfi)*** 50 mole % ZrF,-50 mole % NaF Hexagonal a = 14,1, ¢ = 9.6 NaZrFg . The analogy with crystal forms observed in the NaF-UF4 dystem is the criterion used for assignment of formulas, *re The range of tompositions probably involve solid solutions with resultant distortions of the Na,ZrF, lattice. 2 6 e w A definite compound in this region is supported by thermal analysis. 86 FOR PERITOD ENDING SEPTEMBER 10, 1952 predominant phase 1s the solid so- lTution of NaZrF; 1n the NalUFy lattice. In samples with lower concentrations of NaZrF¢ in NaUFg (made by fusion of NaF, UF,, and ZrF,), however, the presence of unreacted UF, would indi- cate that complete equilibrium had not been attained. It is apparent from these studies that the fluoride fuel NaF-ZrF,-UF, (50-46~4 mole %) may be represented as a solid solution of approximately 8 mole % NaUF, in the lattice of NadrF.. SPECTR@GRQPHIC ANALYSIS 0O0F FLUORIDES C. R, Baldock Stable Jsotope Research and Production Division The mass spectrometry group 1s now engaged 1n making both gqualitative and quantitative examination of reactor fuels. The qualitative examination 1is intended to vield only a rough estimate of the elements and simple compounds that are volatilized from the fuel material at temperatures obtainable 1in the mass spectrometer ovens, Ton sources currently in use can heat a sample of 1 to 2 mgtoabout 1500°C, or larger amounts, up to 50 mg, to about 700°C. A more recently developed source oven 1s satisfactory up to 2100°C. Tons of the base metals are easily obtained, and it 1is anticipated that the temperature will be adeguate to produce measurable guantities of ions from such refractory materials as the oxides of uranium. The spectrometer has recently been emploved in the determination of the contamination of UF, with higher valence compounds of uwranium that yield UF; epon heating, A 2- to 3-mg sample of UF,, which was making fuel studies, was examined at a constant temperature a time sutficiently lomg to vaporize the sample completely. The continuously used 1in for recorded peak heights were then corrected for mutual interference of UF, and UF,, and the resulting in- tensities were graphically integrated over the period of observation. The ratio of these integrated intensities over the period of observation yields the percentage of UF¢ in UF,. This resulted in a value of approximately 1%, which is in close agreement with the chemical estimate for the sample. The sensitivity of this type of analysis is very high, especially when 1isotope dilution is empleyed. The development of higher temperature ion sources will greatly expand theutility of this type of analysis. Qualitative examination of plug materials removed from the cold legs of loops in which NaF-KF-LiF-UF, had been ci1rculated the known compenents chromium, iron, and KNa,CrF, showed amounts - of should: be and also evidence of trace BFs, SiF,, X, and Hg. Tt noted that samples, namely, KNa,CrF,, were not observed with the 700°C source because the temperature obtainable has been too low to produce adequate vaporization. ' the base materials of these iron, chromium, and SIMULATED FUEL MIXTURE FOR COLD CRITICAL EXPERIMENT {Cuneo D. R. L. . Overholser Materials Chemistry Division The fuel composition for the cold critical experiment has been based on a tentative choxrce of NaF-ZrF,-U¥F, (46-50-4 mole %) for the ARE fuel. The powder fuel base was therefore established as 68 wt % Zr0O, and 21. 4 wt % NaF, ‘with 10.6 wt % carbon added to mock up the moderating effect of the missing flvoride ions. A large portion of this work was done in cooperation with the Y-12 Chemical Production Division. ' 87 ANP PROJECT QUARTERLY PROGRESS REPORT The chemically pure NaF used con- tained about 0.,1% water and was used without drying. The hafnium-free (less than 50 ppm Hf) Zr0O, contained about 0.2% water, The carbon was ground in a micropulverizer before mixing. The carbon powder passed 200-mesh screen and the Zr0, passed 100-mesh screens before the blending operation, Blend- ing of the fuel base was accomplished in two batches of about 200 1b each by rolling the mixture in a stainless steel drum for at least 72 hours. Chemical analysis for Zr(O,, Na, and C in samples from various locations in the drum was used to estimate the degree of homogeneity. The moisture content of the final fuel base was less than 0,3%, well below the tolerance value of 0, 5%, About 225 1b of the fuel mixture was prepared by addition of enriched UF, to small amounts of the fuel base to bring the uranium concentration to 10.19 lb/fts, after the fuel was packed into stainless steel tubes 1.1 in. ID and 40 in. long. The powder density of 1.8 was uniform in these tubes. The 70 fuel elements thus prepared were sealed by screw caps of stainless steel, Sixty coolant tubes of stainless steel 1/2 in, ID and 40 in. long were packed to a density of 1.8 with the fuel base, These tubes were sealed by inserting stainless steel plugs and then welding. The remaining fuel base mixture has been retained for use if it should be necessary to change the fuel concentration for another trial. It is still necessary to load the one experimental tube with a mixture of NaF-Zr¥F,-UF, of high density and with proper ratios of Na to Zr to U, All necessary materials are on hand for this operation, which will be accomplished by casting fuel slugs to fit the tube or by packing the tube 88 with sized powder prepared from a fused mixture of these materials., MODERATOR COOLANT DEVELOPMENT E. E. Ketchen L. G. Overholser Materials Chemistry Division Most of the work done during the past gquarter has involved the purifi- cation of sodium and potassium hy- droxides by methods previously re- ported, and virtually no experimental work was done on the development of new methods of purification. Production of purified sodium hydroxide, by removal of Na,CO, from a 50% agqueous solution of NaOH and subsequent dehydration under vacuum at 450°C, has averaged about 1 lb per week, The purified hydroxide continues to be of the same order of purity as previously reported,¢%¢7) with the weight per cent of Na,CO, and of H,0 being less than 0.,1. A batch of lithium hydroxide has been purified by the method previously described, (%7 and the purified hy- droxide was found to contain signifi- cantly less carbonate and sodium than batches previously purified. Several batches of Sr(OH), have been purified by the method described and previously used(%7) for both Sr(OH), and Ba(OH),. The SrCO; content of the purified Sr(OH), was found to be 0.2 to 0.3%, which is higher than the values for batches previously purified by the method. The purification of potassium hy- droxide by removing the carbonate as (6)p. R. Cuneo, E. E. Ketchen, D. E. Nicholson, and L. G. Overholser, Aircraft Nuclear Propulsicn Project Quorterly Progress Repoert for Period Ending March 10, 1952, ORNL-1227, p. 104. (7)E. E. Ketchen and L. G. Overholser, op. cit,, OBNL-12%94, p. 90. FOR in an agueous medium and de- BaCO, hydrating the filtrate under vacuum at 475°C has continued to yield material containing about 0.2% K,CO0; and an equivalent amount of barium, The lowest value for K,CO, attained so far 1s 0.1%. Because of the continual high values obtained for carbonate, this method has been shelved 1n favor of the method using potassium. This method, described previously, (7} jin- volves the interaction of water with pure potassium and dehydration of the potassium hydroxide solution. Four runs have been made since a lavger vacuum dry box became available for trans ferring the potassium to the reaction vessel. Approximately 1/2 1b of potassium hydroxide has been pre- pared per ‘run, and the K,CO; content has varied from 0.05 to 0,15%. It appears possible to prepare potassium hydroxide containing 0.1% or less of both K,C0, and H,0 by this method. The only impurity present in determina- ble amount in the potassium used 1s sodium, which occurs to the extent of 0. 2%. CDOLANT DEVELOPMENT I. M. Bratcher R, E. Traber, Jr, C., J. Bartpn - faterials Chemistry Division Investigation of nonuranium fluoride systems of interest as possible coolants or as a base for preparing fuel mix- tures has continued. Although ZrF, was a compounent of a number of the systems being considered, the field of interest was broadened by the addition of BeF, and AlF; to the list of com- porents., Zinc fluoride received very little attention., Few of the systems have been investigated thoroughly enough to warrant definite conclusions on their usefulness, but preliminary tests have not, for the most part, been very encouraging. PERTOUD ENDING SEPTEMBER 1¢, 1952 RbF-ALF,. RbAIF (%) and Rb,AlF, (%’ have been described in the literature, A melting point of 985°C was reported for the Bb;AlF, compound; the mixture corresponding to this compound, pre- pared 10 this laboratory, showed a melting point of 925°C, The fused mixture has not been analyzed; so the reason for the difference is not known at present. The data obtained on thisg entire system, although incomplete, indicate a eutectic melting at 525°C, Cooling curves for mixtures containing about 40 mole % AlF, also show a thermal effect, which 1s as yet un- explained, at approximately 475°C, RbAl¥, appears to melt at 450 + 10°C. NaF-KF-A41F,. Partial phase diagrams for the binary systems NaF-AlF, and KF-AlF; are given in the literature. Meither diagram extends beyond 50 mole % AlF, because of the volatility of AlF,. The fact that the KF-AlF, system was reported to have a eutectic melting at 565°C led to a study of the ternary system, No low-melting-point compo- sitions bave been found among the few mixtures tested thus far, but the study 1s continuing, | NaF-ZrF,-AlF,;. FEven though the use of two volatile components ina ternary system seems rather unpromising, a number of compositions in the NaF-ZrF, ~ AlF; system has been subjected to thermal analysis., The addition of 5 mole % AlF, to the NaF-ZrF, eutectic and of 5 mole % Zr¥, to the NaF-AlF, eutectic apparently raised the melting point. The cooling curves give no indication of the existence of a eutectic melting at lower than 460°C in this system. | NaF-Zr¥F,. Na,ZrF; preparedin an HF atmosphere melted at 625 £ 10°C. The 8 ' . ( )X-Ruy Diffraction Patterns, A.S.T.M. Standards, First Supplement, August 1945, : (Q)International Critical Tables of Numerical Data, McGraw-Hill, New York {1926-33}. ' 89 ANP PROJECT QUARTERLY PROGRESS REPORT addition of NaF raised the melting point of the Na,ZrF, very sharply to a maximum of 830°C near 25 mole % ZrF,, which indicates the probable existence of the compound NazZrF,. The prepa- ration containing 57 mole % Nal and 43 mole % ZrF, melts at 480°C and appears to be the lowest melting composition of this system. This material expands sufficiently on solidification to break graphite crucibles. Neither Na,ZrF, nor NaZrF, has been found to act this way. ZrF,-BeF,. Melting points of mix- tures 1in the ZrF,-BeF, system were of interest in connection with the NakF- ZrF,-BeF, system. Three mixtures were prepared containing 25, 50, and 75 mole % ZrF,. Their melting points were 650, 740, and 825°C, respectively. These three points, together with the melting point of BeF, (543°C), made almost a straight line. The melting point of ZrF, has not yet been de- termined. The extrapolation of the liguidus curve to 100% ZrF, gives a value of about 910°C, HYDROLYSIS AND OXIDATION OF FUEL MIXTURES R. P. Metcalf Materials Chemistry Division In connection with the oxidation of fuel mixtures, the reaction of uranium tetrafluoride with oxygen 1s of Previous 1nvestigators of (10, 11) hayve attempted interest, this reaction to estimate the free energy change 1in the reaction 2UF, + 0, ——> UF, + UO,F, . S. Fried and N. R. Davidson, The Reaction of UF4 with Dry 02; A New Synthesis of UFG‘ AECD- 2981 (May 1945}, 11)5. S. Kirslis, T, S. McMillan, and H. A. Bernhardt, The Reaction of Uraniur Tetrafluoride with Dry Oxygen, K-567 (March 15, 1950), 90 The availability of an alternate method of estimating the heat of formation of UO,F, has prompted a new calculation using an extrapolation of recently published data on UF, and UQ,F,. Heat of Solution and Heat of For- mation of Uranyl Fluoride. For the calculation of the free energy change for the reaction, the heat of for- mation of uranyl fluoride was required, To aid in estimating this quantity, the heat of solution of uranyl fluoride was measured, A glass Dewar flask was used as the calorimeter, with 0,5 M perchloric acid as the solvent., The determi- nation of the heat capacity of the calorimeter was based on the known heat of solution of sodium nitrate, The result obtained for the heat of solution was UO, F, (crystal) + HY — UO,F* + HF, AH298 = ~-5.3 £ 0.5 kcal/mole . From an estimate of the heat of ionization of the complex ionU02F+,(12) together with known data, the heat of formation of uranyl fluoride was calcu- lated to be U(crystal) + 0,(gas) + F,(gas) ——> UOQ,F,{crystal) , AN = ~-391,3 + 3 kcal/mole . 208 Free Energy Change for the Reaction of Uranium Tetrafluoride with Oxygen. of the thermodynamic for O, and UF, at high temperatures, calculated from spec- troscopic data, are 1n the literature. Values of the functions for UF, and UO,F, were estimated by extrapolation The values functions (12) K. A. Kraus, perscnal communication. FOR PERIOD ENDING SEPTEMBER 10, of published heat capacity data at lower temperatures. Results of the calculations are given in Table 8. Table 8 STANDARD FREE ENERGY CHANGE AND EQUI- LIBRIUNM CONSTANT OF THE REACTION 2UF, + 0, —> UF, + UO,F, T (°K) AF (kcal) K 298.16 -10.7 7 x 107 500 -10.6 4 x 10* 800 -9, 4 4 x 10° 900 -8.4 1 x 107 1000 -7.1 4 x 10 1100 -5.6 1 x 10t 1200 -3.7 5 1300 -1.3 2 For temperatures below 1000°K, the principal uncertainties in the calcu- lations probably arise {from the experimental errors in the heats of formatien of UF,, UFg, and UO,F,. The resulting uncertainties in AF are of the order of +8 kcal, At temperatures above 1000°K, uranyl fluoride 1s known to be un- stable, with one of the principal de- composition products being uranium tetrafluoride.('?? Tt has been pre- dicted that uranium hexafluoride would also be unstable above this tempera- ture, (%) Thus, the present calcu- lations are not expected to have much significance for temperatures higher than 1000°K. Subject to these limi- tatiouns, the calculations confirm the (IB)F. Vaslow and A. 5. Newton, Chemistry of Plutonium Report for Period April 10 - May 10, 1944, CK-1498, p. 10-11, (14)L, Brewer, .. A. Bromley, P. W. Gilles, and N. L. Lofgren, The Thermodynamic Properties and Equilibria at High Temperatures of Ureniunm Helides, Qxides, Nitrides, and Carbides, MDDC- 1543 {(Sept, 20, 1945). 1952 earlier conclusion that the oxidation proceeds at low temperatures except where the reaction rate is too slow. The calculations alse indicate that AF is somewhat more negative than the previous estimates and that the reaction proceeds at temperatures 200 deg higher. However, the dif- ference is still within the limits of error of the heats of formation employed. FUEL PURIFICATION RESEARCH F. F. Blankenship F. P. Boody B. E. Thoma, Jr. C. M., Blood Materials Chemistry Division Purified of fluoride mixtures have been prepared satis- factorily in a routine manner by the hydrogenation-hvdrofluorination tech- nique previously described, (1%) Purification of small samples of enriched fuels for radiation-damage testing has been accomplished by a modification of this technique. samples Purification of pure zirconium fluoride from oxide-bearing material by sublimation has been discontinued, except for special samples, since the Y-12 production facility 1s now 1in positicn to supply the pure material, Preparation of ZrF, in a melt con- taining NaF by ligquid-phase hydro- fluorination appears to be very promising. Purification of Molten Fluorides. A previous document®!®) from this laboratory described the apparatus and techniques used in the high-temperature treatment of fluoride melts with hydrogen and HF. By application of this procedure on a routine basis, a total of about 80 kg of purified fluorides was produced during the (IS)C. M. Blood, F. P. Boody, A, J. Weinberger, and G. J. Nessle, op. c1t., OBNL-1294, p. 97. 91 ANP PROJECT QUARTERLY PROGRESS REPORT guarter, This material was, for the most part, mixtures containing both UF, and ZrF,, but a few fuel-solvent materials possible moderator-coolant mixtures containing BeF, were included, uranium-free and some Although replacement of the trans- fer lines and sintered nickel filters must occasionally be made, the nickel apparatus has performed 1n a very satisfactory manner., Results from this program have justified con- struction of similar apparatus of the same size, as well as equipment to produce 50-1b batches of fuels by this technique, Corrosion of the equipment by the fluorides has the reactor life, of some stainless steel sections 1n the reactor below the liquid led to constiderable contamination of three batches of product with 1ron, chromium, and nickel, Samples prepared in all- nickel equipment generally show less than 100-ppm nickel, Careful exami- nation of the product by chemical microscopy rarely reveals the presence of oxides in detectable amounts. not seriously limited Accidental inclusion The removal of sulfur by reduction of oxidized compounds and evolution of H,S appears quantitative; chemical analysis reveals no trace of this element, Although no analytical data are availlable, 1t appears likely that the traces of chloride present in the ZrF, presently being used are also removed. Preparation of Pure Zirconium Tetrafluoride. FEase of hydrofluori- nation of Zr0, by gaseous HF at elevated temperatures 1s a function of previous history of the Zr0,. Zir- conium hydroxide may be hydrofluori- nated quite easily; Zr0O,, which has been calcined at 800°C or higher, however, reacts incompletely with HF at temperatures as high as 700°C even 92 when reaction times of the order of 20 hr are used.{!®) Unfortunately, the hafnium-free product from the Y-12 production plant is purified by pre- cipitation with phthallic acid and must be calcined at high temperature to remove the last traces of carbon. Accordingly, plans have been made to produce the hafnium-free ZrF, by hydrofluorination of ZrCl, prepared from an oxide by the Bureau of Mines. The difficulty in obtaining com- plete reaction of Zr0O, with gaseous HF is probably due to a combination of several factors. Tt is difficult to obtain good contact of the gas with the particulate solid in most apparatus. The reaction product tends to coat the surface of the unreacted oxide and prevent reaction. In addition, high concentrations of HF are achieved only at elevated pressures. Jn an attempt to improve the con- version of ZrQ, to ZrF,, a series of experiments in which the reaction 1is performed in the liquid state has been performed. The Zr0,, which had been calcined at 800°C, was charged, along with NaF, into a graphite-lined nickel apparatus similar to the fuel purifi- cation apparatus previously described. The HF was admitted at atmospheric pressure and the temperature was slowly raised until liquid phase was present, Evolution of HF accompanied by water of reaction occurred at temperatures above 110°C, and the temperature was slowly raised over a period of 3 to 4 hr until HF evolution ceased at about 500°C. At this stage, HF was bubbled through the fused-salt mixture con- taining suspended Zr0O, while the temperature was raised to 800°C and maintained there for 2 to 3 hours. By this technique, good contact and high HF concentrations are assured at the J. L. Williams and B. 5. Weaver, Prepara- tion of Zirconium Tetrafluoride Progress Report No. 1, Y-619 (June 15, 1950),. FOR low-temperature range. In the high- temperature range the Zr¥, produced dissolves in the melt and leaves the Zr0, free to react. The optimum conditions for smooth and complete conversion of the Zr0, have not yet been demonstrated. Jt has been demonstrated, however, that complete conversion of l-kg charges can be accomplished in 8 to 10 hr and with three to four times the stoichio- metric quantity of HF. These experi- ments are being continued; if the present optimistic results are verified, the technique may be extended to preparation of cryolites and Bel, mixtures, Preparation of Pure Aluminaum ¥luoride. Pure, anhydrous AlF,; has been prepared by dehydration of com- mercially available AIF,-H,0 under a flowing atmosphere of HF in nickel equipment at 300 to 600°C for 6 hours. Several batches of material, totaling 7 1b of anhydrous product, have been made available for phase~equilibrium studies. REACTION OF FUELS WITH ALKALI METALS R. J. She1il E. E. XKetchen N. V. Smith D. C. Hoffman ¥. F. Blankenship Materials Chemistry Division P. Agron Chemistry Division The reducing action of alkali metals on UF, and ZrF, is well known, and under proper conditions the free heavy metals can be produced readily by this method. However, when rela- tively small amounts of alkali metal are added to fluoride mixtures the reaction proceeds not to the metal but to a complex series of intermediate reduction products, including UF,. PERIOD ENDING SEPTEMBER 10, 1952 On the basis of available thermo- dynamic estimates, the reduction of UF, to UF; is more likely than that of ZrF, . However, when the UF, is dis- solved in NaF, where strong com- plexing may occur, and when a large excess of ZrF, 1s present, the re- duction of UF, is not necessarily the first step in the process. In fact, preliminary evidence indicates that, in the presence of excess ZrF,, up to 40% of the NaK theoretically re- quired to reduce all the UF, present may be added without preoducing detecta- ble amounts of free UF,, Although much remains to be learned about the reaction products of NakK with ZrF,-bearing fuels, the following conclusions, typicel of material con- taining 30 mole % Na¥F, 46 mole % ZrF,, and 4 mole % UF,, are of interest, Slow heating of sealed capsules containing the powdered, purified fuel with a few weight per cent of NaK shows that the rapid exothermic reaction takes place at about 250°C, In a typical case, 1 g of NaK with 50 g of fuel yielded a sudden temperature in- crease from 240 to 286°C, The reaction will, presumably, be very rapid at all temperatures above 250°C, Although there is evidence from other sources that UF, forms compounds with NaF, KF, and other fluorides, the bulk ef this material formed such capsule tests appears on examination of the cooled melts as pure UF, at the bottom of the capsules. The solubility of UF, in the reaction melt at temper- atures from 600 to 800°C is not vyet known, but it appears to be relatively low at temperatures well abeve the freezing point of the mixture. in 1f one equivalent of NaK is de- fined as the guantity required to reduce the UF, present toUl,, addition 93 ANP PROJECT QUARTERLY PROGRESS REPORT of less than 0.4 equivalent produces a yellowish tinge in the green, solidified melt., The yellow color replaces the normal green color of the fuel by the time 1 equivalent has been added. Segregated UF; in a yellow matrix results when 2 equivalents are used; the solid solution of NaUF; in NaZrF,, typical of the unreacted fuel, is considerably altered, and a yellow, fine-grained phase, some of which forms a cubic crystal of n = 1,476, has appeared. When 4 equivalents are added, the vellow, cubic phase, a colorless cubic phase of n = 1,430, considerable UF;, an opaque phase that may be metal, and Na,ZrF, appear in the melts. When 8 equivalents are added, the yellow phase and the solid solution are not detected. Some Na,ZrF. and the opaque material are present, and all the uranium seems to be present as UF,. The material appears to the eye as a gray-white matrix containing UF,. Examination of the cooled and ground melts for production of H, on reaction with dilute acids shows that the re- ducing power is much less than expected from the amount of NaK added. This is surprising since UF; yields the theo- retical volume of H, after similar grinding and handling in air. X-ray diffraction studies have indi- cated that prominent but unidentified lines have disappeared from spectra run on the same sample at a later date. It is possible that a low-valence compound of zirconium that 1is unstable when handled in air may be responsible for this behavior. Although it 1s evident that asnpdden large leak of NaK at elevated tempera- tures into the fuel stream would have serious consequences, it is not evident that a leak small enough to permit 94 chemical equilibrium to be maintained would be disastrous,. If reducang conditions are desirable in the fuel, 1t may be possible, on the basis of these data, to provide up to 0.2 equivalent of reducing power (fluorine deficiency) without precipitation of UFy,. It may be that the corrosive behavior of these fuels could be further improved in this fashion., A considerable effort will be maintained on this complex problem. SERVICE FUNCTIONS J. Truitt C. J. Barton Materials Chemistry Division The need for preparation of small batches of fluoride mixtures for standards in x-ray diffraction or petrographic microscope work, for enriched uranium fuels, and for other purposes still exists. About 30 batches of fluoride mixtures ranging in size from 10 g for enriched uranium fuels to about 200 g were prepared during the last quarter. In addition, various containers were filled with fluoride mixtures, as shown in Table 9, Fluoride mixtures were removed from 406 corrosion-testing tubes by melting in an inert atmosphere; the recovered melts were prepared for analysis by grinding to a fine powder, Table 3 CONTAINER-FILLING SERVICES TYPE OF CONTAINER NO. OF CONTAINERS FILLED Cyclotron tubes 59 Corrosion tubes 286 Radiation-damage capsules 5 Heat-capacity capsules 4 Vapor-pressure pots 3 FOR PERIOD ENDING SEPTEMBER 10, 10. 1952 CORROSION RESEARCH W. B. Grimes, Materials Chemistry Division W. D. Manly, Metallurgy Division H. W. Savage, ANP Division The emphasis in dynamic corrosion testing i1in thermal convection loops has shifted from testing the fuel NaF-KF-LiF-UF, to testing various coolants and fuels containing zirconium fluoride. The coolants and fuels containing zirconium fluoride are not so corrosive to Inconel as the NaF-KF- LiF-UF, fuel. This is probably due to the more careful preparation and handling prior to the corrosion tests. With the zirconium fuels, the corrosion in the hot leg is approximately 5 mils after 500 hr of operation compared with the 10- to 15+-mil depth of corrosion that was previously ex- perienced. The effect of different cleaning procedures on the corrosive action of the fluorides on Inconel 1is being studied. Static tests and modified dynamic tests have been used to de- termine the effect of time, tempera- ture, and various additlives on the corrosion behavior of the fluorides. It has been found that most of the cerrosion occurs in the first 500 hr of operation. Strange results were obtained in corrosion testing at extremely high temperatures; the amount of corrosion appreciably decreased. This was probably due to the formation of a thin film of uranium oxide on the surface of the test capsule. The additionof strong getters has markedly decreased the corrosive action of the fiuorides. The most striking example was the addition of zirconium hydride to the NaPF-KF-LiF-UF, fuel that was circulated in a thermal convection loop. Upon examining the thermal convection loop after 500 hr at 1500 °F, it was found that the loop suffered the least amount of attack of any thermal couvection loop run with the fluorides; the maximum depth of attack was 0,5 mil. Tests of chemical compatibility of several possible moderator coolants with the beryllium oxide moderator material have shown that Zr¥, -bearing melts are not useful in this connection. Sodium, NaK, and molten lead appear to be satisfactory, as do fused salts containing beryllium fluoride. Tests on the mass transfer properties of lead have been conducted 1n which additions of 3, 5, and 50% sodium were made to lead. These additions had a very marked effect on the mass transfer characteristics and the corrosive action on the container materials. Additional static and dynamic tests were run in an attempt to find in- hibitors for the mass transfer and corrosion exhibited in the metal- hydroxide systems. Additional tests have shown that the mass transfer experienced in nickel-hydroxide systems can be stopped by the use of purified hydrogen over the system. In additien to the program of empirical corrosion testing, a number of studies designed to explain the 65 ANP PROJECT QUARTERLY PROGRESS REPORT corrosion mechanismhas been conducted. Careful examination by physical and chemical means has been made of the corrosion products formed during large-scale dynamic corrosion tests, and studies have been made of the reactions of structural metals with high-temperature liquids in simple systems and the possible reactions of molten fluorides and hydroxides under applied potentials. A considerable number of complex fluorides of the structural metals has been prepared and characterized to assist in the identification of corrosion products, A study of the compounds formed by the interaction of nickel in the hydroxides has shown the appearance of two new compounds, NaNi0O, and LiN10,. These compounds, which have been named sodium nickelate{(III) and lithium nickelate(II1), have been 1identified by chemical analysis for nickel and the alkali metals; by several chemical reactions, including a quantitative determination of the oxidizing power; by the reaction that gives a known trivalent nickel com- pound; and by complete crystal structure analysis. PARAMETRIC STUDIES OF FLUORIDE CORROSION A. des Brasunas E. E. Hoffman I.. S. Richardson R. B. Day D. C. Vreeland I.. D. Dyer Metallurgy Division F. Kertesz Materials Chemistry Division The efforts to understand and minimize fluoride corrosion have logically led to a study of the many parameters that enter into corrosion phenomena and corrosion tests, In- cluded among these parameters are temperature of test, length of test, effect of residual stresses in the 96 corrosion specimen, carbon content of the specimen, effect of additives, and effect of pretreatment, Each of these parameters, except for the carbon content of the specimen, affected the corrosion in some measur- able manner. Additives were both good and bad, pretreatment was bene- ficial, and corrosion decreased with increasing time or temperature. Temperature of Test (A, des Brasunas, L. S. Richardson, D. C. Vreeland, E. E. Hoffwman, R. B. Day, and L. D. Dyer, Metallurgy Division). Inconel was tested at a series of temperatures for 200-hr periods in molten fluoride NaF-KF-LiF-UF, (10.9-43.5-44,5-1.1 mole %). A rather surprising effect was noted; at the higher temperatures void formation became less prominent, as shown by Fig. 45 and by the data of Table 10. The analyses of the fluoride baths, after the tests, were 1n agreement with the observations given in Table 10; in the test at 800°C, 1800 ppm chromium was detected in the bath, whereas the fluoride of the test at 1300°C yielded only 65 ppm chromium. The nature of the film detected on the Inconel surface of the high-temperature tests has been identified as U0, by x-ray diffraction, Length of Test. A series of tests was made to determine the extent of corrosion as a function of time. Inconel test specimens were prepared for testing at several time intervals ranging up to 3000 hr in the seesaw test apparatus, 1n which the cold zone was at about 800°C and the hot zone at about 650°C, The test data are summarized 1in Table 11 and plotted in Fig. 46. These data indicate that most of the corrosion occurs in the first 500 hr; beyond this time the additional amount of attack 1s quite small. FOR PERIOD ENDING SEPTEMBER 10, 1952 UNCLASS I F IED v.7183 = in jo> ‘.' UNCLASSIFIED UNCLASS I FIED Y.7184 : Y.7185 £ AN Fig. 45. Depth of Veid Formation as a Function of Temperature in Inconel by NaF-KF-LiF-UF,. {(a) 900°C. {(b) 1000°C. (¢) 1100°C. {(d) 1200°C. Un- etched. 250X ' .97 ANP PROJECT QUARTERLY PROGRESS REPORT Table 10 CORROSION OF INCONEL BY NaF-gKF-LiF-UF, AS A FUNCTION OF TEST TEMPERATURE © TEST TEMPERATURE DEPTH OF VOID FORMATION THICKNESS CHANGE (°C) (mils) OF SPECIMEN 800 20 None 900 15 None 1000 10 None 1100 10 None 1200 0.5* None 13060 0.2* None *Thin film of UO2 observed on metal surface. Table 13 EFFECT OF TIME ON THE DEPTH OF FLUORIDE ATTACKX ON INCONEL AS DETERMINED BY THE SEESAW TEST METALLOGRAPHIC OBSERVATIONS L TEMPERATURE ( °C Depth of Void TIME NO., OF e (‘ml Formation in TEST (hr) CYCLES Hot Zone COLD Z20NE Hot Zone Cold Zone Average Maximum (mi]s% {mils) S5F-3 66 18, 500 g12 618 0.5 1 No evidence of reaction SSF-4 115 30, 200 780 600 1 2 No evidence of reaction SSF-5§ 210 §5, 200 780 630 2 5 No evidence of reaction SSF-8 500 132,000 750 610 3 7 Metallic deposit 0.5 mil thick SSF-8 750 198,000 816 620 3.5 6 Metallic deposit | 0.5 mil thick SSF-17 3000 186,000 300 560 4 13 Metallic deposit 0.5 mil thick 98 FOR PERIOD ENDING SEPTEMBER 10, tp— P {2 e o e D J,/ 0 ~w§§¢fifl ---------------------------- l‘I" gk Ll e el L et e - % Q0 o o L . & BYERAGE et b 500 1000 1500 TTEOGD T 3500 3000 TE (hr) Fig. 46. Corrosion vs. Time for Inconel in NaF-KF-LiF-UF, from Seesaw Tests at 800°C. Residual Stresses in Specimen. From quench-annealed flat stock a series of cold-worked Inconel speci- mens were prepared that ranged from 0 to T1% reduction 1in thickness. These specimens were then corrosion tested in NaF-KF-LiF-UF, for 100 hr at 815°C to determine possible differ- ences 1in corrosion behavior. The test results listed in Table 12 show that there is a small, but never- theless perceptible, effect. 1952 There appears to be maximum cor- rosion in the range of 5 to 25% reduction in thickness. Such results are not surprising, since hole for- mation, which is believed to be caused by the diffusion of chromium from the metal into the fluoride bath, is accelerated by the presence of cold work. The recrystallization tempera- ture range for Inconel is about 1500°F, which 1s the test temperature, and hence heavily cold-worked specimens recrystallize in a relatively short time. Therefore specimens that have not been cold worked and specimens that have been heavily cold-worked may be essentially stress free for most of the period of the corrosion test, and the intermediate stress range may not result in recrystal- lization throughout the exposure period. Carbon Content of Specimen. To establish the effect of carbon content, 1f any, on the corrosion behavior of face-centered-cubic metals in NaF-KF- LiF-UF,, two seesaw test specimens were run simultaneously so that identical time and temperatures would Table 12 VARIATION IN SUBSURFACE VOID FDRMATION AS A RESULT OF COLD WORK REDUCTION IN THICKNESS | WEIGHT LOSS | DEPTH OF stfn%s]iJsBFACE VOIDS | mpicKNESS CHANGE 0 0.0020 1 None 5.5 : 0.0043 2 None 11.5 0.0036 1.5 None 26.0 0.0036 1.5 None 50.0 0.0023 1 None 71.0 0.0028 1 None 99 ANP PROJECT QUARTERLY PROGRESS REPORT prevail. One specimen was a regular, type 304 stainless steel (containing 0.08% C), and the other was an extra- low-carbon type 304 stainless steel (0.006% C). The temperature conditions for both specimens were (20°C in the hot zone and 600°C in the cold zone. The duration of the test was 190 hr, during which time the specimens were subjected to 51,150 cycles. Examination of the specimens after testing gave 1dentical results, namely, 5 mils of void formation in the hot zones of both. No attack or deposition was detected in the cold zones. Corrosion Inhibitors, Static and dynamic corrosion tests at 816°C for 100 hr have been continued with various agents being added to check the possibility of the additions acting as corrosion inhibitors, Particular attention was given to additions of chromium because of the part chromium plays in the fluoride corrosion of Inconel (cf., paragraph on ‘Mechanism of Fluoride Corrosion,” this section). Tests have been run with 0.17% additions of sodium (previous addition tests were with higher percentages ot sodium) and also ad- ditions of MnO,, CrF,, NiF,, and FeF,. In the case of the sodium addition and also in the previous tests with additions of lithium, potassium, calcium, titanium, and manganese, 1t was thought that these elements might possibly act as getters to reduce the available amount of oxygen in the system and thus help to reduce cor- rosion. The MnO, additions were made to determine whether oxygen added to these tests in this form would in- crease corrosion, 100 " The metal fluorides (and also sodium, lithium, and potassium) were added with the thought that perhaps a reaction of the following typemight be the corrosion mechanism: 2¢F + [ — BF, + 2a , where a could represent sodium, lithium, or potassium and S could represent iron, nickel, or chromium. If this were the case, then additions of iron, nickel, or chromium fluorides and also sodium, lithium, or potassium should tend to drive the above re- action to the left and help reduce the corrosion. However, additions of the iron, nickel, and chromium fluorides had no effect in reducing corrosion. It should be mentioned that the purity of the salts used is not known. Ad- ditions of sodium, lithium, and potassium do seem to reduce corrosion, but, as mentioned above, their action may be that of getters rather than that of participators in the reaction. In Table 13, the additions (in both the previous and the most recent tests) are classified according to the results of the tests. It should be emphasized that the classifications are based on tests that were run with certain percentages of these addition agents, and 1t 1s conceivable that different amounts might result in different corrosion results, Table 14 shows the results of the most recent tests in which addition agents were used. The beneficial effect of the addition of 1/4% chromium to the fluoride bath i1s demonstrated in Fig, 47, which shows the corrosion of Inconel by NaF-KF-LiF-UF, with and without the chromium additive. The absence of subsurface voids when chromium is added agrees with the hypothesis that the voids are caused by chromium depletion of the metal by the fluorides. FOR PERIOD ENDING SEPTEMBER 10, 1952 Tab}e 13 RELATIVE EFFECTS OF ADDITION AGENTS ON CORROSION BY FLUORIDES AGENTS AGENTS AGENTS WITH LITTLE REDUCING INCREASING OR NO EFFECT oN COBROSION CORROSION CORROSION Na MnO3 CrF, L1 NiF, NiF, + CrF, + FeF, K Fel, C Ca Nal Mo Ti KI Si Mn Lii0, Zn Be KC1 W Al NaCl Fe U KBr Zn Cr Mn Cu Ag NaH Table Pretreatment. In an attempt to evaluate the influence of the pre- treatment procedure, parallel dynamic tests of material containing NaF-KF- ZrF,-UF, (36.6-14.0-45.6-3.8 mole %) were made with samples prepared 1in graphite and in nickel containers without hydrofluorination and in nickel containers with hydrogenation and hydrofluorination. The material that had been hydrofluorinated was somewhat superior, although in these tests no marked differences were noted. In other tests the fluoride mixture NaF-KF-LiF-U¥F, was treated by bubbling hydrogen through it for 2 hr while the molten fluoride was held at a tempera- ture of approximately 980°C. There appeared to be no significant re- duction in corrosion of Inconel, type 321 stainless steel, or 4 nickel when the hydrogen-pretreated fluoride was used. 14 EFFECTS OF VARIOUS ADDITIONS T0 NaF-KF-LiF-UF, IN STATIC CORROSION TESTS FOR 100 hr AT 816°C IN VACUUM MATERI AL ADDITION R e METALLOGRAPHIC NOTES Inconel 0.17% Na No attack Type 304 stainless steel 0.17% Na No attack Inconel 10% Cer 4 Subsurface voids Inconel 10% NifF, 7 Subsurface voids Inconel 10% Fek, 9 Subsurface voids Inconel 3 1/3% CrF, 3 1/3% NiF, 3 1/3% Fel, 2 Subsurface voids Inconel 10% MnO, 9 - Subsurface voids and intergranular penetration Type 309 stainless steel 10% MnO, 4 Subsurface voids and intergranular penetration 101 ANP PROJECT QUARTERLY PROGRESS REPORT IR (i ASSIFIED § I v_7180 {0/ . L | - -w—‘k‘- - . ( jn.‘ '. i cfi..s S UNCLASS I FIED @ e .00 B - ., . . ' - . - - . * T v - a ¢ . : -y » - . . - i ~: e .\ e L!f e e ) . “ ’ ~ ] . . « 0 Y - =T - ! - - v T * . - - . « . - S CE i , AU S oo TN . - ' ‘ - . i T - .l : ¢ - . ‘ ’ 3 [ . . - v : - 4 a . {5/ ¢ oY - . " - - NS L s - . : .o .- P . e . N . L » » ’ - a4 Y s P “ g - - e R . - . . + Fig. 47. Corrosion of Inconel in Seesaw Tests by NaF-KF-LiF-UF4 With and Without Added Chromium After 200 hr at 780°C. (a) No chromium added,. (6) Chromium added, 1/4%. 250X 102 FOR SEESAW AND STATIC TESTS WITH FLUORIDES CONTAINING Zr¥, C. R. Croft N. V. Smith R. E. Meadows H. J. Buttram Materials Chemistry Division Ten different Zr¥F, -bearing fused salts ranging from 58 to 44 mole % ZxF,, 0 to 52 mole % NaF, 0 to 46 mole % KF, and 0 to 4 mole % UF4 have been tested in sealed capsules of type 316 stainless steel and Inconel under static and dynamic conditions. The static tests were maintained for 100 hr at B00°C. The dynamic tests were performed with the tilting furnace previously described;(!) the hot-zone temperature of 800°C and the cold-zone temperature of 650°C were maintained for 100 hr, during which time the fuel made 24,000 cycles. Comparison of results from the two test procedures substantiates previous statements that corrosion, as measured by penetration, is significantly greater in the dynamic test and that mass transfer of small quantities of metal is often observed in the dynamic tests. However, even the dynamic tests are not sufficiently severe to permit evaluation of small differences in corrosion behavior. The zirconium~-bearing fluoride mixtures were shown to be less cor- rosive than previously tested NaF-KF-~ LiF-UF4 mixtures. Variation of corresiveness with composition over the concentration ranges listed above could not be ascertained with accuracy. In general, corrosion in each case (as 1ndicated by penetration data) was scattered and the depth of attack ranged from 0.5 to 2 mils on Inconel, with slightly higher values for (Dp. c. Vyeeland, R. B. Day, E. £. Hoffman, and L. D. Dyer, Aircreft Nuclear Propulsion Project Quarterly Progress Report for Period Ending March 10, 1952, ORNL-1227, p. 120. PERIOD ENDING SEPTEMBER 10, 1952 type 316 stainless steel. Weight changes of the specimens after test were slight and metal dissolution in the fuels was less than with fuel mixtures tested previously. Slight penetration or surface roughening was found in the hot zones of the capsules, and occasional small metallic deposits were observed in the cold zones. The corrosion observed appeared independent of the UF, concentration over the range studied. In a series of static exposures of Inconel to NaF-ZrF,-UF, mixtures for 100 hr at 1000°C the only corrosion observed was a slight surface deposit similar to that observed at lower temperatures. Tests at higher tempera- tures are under way. FLUORIDE CORROSION IN THERMAL CONVECTION LOOGPS G. M. Adamson, Metallurgy Division The emphasis in dynamic corrosion testing has shifted from testing NaF-KF-LiF-UF, to testing various coolants containing zirconium fluoride. As a class, the zirconium fluoride coolants are less corrosive to Inconel. The method of attack appears to be the same for the two types. The zirconium fluoride coolants do have the dis- advantage of a high vapor pressure that results 1n the sublimation of zirconium fluoride i1n the expansion pots of the loops. Some work 1s still being done with the mixture NaF-KF-LiF-UF, because it 1s more readily available; 1t 1s more corrosive, trends are easlier to discern. A portion of the work has been on the addition of possible 1nhibitors to the coolants. From the results obtained with a single loop, zirconium hydride will reduce corrosion, but the effect of NaK 1s questionable. Corrosion from also, since 103 ANP PROJECT QUARTERLY PROGRESS REPORT the fluorides appears to increase if crevices are present in the loop walls. Tables 15 and 1,5, which summarize the results obtained from the thermal convection loops operated during this period, are continuations of Tables 12 and 13 in the previous report,¢?) Even with the additional fluids being tested, none of the data in these tables conflact with the conclusions presented in the last report. Fluoride Mixtures Containing ZrF4. As shown in Table 15, four additional fluoride mixtures containing zirconium fluoride have been tested in Inconel loops. The hot-leg attack 1in every one of these loops i1s less than that normally found with NaF-KF-LiF-UF, fuel. Since loop 234 was cleaned only by degreasing, which previously was the standard procedure, it can be compared with the standard loops. The maximum attack in this loop was only 5 mils rather than the 10 to 15 mils normally found with the earlier fuel composition., The amount of attack was also greatly reduced, as shown in Fig. 48. Some of the other loops in which the zirconium fluoride fuels and coolants were circulated did show a maximum attack greater than 10 mils., As discussed below, the attack seems to be more a function of the cleaning procedure than of the mixture. There does not appear to be much variation in the attack by the various zirconium-base mixtures, since all produced about the same attack under comparable conditions. The attack was about the same as that shown 1in Fig. 54 of the previous report,(3) which 1llustrated attack by a mixture [2)G. M. Adamson and K. W. Reber, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Fnding June 10, 1959, ORNL-1294, p. 112 and 113, (S)Ibid., p. 111, 104 containing ZrF,. Even though the number of loops run with these mixtures 1s limited, the attack mechanism seems to be the same as the process of leaching and diffusion of chromium, which was discussed previously(*) in connection with the NaF-Kb-LiF-UF, fuel, The expansion pots of twoe Inconel loops 1n which NaF-ZrF,-UF, was circulated for 500 hr were cut open and over 200 g of sublimed zirconium fluoride was recovered from each, The material was found both on the insulated sides and exposed top of this chamber. The area of the surface from which this material escaped was only 7.4 in.?, and the surface was at a temperature of about 1300°F. 1In general, large crystals were found on the sides and small ones on the top. This vaporization and subsequent sublimation of the ZrF, fluoride is a major problem in the handling of these mixtures (cf., “Experimental Engi- neering,” sec. 2). Effect of Cleaning Loops. The standard procedure for cleaning thermal convection loops before operation has been vapor degreasing and flushing with a fresh organic solvent. This procedure was felt to be more desirable than pickling because of the variables introduced by pickling, even though it does not remove the thin oxide layers at the welds. Heating 1n a dry-hydrogen atmosphere seemed to be a possible method for reducing these oxides without attacking the metal. This operation was tried on several Inconel loops in which the zirconium- bearing fluorides and the earlier nonzirconium-bearing fluorides were circulated. The corrosion results were all comparable, no matter which fluoride mixture was later circulated. From this 1t 1s apparent that the (4)G. M. Adamson, A. des Brasunas, and L. S, Richardson, op. cit., ORNL-1294, p. 126. S01 Table 135 CORROSION DATA FROM INCONEL THERMAL CONVECTION LOOPS CONTAINING VARIOUS FLUORIDE MIXTURES METALLOGRAPHIC NOTES 1 E ) s p . B . TIME OF REASON HOT LEG 1;\%01}' e PLUORIDE ! CTRCELATION FOR TEMPERATURE CHEMICAL NOTES NO. 1 PROCEDURE MIRTURE (inx) TERMTNATION {°F) ‘Hot Leg Cold Leg 591 | Degreased NaF-KF-ZIF4-UF4 500 Scheduled 15G0 Light to moderate inter- Rough surface; thin noxne Sgme increase in Cr, decrease in {4.8-50.1-41,3-2.8 mole %} granulaer pictting, 3 to 7 metallic layer Ni, smell increase in Fe, de- mils crease in U 222 | Degreased | NaF-KF-LiF-UF, 500 Scheduled 1650 Intergranular attack, up Lo Thin deposit Cr snd Fe bolh decreased; ko {16.9-42.5-44.5-1.1 mole %) 18 wmils; pitting, § mils systematic variations 228 } Hydrogen NaF-KF-Zol «UF, 500 Scheduled 1500 Light intergranular pitting, Rough surface; thin de- Small increase in Cr, decrease in (4,8-50,1-41.3-3.8 mole &) 3 to 13 mils, average 3§ posited layer Ni, U variahle mils 224 | Depreased | NaF-KF-Lif-UF, + NeK ! 500 Scheduled 1500 Intergranuiar pitting, up to | Thin surface lever with Cr increased only siightly, Fe {10.9-43.5-44.5-1.1 mole %)! 10 »ils some adhering nonme- decreased tullic parcicles 230 | Hydrogen NaF-KF-ZrF, 500 Scheduled 1500 Light intergranular pitting, No deposited layer Gr increessed slightly, Fe de- {36.0-18,0-46.0 mole %) average 4 mils with maxi- creasad, others vary mum 7 mils; grains were not outlined 226 ! Hydruogen NaF-KF-21F, -UF, 500 Scheduled 1500 ¥idely scattered inter- Both metallic apd non- Cr increased, Fe and Ni de- (36.6-14.G-45,6-3.8 mole %) grannlar pitting, average wetallic deposits creascd, others vary T omils with maximum of 10 mils 229 | Degreased | NaF-KF-Lif-UF, 500 Scheduled 1500 Moderate to heavy inter- Thin deposit ut least Large increase ia Or, Fe de- {10.9-43.5-44.5-1. 1 mole %) granular pitting, up te 18 partially nonmetallic creased, small decresse inUand F Mot treated mils, average 8§ mils 227 | Degreased HaF-RF-LiF.UF, 500 Scheduled 1500 Moderate to heavy pitting, Yery thin metsllic layer | Cr increased, Fe decreased, U {10.9+43.5-44.5-1. 1 mole %) § to 16 mils; mainly guite uniform intergranular 234 | Degreased | NaF-ZrF -UF, 480 Pover 1500 Light seartered intergranular! Both mevallic and nen- Cr increased, large decresse in {46,0-50,0-2.0 mole %} failure pittiag, up to 5 wils, mevallic crystals on Fe and Ni, U increased average 2 mils; dirty the walls fnconel used 236 : Hydroger NaF-Zrl, -UF, 462 Pewer 1560 Moderute intergranular pitting, | Both metallic and non- Cr increased, Fe ond Ni de- {46.0-50.0-4.0 mole %) failure average B mile with maximum of metellic particles on creased; LU increased 10 wils; scme pils 1n grains wall 225 | Degreased | NaF-KF-LiF-UF, + 1,2% ZrH, 500 Scheduled 15090 Very little attack; surface Metallic layer, 9.2 Cr, Fe, U, and Ni all decreased {19.9-43,5-44,.5-1.1 mole %} rough, few pits, 0.5 mil; mil; layer was guite possibly some atrtached adherent i crystals; precipitate i formed in crystal boundaries 231 | MNaK Nab-ZvF, -UF 500 Scheduled 1500 Moderate general pitting, up { Both metallic and non- Cr increased, Fe, Ni, and U all {46.0-50.0-4.0 mole R) to 9 pile metallic particles decreased deposited on wall 237 | Na Naf -ZrF -UF 560 Scheduled 1500 Tvpical Inconel pitting, Rough surface with both 4 P {50,0-46.0-4.0 mole %) maximum penetrution 5 mils metallic and non- with average 3 mils metullic crystals ad- hering 238 | Nuk NaF-KF-LiF-LUF, 500 Scheduled 1500 (10.9-43.5-44.5-1.1 male %) 223 | NaK Nal-KF-LiF -UF 500 Scheduled 1500% £10.9-43,5-44.5-1.1 mole %) “Hot leg emnealed 1/2 hr at 1750°F, **Lap joint in hot leg. ‘01 HAGWILJES ONIGNd dOI¥dd HOd c461 901 Table 16 CORROSION DATA FROM STAINLESS STEEL THERMAL CONVECTION LOOPS CONTAINING VARYOUS FLUORIDE MIXTURES i 1 TIME OF CIRCULATION {qr) 500 309 igg 62 91 500 43 COF Loop | (AYPE OF FLLUORIDE . STAINLESS LRE 0. STEEL MixTl 123 316 NaF-4F-LiF-UF | (10.9-43.5-44.5-1. 1 mole %); 126 316* | NaF-KF-LiF-UF, i (10.9-43.5-44.5-1. 1 mole %) H, treated : 1 i 124 3156 NaF-XF-ZrF, -UF, {4.8-50,1-41.3-3.8 mole %) 1 120 | 316 Nab-KF-LiF-UF, (10.9-43.5-44.5-1.1 mole %) Not treated 125 316 NabF-KF-LiF-UF, (10.9-43.5-44.5-1.1 moie %) 1238 316 NaF-ZrF, -UF, § (46.0-50.0-4.0 mole %) 127 ELC NaF-KF-LiF-UF, i6 {10.9-43,5-44.5-1. 1 mole %) REASON METALLOGRAPHIC NOTES HOT LEG ] kQH , TEMPE@ATURE Hot Leg i Cold Leg CHEMICAL NOTES TERM INATION (°F) : 1 Scheculed 1650 Intergranular attack, up to | Metallic layer, 1 mil Fe, Cr, and U all decreased; hot I 13 mils; some grains re- thick feg highest in all i moved Plug 1500 ‘ Heavy intergranular attack, MNonuni form metallic i Cr increased, Fe decreased, U up to 14 mils; some grains deposit; few non- § varied, but little change, Ni removad metallic crystals in same top pertion; slighc attack in seme areas l.eak 1500 Intergranular attack, usually | Some evidence of a thin : Cr up slightly, Ni decreased, Fe S mils but occasionally up metallic deposit i varied but is unchanged, U lower to 11 mils Plug 1500 Heavy intergranular attack, Thin surface deposit with| U increased, all other elements up to 12 mils; some grains both metallic and non- varied removed metallic crystals ad- hering Piug 1650 Heavy intergranular attack, Metallic deposit with ve and Cr varied, Ni decreased up to il mils some crystals not at- tached to wall Scheduled 1500 Heavy intergranular attack, Surface rough; heavy Cr increased, Fe and Ni de- up to 8 mils; some grains layer of nonmetallic creased, U increased removed crystals in which some metallic crystals were embedded; some non- ; i metallic crystals were | i encased in metai ; i i Plug 1500 i Heavy intergranular attack, Metallic deposit with i Cr increased, Fe decreased, others up to li mils some metallic crystals not adhering varied *Hydrogen fired; hydrogen atmosphere. LHOdA SSHHI0Ud A THALYVYNO LJAT0Md dNV FOR PERIOD ENDING SEPTEMBER 10, 1952 Fig. 48. cleaning cycle is just as important as the choice of fluoride mixture. Comparison of these loops with similar loops that have only been degreased shows that the amount of attack, as indicated by the number of holes, is reduced. The attack 1s concentrated in a few grain boundaries and actually extends to a greater depth. This is shown by Fig. 54 in the previous report.(3) In cleaning with hydrogen, temperatures in excess of 1850°F are necessary to provide a minimum temper- ature above 1750°F. These temperatures cause grain growth and possibly concen- tration of impurities in the grain boundaries, Since hydrogen firing did cut down the amount of attack, a method of removing the oxide without DT e . % . W " Jwts " .y . e s gr Corrosion of Inconel Thermal Convection Loop by NaF-ZrF4-UF4 (46-5%0-4 mole 9) After 500 hr at 1500°F. Etched with aqua regia. 250X the resulting grain growth would be desirable. It was found that NaK at 1500°F would remove weld scale from Inconel. Therefore several Inconel loops were cleaned by heating the hot leg to 1600 °F and the cold leg to 1500°F and circulating NaK for 4 hours. The NaK was drained while still hot, and the loops were then heated and evacuated to remove any remaining NaK. The one loop examined showed attack as deep as would be obtained by degreasing, or even deeper. It seems likely that when the NaK was drained, sodium or potassium oxides were left behind in cold spots, or possibly some oxides were deposited in the cold leg by the 107 ANP PROJECT QUARTERLY PROGRESS REPORT 100°F temperature difference. This method of cleaning will be checked further and other methods will be tried. Corrosion Inhibiiors. In the static corrosion studies 1t was shown that the addition of zirconium metal to NaF-KF-LiF-UF, reduced the attack. Because of the great affinity of zirconium metal for the various gases and the difficulties i1n obtaining adequate mixing, 1t was not practical to add the zirconium metal to a loop either before or after heating. Zirconium hydride was also tried as a possible inhibitor because 1t does not decompose until slightly below operating temperature, and the nascent hydrogen liberated during decomposition might helpin cleaningup the fluorides. Fig. 49. with 1/2% Added ZrH, After 500 hr at 1500°F. 108 The NaF-KF-LiF-UF, with 1/2% ZrH, added was circulated in an Inconel loop for 500 hr at 1500°F. After circulation, the hot leg was found to have a roughened surface but the attack extended only to a depth of 0.5 mil; the surface 1s shown 1n Fig. 49. The attack was the least found in any loop in which fluorides have been circulated. 1In the hot leg, some precipitated crystals were observed just under the surface and in the grain boundaries. These crystals do not seem to be continuous enough to act as a protective layer. An adherent metallic layer about 0.2 mil thick was found in the cold leg. The lack of corrosion in this loop was further evidenced by the very low amounts of impurities in the fluorides. This test is now being repeated, and IPHOTO T- 1764 . s Voo L ; . Corrosion of Inconel Thermal Convection Loop by NaF-KF-LiF-UF4 Etched with aqua regia. 250X FOR PERIOD ENDING SEPTEMBER 10, Tt the effect addition to a zirconium-base fluoride fuel is also being studied. * It was pointed out in the previous report{3) that a small NaK addition reduced the corrosion obtained when neonuranium-bearing coolants were used. As an extension of this work, NaK was added to two other Inconel loops. l.oop 224 was used to circulate Nal-KF- LiF-UF,, and Loop 241 to circulate NaF-ZrF, -UF,. About 8 cc of NaK was added to each loop. In both loops the NaK was placed in the pot and the fluorides were added to 1t. Loop 224 was started without difficulty and operated satisfactorily for 500 hours. The maximum hot-leg attack measured 10 mils, which is lower than the usual average but not enough to show a definite effect. However, since the metallic impurities found in the fluorides after circulation were also Jower than normal, it seems possible that some actual improvement took place. Loop 241 1s now operating satisfactorily, but there was con- siderable difficulty 1in start-up, which may have been due to a reaction that yielded some solid precipitates. The cold-leg temperatures were lower than normal and kept decreasing. (S)Adamson and Reber, Ibzd., p. 109, 1952 Another Inconel loop, in which chromium metal powder was added to the circu- lating NaF-KF-Li¥F-UF, mixture, 1s operating normally. Crevice Corrosion. Crevice cor- rosion has been observed but requires further 1nvestigation. In an Inconel pump loop operated with NaF-KF-LiF €11.5-42.0-46.5 mole %) very little attack was found in most of the loop, but heavy attack was found 1n two different types of crevices, One type of ecrevice resulted from sleeves joining pipe sections and the other from a thermocouple well. This loop had operated for about 1000 hr at 1250 °F and below and 100 hr at 1500°F. Several other loops were then checked for crevice corrosion by examining the top welds. The corrosion of crevices found i1n the hot legs of loops 219, 229, and 234 1s compared 1in Table 17 with the corrosion of the wall directly opposite the crevice. Figure 50 shows a section through the joint of Inconel loop 219. It should be pointed out that the attack in this portion of the hot leg 1s con- siderably less than that given in Table 15, since the sections were at slightly different temperatures. One additional piece of evidence for the existence of crevice corrosion 1s the fact that the attack at the mouth of Table 17 CREVICE CORROSION IN THERMAL CONVECTION LOOPS METALLOGRAPHIC NOTES LOOP MAXIMUM ATTACK (mils) FLUORIDE MIXTURE : NO. Hot Leg Crevice 219 | NaF-KF-LiF-UF, | 3 6 229 NaF-KF-LiF-UF, | = 4 6 234 | NaF-ZrF-UF, 3 5 Two to three times more attack in the crevice Many times more attack in the crevice Very little attack in hot leg with moderate amount in the crevice; tendency for attack to follow grain boundaries 109 ANP PROJECT QUARTERLY PROGRESS REPORT 5 Ll W o M . . .- 3 L kT RA] A / F t VRS dfi%‘;i;'f : f;\x L r"Pdor T-1802 ° P e RN R T vl } o o - L4 '5’ ¢ q‘-u‘ o ‘ ‘.. 1. i. ’i‘ "i "‘ ‘..‘ 0 \, "‘0J : sV of _5%; flflkfi? . ‘,,h:i ™ -u Y f"(o _ CREVWE Ng"- "i,f . ) * g *fi A .. re .- ‘ K.‘.;‘gf*; ::_ - @ '# ‘ o L% ‘5 ° 4 . b RN -* "'\w ,".f.. * .‘ ,‘i:ayn-, . - » Fig. 50. After 500 hr at 1500°F. the crevice 1s less than that found within. The mechanism of this attack is not yet known. Further tests will be made to determine just how serious this effect will be. Temperature Variations, To de- termine the effect of temperature on the corrosion of Inconel, two loops (218 and 222) were run at 1300 and 1650°F, respectively. It apparent that the attack increased slightly with temperature. The type of attack was the same at each temper- ature, was A type 316 stainless steel (123) was operated NaF-KF-LiF-UF,. 300-series loop at 1650°F with This was the first stainless steel loop to 110 *’fw//‘”w Effect of a Crevice on Corrosion of Inconel by NaF-KF-LiF- UF , Etched with aqua regia. 250X operate 500 hr without plugging. The attack in the hot leg was inter- granular and extended to a depth of 13 mils. The test was repeated (loop 125), but the loop plugged in only 91 hours. The insulation in the second loop was not so good as in the first and the cold leg operated at a slightly lower temperature. Further work 1s necessary to determine whether 1t could be the small temperature difference that caused the plugging. Variations in Alloy Composition of the Wwalls. Tables 18 and 19 present the results of chemical analyses of successive layers drilled from the inside of two loops (78 and 121) for comparison with similar data obtained from two other loops (210 and 219) and FOR PERIOD ENDING SEPTEMBER 10, 1952 Table 18 CHEMICAL ANALYSES OF SUBSURFACE LAYERS OF INCONEL WALLS OF LOOP 78 IN WHICH : NaF-KF-Li¥F WAS CIRCULATED DEFTH OF | COVPOSITION (%) LAYER : : : (mils) Fe | N | Cr | Fe/Ni | KF* | LiF* | NeF* | Other nga;njgé Total 5 7.03 | 78.3 | 12.0 | 0.089 | 0.09 | 0.3 | 0.07 | 1.68%* - 97.4 99.5 10 8.66 | 77.5 | 11.8 | 0.112 | 0.03 | 0.15 o.o? 1. 46 - 98.0 99.6 15 6.93 | 77.6 | 13.6 | 0.090 | 0.02 | 0.1 | 0.02 | 1.18 98.1 99. 4 20 6.81 | 77.7 | 13.7 | 0.088 | 0.02 | 0.1 | 0.02 | 1.30 98.3 99.6 25 6.66 | 77.9 | 14.3 | 0.086 | 0.06 | 0.2 0.07 | 1.78 - 98.9 101.§ Exterior | 7.15 | 77.4 | 15.2 | 0.092 : 1. 48%ss 99,8 161.0 These values calculated from spectrographic analysis of metals by assuming them to be present as fleorides. - : : Total of spectrographic analysis for Al, Co, Cu, Mn, Mo, 51, and Ti. 5 : Average of values above.: Tabie 19 CHEMICAL ANALYSES OF SUBSURFACE LAYERS OF TYPE 316 STAINLESS STEEL WALLS OF LOOP 121 IN WHICH NaF-KF-LiF-UF, WAS CIRCULATED DEPTH OF | : COMPOSITION (%) LAYER . - e o Total Fe, ‘ (mils) Fe Ni .Cr Mo KF* Lif Nak* | Other Ni, Cr, Mo Total 2 60.4 | 16.9 | 14.5 | 2.8 | 1.5 | 3.7 | 1.1 | Avg. 1.6 94.6 102.5 = : Co, Cu, : - 5 60.8 | 16.1 | 147 | 2.7 | 0.9 | 3.7 1.1 Mn, Si 95.2 101.6 10 63.5 | 14.8 | 15.1 | 2.5 | 0.5 | 2.2 | 0.5 95,9 160.7 15 64.1 | 14.6 | 15.7 | 2.5 | 0.3 | 1.1 | 0.5 96.9 100. 4 20 0 65.2 | 14.3 | 16.2 | 2.5 | 0.2 | 0.7 | 0.5 98.2 101.1 25 65.1 | 14.0 | 16.8 | 2.4 | 0.3 | 0.7 | 0.5 - 99.3 101. 4 Exterior | 66.0 | 13.3 | 16.8 | 2.4 | | 98.5 100.1 * : : These values calculated from spectrographic analysis of metals by assuming them to be present as fluorides. ‘ : 111 ANP PROJECT QUARTERLY PROGRESS REPORT presented in the previous report, (%) [.oop 78 was constructed of Inconel and circulated NaF-KF-LiF for 500 hours. In the first loop{®’ the variations 1n chemical composition appeared to be the same as those found in loop 78, but in loop 78 they were much smaller. Chromium was depleted to a depth of at least 25 mils, Since the fluoride content was low and constant, 1t was evident that there was practically no penetration. In the type 316 stainless steel loop, chromium was reduced slightly, and the iron was the major material leached out. The nickel and molybdenum contents were essentially unchanged. In this loop the fluoride content was much higher than in the other loops and decreased gradually with depth. Evidently the fluorides are trapped in crevices. A loop in which zirconium- bearing fluoride mixtures will be circulated 1s being prepared. Postrun Examination of Fuels (D, C. Hoffman, F. F. Blankenship, Materials Chemistry Division). Examination of sections from a number of thermal convection loops operated by the experimental engineering group 1n which zirconium-bearing fluoride fuels were circulated in loops of type 316 stainless steel and Inconel has been accomplished by the method and techniques previously described.(7? The results of these investigations are briefly summarizedin the following statements. The zirconium-bearing fluoride fuels show considerably less corrosion transfer than the fuels tested previously. No NaK,-CrF_ has been discovered in the loops run with these fuels. Inconel seems generally superior to stainless steel. and mass (6)rpia., p. 115, Table 1f. (7)[). C. Hoffman and F. F. op, cit,, ORNL-1294, p. 121. Bl ankenship, 112 In loops of type 316 stainless steel, a small amount of magnetic material showing an x-ray pattern resembling an iron-chromium alloy appears 1in and above the trap at the bottom of the cold leg. Cne Inconel loop that had been hydrogen fired showed somewhat more mass-transferred metal in the cold leg than a similar section of an Inconel loop that had simply been degreased. The metal gave an x-ray diffraction pattern for iron i1in which the lines were shifted slightly. In one type 316 stainless steel loop in which a mixture containing NabF-ZrbF,-UF, (46-50-4 mole %) was circulated for 500 hr, a brown un- known material (average index of refraction = 1.556) was found in the vertical hot section. A con- siderable amount of metallic deposit was found in the trap of this loop. Identification of the deposit is 1ncomplete. COMPATIBILITY OF BERYLLI1IUM OXIDE WITH VARIOUS FLUORIDE MIXTURES R. E. Meadows Materials Chemistry Division H. J. Buttram In the present ARE design a liquid heat transfer agent will be pumped slowly through the moderator space to maintain a uniform temperature in the beryllium oxide. The flow rate of this liquid past the beryllium oxide blocks should be at most a few feet per minute. The number and irregular shapes of the beryllium oxide speci- mens would seem to preclude canning or electroplating of the pieces., One of the essential characteristics of the moderator coolant therefore is chemical compatibility with the beryllium oxide. FOR PERIOD ENDING SEPTEMBER 10, 1952 Small specimens sawed from hot- pressed beryllium oxide blocks typical of those proposed for use in the ARE have been used in these studies. These pieces have been exposed for 100 hr in sealed capsules of type 316 stainless steel to molten alkali fluorides with and without ZrF, or Bng.(s) ‘In all static tests the temperature was maintained at B800°C; in all the:dynamic capsule tests the capsule was tilted every 15 sec, with the cold end at 650°C and the hot end at 800°C. ’ : Preliminary studies of the bebavior of beryllium oxide in a static bath of molten NaF~KF'ZrF4 (5-52-43 mole %) indicated that the rate of reaction was appreciable but that i1t decreased with time. JIdentification by x-ray diffraction and petrographic microscopy of a film of ZrO, on the beryllium (B}Addibional beryilium oxide compatibility vests, with Na, NaK, and Pb, are reported in this section under * lLiquid Metal Corrosion.” oxide blocks suggested thata protective film of ZrO, might make it possible to use a coolant containing zirconium. Exposure of the blocks for longer times in static tests and for 100-hr intervals in dynamic tests, however, clearly demonstrated that zirconium- bearing mixtures are not suitable in contact with beryllium oxade. llata shown in Table 20 are representative of those obtained. The attack seems to be due to a uniform dissolution of beryllium oxide, since the corners and sharp edges of the specimens retained. are However, as the data in Table 20 indicate, systems of alkali fluorides containing Bel, are compatible with beryllium oxide. Microscopic and x-ray examination of specimens crushed afrer dynamic testing show BelF, in the body of the beryllium eoxide block. The penetratien, which appears to do no damage, is undoubtedly responsible for the weight gain observed. Further studiesof the Be0-BeF,-alkali fluoride systems are being made. Table 20 REACTI0N OF BERYLLIUM OXIDE WITH FUSED SALTS CONTAINING ZfF4 O/ Bek, UNDER DYNAMIC CONDITXIONS . WEIGHT CHANGE COOLANT B . 5 REMARKS : In % In mg/dm®/day NaF-KF-ZrF, ~66 ~5879 Specimen in hot end of tube (5-52-43 mole %) | | | . -16 ~-1432 Specimen in cold end of tube ~60 -4907 Specimen dissolved uniformly® NaF-KF-BeF, +4.1 +362 Specimen appeared unchanged (30-5-65 mole %) 4 fi : : : 6,9 +594 Salt apparently penstrates blocks Original dimensions, 0.508 by 0.253 in.; final dimensions, 0.410 by 0.176 in. ANP PROJECT QUARTERLY PROGRESS HYDROXIDE CORROSION F. Kertesz Materials Chemistry Division A. des Brasunas E. E. Hoffman G. P. Smith Metallurgy Division The effectiveness of a hydrogen atmosphere in inhibiting hydroxide corrosion was demonstrated previ- ously(?) in at least two independent tests. Recent tests 1n a thermal gradient apparatus have substantiated the results. Furthermore, addition of the reducing agent, sodium hydride, has reduced mass transfer 1n an Inconel system tested with potassium hydroxide. A series of temperature- dependence tests indicated that the corrosionof Inconel by sodium hydroxide is negligible up to 450°C; more ex- tensive tests at 600°C gave evidence of severe attack. Corrosion tests of a stressed specimen 1indicated in- creased corrosion, Temperature of Test. A series of temperature-dependence testsof Inconel in sodium hydroxide has been completed. From the results listed in Table 21, it would appear that in 100-hr tests, corrosion of Inconel 1s negligible unti]l the test temperature 1s raised to above 450°C. (Q)Brasunas and Richardson, op. cit., p. 116. REPORT Nickel was tested similarly 1in sodium hydroxide at a series of hot- zone temperatures ranging from about 400 to 800°C. This was done in the hope of finding a threshold tempera- ture below which no mass transfer occurs. However, the results indicated that as the maximum temperature was diminished, the amount of nickel crystals in the cold zone was di- minished; mass transfer was not eliminated. More extensive corrosion data are available at 600°C, at which tempera- ture a variety of materials was subjected to 100-hr exposures to sodium hydroxide. These data are summari zed in Table 22. Crystal-bar zirconium was heavily attacked even at this temperature, although it 1s possible that the canning in type 347 stainless steel may have adversely affected its behavior. Of the alloys tested, Monel showed the least cor- rosion. It 1s obvious that the use of these alloys with sodium hydroxide under an inert atmosphere 1s not possible. Residual Stresses in Tests have been run to check the effect of residual stresses on the corrosion resistance of Inconel 1n sodium hydroxide at 816°C. The stresses were induced by cold rolling. The tests were first run for 50 hr, but Specimen, Table 21 TEMPERATURE-DEPENDENCE TESTS OF INCONEL IN SODIUM HYDPROXIDE FOR 100 HOURS TEMPERATURE OF TEST WEIGHT CHANGE o .2 METALLOGRAPHIC NOTES (°C) (g/in.%) 350 None No attack 450 None No attack 550 +03. 0050 3 to 4 mils of intergranular penetration 593 +0.0060 3 to 4 mils of intergranular penetration 114 FOR PERIOD ENDING SEPTEMBER 10, 1952 Table 22 STATIU CORROSION TESTING MATERIALS Ifl SODIUM HYDROXYIDE AT 606°C FOR 100 H@URS - WEIGHT CHANGE. - Crystal-bar zirconium _ ~1199 Specimen retained its original shape and {14.5% of specimen) appeared to be uniformly dissolved ~8787 Specimen retained its original shape and (88.4% of specimen) appeared to be uniformly dissolved Inconel +459 Surface oxidized to depth of 6 to 7 mils with a very thin layer on outside of attack; attack proceeded intergranularly Type 316 stainless steel -79 Surface oxidized to T mils deep; attack proceeded evenly Type 321 stainless steel -182 Surface oxidized te 6 mils deep; attack proceeded intergranularly +13 Surface oxidized to 6 mils deep; sttack proceeded intergranularly Menel +8 Light intergranular attack 1 to 2.5 mils deep +10 Light intergranular attack 1 to 2.5 mils deep ' Hastelloy B ~350 General pitting, 4 mils deep; specimen very brittle the specimens were all corroded throughout their entire thickness. The time was then cut to 8 hours. The results of the 8-hr tests, listed in Table 23, seem to indicate that residual stresses of this nature do have a deleterious effect on the corrosion resistance of Tnconel 1in sodium hydroxide. Similar tests were run on A nickel for 50 hr, but there was no evidence of attack even on the most severely cold worked (71% re- duction in area) specilmens tested. Corrosion Inhibitors. Inconel was tested in potassium hydroxide with a 2% NaH addition. Very little evidence of mass transfer was evident after 100 hr (51,200 cycles) with a hot-zone temperature of 750°C and a cold-zone temperature of 550°C. The tempera- tures were measured by spot-welded external thermocouples. Hydrogen Aimosphere. Earlier(1!?) (10)e p. Swith, J. V. Cotheart, snd W, H. Bridges, op. cit,, OBNL-1227, p. 127. 115 ANP PROJECT QUARTERLY PROGRESS REPORT Table 23 EFFECT OF RESIDUAL STRESSES ON CORROSION OF INCONEL BY SODIUM HYDROXIDE FOR 8 hr AT 816°C REDUCTION IN ORIGINAL THICKNESS WEIGHT CHANGE , . THICKNESS (%) (mils) (&/in.?) METALLOGRAPHIC NOTES 71 10 -0.0011 Attacked throughout 50 17.5 +0.0088 9 mils of penetration, 2 to 3 mils of unattacked material 26 26 10.0188 7 to 8 mils of penetration, 12 nils of unattacked material 11 31 10.0171 4 to § wils of penetration, 23 mils of unattacked material 5 33 10,0203 3 to 5 mils of penetration, 25 mils of unattacked material 0 35 +.,0210 3 to 5 mils of penetration, 27 mils of unattacked material studies of mass transfer with small melt. 1In 40 hr a copilious quantity of thermal convection loops 1indicated that a hydrogen atmosphere had an appreciable effect i1in reducing nickel mass transfer. This effect was more closely investigated with the thermal convection apparatus described in the last report.{?!) A number of ex- periments has been performed with this apparatus. The results are 1llustrated by two typical tests. In both tests the temperature of the nickel cup was about 725°C, whereas the temperature of the inner nickel tube was 525°C, a temperature difference of 200 degrees. The first test was conducted with a static vacuum over the hydroxide (ll)J. V. Cathgart, W, H., Bridges, and G. P, Smith, op., cit,, ORNL-1294, p. 118, 116 dendritic nickel crystalshad encrusted the itnner nickel tube and extended more than half way across the hydroxide melt. This 1s a case of normally expected nickel mass transferin sodium hydroxide {for a large temperature difference. In the second test a purified hydrogen atmosphere replaced the static vacuum. In 170 hr, as compared with 40 hr for the first test, the only evidence of mass transfer obtained was a tiny band of nickel crystals on the inner nickel tube at the gas-liquid interface. This band was about 1 mm wide and 0.1 mm thick. Confirmatory evidence for the inhibiting effect of hydrogen was also obtained with the tilting-furnace apparatus. The test with nickel and FOR sodium hydroxide at a hot-zone temper- ature of about 750°C and a cold-zone temperature of 600°C gave very little evidence of mass transfer after a 100~hr test when a hydrogen atmosphere of about 1 atm was maintained. This represents a very appreciable re- duction 1n mass transfer when compared with tests made under similar con- ditions in vacuum. LIQUID METAL CORROSION A, des Brasunas G. P. Smith D. C., Vreeland Metallurgy Division F. Kertesz Materials Chemistry Division The desirability of lead as a coolant is largely offset by 1ts susceptibility to mass transfer and 1ts attack on most commercially available heat-resistant alloys. The use of a lead-sodium mixture would be advantageous i1f 1t possessed predominately the corrosion character- 1stics of the sodium and the radio- active characteristics of the lead. Tests indicate that the corrosiveness of lead 15 diminished with the ad- dition of increasing amounts of sodium. In other corrosion tests with lead, the mass transfer phenomenon and the effect of temperature have been examined further. Sodium cor- rosion was measured in the “spinner” test 1n a new, revolving, corrosion apparatus. Testsof type 1020 chromized steel 1in various low-melting-point alloys show that this steel cannot be recommended as a container for the alloys tested. In addition, the compatibility of beryllium oxide with such probable moderator coolants as lead, sodium, and NaK i1s being in- vestigated. PERIOD ENDING SEPTEMBER 10, 1952 Mass Transfer in Liguid Lead. Studies are being undertaken of the mass transfer of molybdenum and Inconel in liquid lead. Preliminary to the corrosion studies, research has been conducted on the handling and purifi- cation of liguid lead, and a few mass transfer measurements have been made. Since molvbdenum 1s difficult to weld and oxidizes excessively at high temperatures, the molybdenum corrosion tests were conducted on specimens of the metal in convection loops made of quartz or pyrex. These materials are resistant to corrosion by lead(!?) and do not present the problem of a bimetallic system, Studies were made of the de- oxidation of high-purity lead. From the quantity of oxide that collected on the surface of lead melted under purified hydrogen, it was concluded that the original material had con- tained a considerable amount of occluded oxide. A number of ex- periments was performed in which hydrogen was passed in fine bubbles formed by a fritted pyrex disk through molten lead at 450 to 500°C for 20 hr, and the lead was then filtered through a fritted pyrex disk. There were no visible oxide films on the surface of the filtered lead, but a small amount of oxide was found on the filter. This indicated that the hydrogen reduction had not been complete. Further studies of the conditions required for complete reduction by hydrogen will be made. No results of the corrosion of either molybdenum or Inconel in the convection loops have yet been ob- tained because of difficulties as- sociated with the operation of such 12 ' ' ( )R. N. Lyon {(ed.), Liquid Yetals Handbook, NAVEXOS P~T733, p. 105 {(June 1, 1950}, 117 ANP PROJECT QUARTERLY PROGRESS REPORT loops. However, a preliminary ex- periment was performed to determine whether there 1s mass transfer of molybdenum during temperature cycling in the manner reported{!'3) for copper in bismuth, A mechanically polished molybdenum specimen, contained in lead, was cycled every 12 min for 42 hr between 500 and 510°C. At the end of the test period the lead was filtered at 500°C. No mass transfer crystals were found. This result should be taken as preliminary because the molybdenum was undoubtedly coated with an oxide film that may have prevented contact with the molten lead. (IS)G. P. Smith, and J. V. Cathcart, Metallurgy Division Quarterly Progress Report for Period Ending Jenuary 31, 1952, OBNL-1267, p. BRS. Effect of Tempevature on Lead Corrosion. Temperature-dependence tests of several materials i1n molten lead have been run. The only attack that was noted 1n these tests was at 700°C and above. The results of the tests, which were run for 100-hr periods, are given in Table 24. Lead-Sodium Mixtures, Seesaw corrosion testsof lead-sodium mixtures were very encouraging 1n that both corrositon and mass transfer were minimized. The tests in type 310 stainless steel were made with ad- ditions of 0, 3, and 50 wt % sodium to the lead. The test capsules had a hot-leg temperature of 780°C and a cold-leg temperature of 475°C, The Table 24 TEMPERATURE-DEPENDENCE TESTS OF VARIOUS MATERIALS IN MOLTEN LEAD TEMP(%P::A;TURE MATERI AL METALLOGRAPHIC NOTES 400 Inconel No attack Type 446 stainless steel No attack Type 309 stainless steel No attack 500 Inconel No attack Type 446 stainless steel No attack Type 309 stainless steel No attack 600 Inconel No attack Type 446 stainless steel No attack Type 309 stainless steel No attack 700 Inconel Specimen attacked to about 1/2 mil é Type 446 stainless steel No attack | Type 309 stainless steel No attack \ 816 Inconel Specimen attacked 5 to 9 mils Type 446 stainless steel Slight roughening of surface 118 FOR tests were for 190-hr periods, during which time the capsules were cycled 51,000 times. ' Examination of the specimens after test indicated: {1) the usual mass transfer 1n the cold zone of the sodium-free test, (2) reduced mass transfer when 3% sodium was added to the lead, (3) no mass transfer when 50% sodium-50% lead was used. The attack in the hot 2zone was reduced from 6 mils in the case of the sodium- free lead ©to 1 mil in the case of the 50% sodium~50% lead mixture. Spinner Tests with Sodium. The spinner test, designed to supply what might be called *“high-velocity” corrosion data, has been used for two tests with molten sodium. consists of a large pot of type 347 stainless steel for holding the molten bath and a rotating shaft with speci- men holders that extend into the pot. The spinner i1s equipped with a gas- tight cover so that an inert gas atmosphere can be contained during test. The entire assembly is lowered into a pot furnace to attain the required test temperature. | The spinner The first spinner test was conducted with Incenel specimens in sodium. During the test the specimens were immersed in sodium first at 450°C, but not in motion, for 189 hr; next at 450 to B816°C, in motion, for 11 hr; and finally at 816°C, in motion, for 72 hours. The total time of the test 272 hours. The spiuning speed 400 ft/min; a helium atmosphere was used; and 8.9 1lb of sodium was 1n the test container. The specimens were three 1 3/4-in. lengths of 1/2- in.-0D Inconel tubing with the leading end flared to 3/4 in. in diameter. Afver the test, a very light, crystal- line deposit approximately 1/2 mil in depth was noted on the three specimens and all had weight gains of approxi- mately 1%. Was was Under a microscope 1t was PERIOD ENDING SEPTEMBER 10, 1952 apparent that beneath this deposit there was a layer of voids approxi- mately 1 mil deep in the Incomel. Some of the crystalline deposit was scraped from the tubing and sent for chemical analysis. The analysis reported was 64% nickel, 27.5% chromium, 8.3% 1ron, 0.15% aluminum, 0.23% copper, and 0.12% titanium. There was an 1ncrease in chromium and i1rton content and a decrease 1in nickel content as compared with the normal Inconel composition (80% nickel, 15% chromium, 5% i1rom). Since Lhe spinner pot is constructed of type 347 stain- less steel, 1t i3 not known whether the surface layer and voids a velocity effect or were due to the presence of the two dissimilar metals, the Inconel of the specimen and the type 347 stainless steel of the con- tainer. [t is planned to run using type 347 stainless steel speci- mens to obtain a moncometallic system. wWeT e a test Type 1028 Chromized Steel in Low~ Melting-Point Alloys. Tests of type 1020 chromized steel have been run in various low-melting-point alloys. In some cases 1t was difficult to estimate the exact extent of attack because of the adherence of bath metal to the specimen. 1n all these tests, however, attack appeared to be suf- ficiently severe to preclude the use of type 1020 chrowized steel as a container material for the low-melting- point alloys tested. However, the chromium layer on these materials appeared to be quite porous 1in nature. The tests were for 100 hr at 816°C, and except 1in one case both the speci- mens and chromized steel. Results of the tests are presented in Table 25. tubes were Compatibility of Beryllium Oxide with Sodium, Nak, and Lead. Metallic lead would be a satisfactory moderator coolant from the standpoint of com- patibility with the fuel and with the beryllium oxide. As the data in 119 ANP PROJECT QUARTERLY PROGRESS REPORT Table 25 RESULTS OF TESTS OF TYPE 1020 CHROMIZED STEEL IN VARIOUS LOW-MELTING-POINT ALLOYS FOR 100 hr at 816°C BATH METALLOGRAPHIC NOTES 67.8% Sn-32, 2% Cd 67.8% Sn-32.2% Cd 43% Sn~57% Ba 60% B1-40% Cd Test of chromized specimen in Inconel tube; specimen apparently unattacked; however, it was only partially covered by bath metal Specimen completely penetrated in places Erratic attack to 10 mils Erratic attack to 8 mils Table 26 show, insignificant weight gains without appreciable change 1n appearance and dimensions result when beryllium oxi1de 1s tested with lead in static systems. Dynamic tests have as yet not been run with lead, but 1t appears likely that the conclusions stated above will be substantiated. As the data in Table 26 indicate, the beryllium oxide specimens often crack during treatment with NaK. The liberation of considerable hydrogen for intervals up to 1 hr when the specimens are placed 1in water after test suggests that the NaK 1s present in considerable amounts 1in fine cracks Table 28 CORROSION OF BERYLLIUM OXIDE BY NaK AND LEAD WEIGHT CHANGE TYPE OF EXPOSURE METAL USED METALLOGRAPHIC NOTES In % In mg/dm?/day Static Pb +3 Slight coat of green material Pb +8 No attack visible Pb +7 No attack visible NaK +1. 45 +128 No attack visible NaK +0.98 +83 Specimen cracked Dynamic NaK 1.0 +79 Specimen cracked NaK 1.3 +99 Specimen cracked 120 FOR PERIOD ENDING SEPTEMBER 10, 1952 in the structure. Tt 1s possible that prior to heating the NaK enters fine cracks that are caused 1n the sawing operation and that subsequent expansion of the trapped NaK enlarges the fissures until theyare easily visible, Attempts to crack the blocks by thermal shock, which might occur because of the superior heat transfer properties of NaK, have not been successful when the capsules contained only beryllium oxide and helium, The spinner test was also run with molten sodium and specimens of hot- pressed beryllium oxide such as will be used in the ARE. After the test i1t was noted that some spalling had particularly on specimen the data given 1in the weight losses and taken place, No. 1. In Table 27, test Tablie 27 DATA FROM SPINNER TEST OF BERYLLIUM OXIDE IN SODIUM SPECIMEN 1 SPECIMEN 2 Weight Data Before test {g) 21. 2583 23.7735 After test (g) 20,0749 23.7465 Loss (g} 1. 1834 0.027¢ Loss {%) 5.6 0.11 Mmensional Data Length {(in.) Before test 1,443 1. 545* After test 1.427 1. 545¢% Width (in.) Before test 0.545 0. 548 After test 0.528 0. 546 Thickness {(in.) Before test 0. 568 0.556 After test D.556 G.553 R U *Surface very irregular, acturate measurements difficult to obtain, dimensional changes are due to this spalling. It was also noted that sodium had apparently penetrated the specimens, so they were placed 1in alcohal and then water for 24 hr, that is, until the absorbed sodium ceased reacting. In the test the spinning speed was 400 ft/min, and the specimens were submerged in sodium, first at 450°C, but not spinning, for 31 hr; next at 450 to 816°C, spinning, for8 hr; and finallyat 816°C, spinning, for 100 hours. The total time of the test was 139 hours. Test data are summarized in Table 27. FUNDAMENTAL CORROSION RESEARCH In addition to the program of empirical corrosion testing, siderable effort has been devoted to a number of studies designed to assist in the discovery of the corrosion mechanism. These studies include the synthesis and identification of cor- rosion products and studies of re- actions in fluorides and hydroxides under applied potentials. Of particular significance is the mass of experi- mental data supporting the postu- lated(!*) fluoride corrosion of Incomnel, that is, chromiuw depletion of the metal and precipitation of the resulting voids. con - Interaction of Fluorides with Structural Metals (H. Power, J. D. Redman, L. G. Overholser, Materials Chemistry Division). FExamination of the reaction between iron, chromium, and type 316 stainless steel with a fluoride melt containing [iF, NaF, and KF under an inert atmosphere has been extended to cover a range ol tempera- tures and reaction times. lo this study the finely divided metal has been contacted with the melt under an atmosphere of helium at 600 to 800°C (14)Adamson, Brasunas, amd Richardson, op., cit,, ORNL-1294, p¢. 126. 121 ANP PROJECT QUARTERLY PROGRESS REPORT for various periods followed by filtration at high temperature. The concentration of structural elements in the filtrate 1s then determined by chemical analysis. Previous examination(!%) indicated that little (20 to 100 ppwm) chromium and nickel and large amounts (1500 to 3000 ppm) of iron were solubilized by a 6-hr treatment of type 316 stainless steel with the NalF-KF-LiF eutectic at 800°C. Parallel tests have shown that exposure for 72 hr raised the chromium and nickel values slightly (40 to 400 ppm) without perceptible 1ncrease in iron concentration. The amount of reaction required to produce soluble products appears similar when structures of graphite or nickel as the reaction vessel. There 1s some in- dication that the values for nickel and chromium are slightly higher when filtration is performed at 600 rather than 800°C. It was reported previously that when the NaF-KF-LiF eutectic contained added UF,, the iron content of the filtrate roseto 6000 to 7000 ppm and the amount of soluble chromium serve increased to 500 to 2000 ppm. When metallie 1ron 1s exposed to the eutectic mixture at 800°C, 1500 to 2500 ppm of 1ron appears 1n the filtrate. The dissolved 1ron seems to be trivalent in all cases. When chromium powder is used, 150 to 450 ppm of the element 1s solubilized. Examination of the cooled filtrate indicates the presence of NaK,CrF. However, when materials that exposed 1n circulating-loop corrosion tests arerecovered, filtered, and analyzed, a different pattern 1s revealed. have been The NaF-KF-LiF eutectic containing 2 mole % UF4 when filtered after exposure 1in equipment of Inconel or stainless steel shows 100 to 300 ppm (l‘q}H. Powers, J. ND. Redman, and L. G. Overholser, op. c¢it., ORNL-1294, p. 119. 122 Fe, 20 to 100 ppm Ni, and 100 to 2000 ppm Cr. The decreasein solubility of iron without an appreciable in- crease 1in solubility of the other elements may, perhaps, be due to the deposition of iron by the mass transfer mechanism. Experiments of this type will be continued. Synthesis of Complex Fluovrides (B. J. Sturm, L. G. Overholser, Materials Chemistry Division). There has been no indication of simple fluorides of the structural elements in the fluoride mixtures following interaction of the melts with the containers, However, the presence of NaK,CrF, has been confirmed, and it is not unlikely that more sensitive means of detection will show the presence of other complex fluorides. The preparation of complex fluorides has continued as an aid in identifying the various materials formed during static and dynamic corrosion tests, The complex fluorides synthesized in these studies have been identified or characterized by x-ray diffraction patterns and/or by optical crystal- lographic data. It has been demonstrated that NaK,CrF, is the only stable complex fluoride of sodium, potassium, and trivalent chrowium, Attempts to prepare complexes of other proportions resulted 1n mixtures contalning NaKzCrF6 and either NaF or KF. Fusion of various fluorides 1in ratios corresponding to LiK,CrF_, Li,KCrF,, Na,LiCrF,, and NaLi,CrF, yielded products with characteristic x-ray diffraction pattern, indicating that each 1s probably a definite compound. In addition, materials that should correspond to the com- pounds RbK,CrF_. and CsK,CrF, have x-ray patterns similar to that of NaK,CrF,, but they show slight shifts of the lines that indicate somewhat different lattice constants. FOR PERIOD ENDING SEPTEMBER 10, 1952 K,FeF, has been precipitated by slow addition of KF in aqueous so- lution in less than stoichiometric quantity to an aqueous solution of FeCl, acidified with HF. K,;FeF. may be prepared by fusion of K,FeF, with KHF, or by fusion of KF with NH FeF, . The optical properties of these preparations agree with those of K;FeF,, which is prepared by addition of aqueous FeF, to aqueous solutions of KF, as described 1n the litera- ture.{18) Fusion of K,FeF, with NalF, yields NaK,FeF,, which appears to be isomorphous with NaK,CrF. Hydrated KNiF, is formed when proper quantities of aqueous KF so- lutien are added to NiF2 in agueous hydroflueric acid solution. Fusion of a mixture of KHF, with hydrated NiF, and subsequent extraction with water vields a yellow powder that corre- sponds, by chemical analysis, to K,NiF,. The analogous complex com- pounds of LiF and NaF are prepared in a similar manner. The x-ray data from this series of materials are in- complete, Manganic fluoride forms a series of purple to reddish-violet products upon fusion with the alkali fluorides. Compositions corresponding to K MnkF, and K;MnF, yield x-ray patterans typical of pure compounds after fusion. Other materlials prepared in similar fashion have not yet been i1dentified. Since only NaK,CrF, has been positively i1dentified in the products of corrosion and since many other compounds are readily formed from the corresponding fluorides, it might be surmised that complex compounds of the other metals are unstable in the presence of chromium. Fusion of NaK,FeF, (m.p., 970°C) with metallic (1) Minder, z. Krist. A96, 15-19 (1937). chromium for 18 hr in aw inert atmos- phere demonstrated that the reaction Cr + NaK,FeF, ——s Fe + NaK,CrF, takes place almost quantitatively. The melting points of the pure com- pounds have been shown to be: K,CrF,, 1055°C; NaK,CrF,, 1000°C; and NaK,FeF,, 970 °C. EMF Measurements in Fused Fluorides (L. E. Topol, L. G. Overholser, Materials Chemistry Division). De- composition potential measurements inmolten fluorides have been continued, Whereas previous work dealt with the electrolysisof KF atnickel electrodes, subsequent work deals with the effects of additions of small amounts of NiF, (approximately 0.3 wt %) to the KF. Containers of graphite, magnesia, and alumina were used. Graphite 1is advantageous in that 1t 1s fairly stable 1n the presence of fused fluorides but 1s disadvantageocus in that 1t is an electrical conductor. The plot of £ vs., I was a straight line through the origin and yielded a resistance of 0.7 ohm when magnesia was used as the container. In this test neither electrode changed in appearance, but the amount of nickel formed in the cell was greater than conld be accounted for by simple electrochemical processes. The magnesia vessel absorbed the molten fluoride and cracked in numerous places. Norton Alundum crucibles (RA 7232) are the most satisfactory of the containers tried despite the slow dissolution of the Al,0, by the fused fluorides. The degree to which this effects electrochemical results is unknown. Breaks in the voltage- current curves were noted at 0.4 and 0.8 to 0.9 volts. Both electrodes had undergone attack in this test, and the 123 ANP PROJECT QUARTERLY PROGRESS REPORT presence of nickel throughout the melt and of NiF, near the anode was noted. The significance of the breaks 1in the voltage-current curves 1is difficult to ascertain because the dissolution of Al,0;, complicates the picture . and because there 1s evidence that the NiF, thermally decomposed at the temperatures used (850 to 900°C). The thermal decomposition of NiF, appears impossible based on thermo- dynamic considerations (AF = 120 kcal and the equilibrium partial pressure of F, = 10°23 atm(t7)), However, the nickel formation 1in the electrolysis experiments suggested that decomposition was occurring and led to experiments that verified it. excessive A series of experiments was run 1in which various stock samples of NiF, were heated for 4 to 5 hr under helium, first with KF and then with NikF,, alone, 1n cups of nickel, graphite, alumina, and porcelain. In all cases with temperatures from 675 to 900°C, the compound, which 1s somewhat volatile, decomposed and yielded various amounts of nickel. Neither the introduction of a CuO train, heated at 600°C, into the helium line nor the use of an activated charcoal- liquid nitrogen trap had any effect. This indicates that reducing substances in the helium were not responsible for the reduction to nickel, which leaves possible impurities 1n the NiF, as the only cause for the nickel formation other than thermal de- composition. Experiments with K,NiF, gave similar results. In studies of other materials it was found that FeF, heated at 860°C showed some magnetic properties, whereas CrF, at the same temperature appearecd to have been converted to CLTXp L. Quill (ed.), The Chemistry and Metallurgy of Miscellanecous Materials: Thermo- dynamics, McGraw-Hill, New York, 1950. 124 CrF,. Nickel oxide that was heated above 800°C showed some decomposition, but not so wuch as NiF,. Samples of NiO studied in the mass spectrometer have shown dissociation into nickel and the diatomic gas molecule at temperatures of 1000 to 1050°C, Tentative results with NiF, also indicate that dissociation occurs at lower temperatures. Polarographic Studies in Sodium Hydroxide (R. A, Bolomey, Materials Chemistry Division). The collection of polarographic data on the sodium hydroxide—~platinum system has con- tinued. The data have been analyzed in terms of the location of the maximum and half-wave values as affected by temperature. Many of the polarographic waves were found to be too poorly defined to obtain reliable half-wave-potential values. In general, however, the half-wave potentials found to closely parallel the maximum values. Data obtained in the past year have been analyzed and a status report has been written, were Electrochemistry of Sodium Hydroxide (A. R, Nichols, Materials Chemistry Division). Measurements of thermo- galvanic potentials were made to study the mechanisms of corrosion and metal transport in fused hydroxide systems. These measurements were made by de- termining the potential between two nickel electrodes maintained at different known temperatures 1n a melt of sodium hydroxide 1n the apparatus described previously.(!%) Since the voltage developed by the thermo- galvanic cell depends upon the temper- ature difference between the electrodes and also upon the part of the tempera- ture in which the cell 1s operated, the results are conveniently expressed by plotting the guantity scale (18) p. 124. A, R. Nichols, Jr., op. eit,, ORNL-1294, FOR PERIOD ENDING SEPTEMBER 10, 1952 Q¢ = E/AT, in millivolts per degree, against the mean of the electrode temperatures. The hotter electrode was negative in these trials. Although curves showing some features in common were obtained under an atmosphere of purified helium, erratic changes 1in potential were observed, whether hydrogen-fired or untreated nickel electrodes were used. However, when a hydrogen atmosphere was used, the results were fairly reproducible. Figure 51 shows the result of a trial in which the charge consisted of sodium hydroxide and approximately 0.2 mole % nickel oxide. It 1s noted that a maximum value of Q of between 0.8 and 0.9 1s maintained over a range from approximately 400 to 600°C, with a gradual decrease to about 0.2 at a temperature above 750°C. These values do not change appreciably with time; it 1s possible to retrace the curve through a cooling and a second heating and cooling, with good agreement. Trials with a helium atmosphere, in addition to giving less smooth and reproducible curves, showed gross changes with time and gave generally higher @ values in the higher temper- ature range. ' !afi,ls_?iq o @ {mvideg) 360 400 500 eno TR T ano TEMPERATURE (°C) Fig. 51. Potentials in Molten Sodium Hydroxide as a Function of Temperature Gradient. In addition to the trials in which nickel oxide was added, several trials were made with sodium hydroxide alone. With hydrogen-fired electrodes and a helium atmosphere, it was found that the valueof Q was erratic and actually reversed 1ts sign occasionally, which indicated that the cold electrode became negative. When this experiment was carried out in hydrogen, the value of Q became negative after an initial positive interval, during which, presumably, traces of oxide were removed from the electrodes, and remained slightly negative during successive heating and cooling cycles. It is evident that if some form of dissolved nickel, represented in these experiments by the added nickel oxide, is present in molten sodium hydroxide, which 1s 1n contact with nickel surfaces at different temperatures, a potential will develop that is of the proper sign to correspond to oxidation (attack) of the nickel metal at the hot surface and to reduction (de- position} of nickel at the cooler surface. The process may be represented by the following equations: Ni (metal) ——3> (Ni**) + 2 electrons (at hot surface) , (Ni**) + 2 electrons——> Ni (metal) (at cold surface) . Since little is known of the actual state of the nicke}l in these melts, the term (Ni*") must be interpreted as being some 1on containing divalent nickel in a form capable of electro- chemical equilibrium with metallic nickel. The experiment with sodium hydroxide containing no added nickel oxide and a hydrogen atmosphere showed that under these conditionsno concentration of dissolved 1onic nickel develops 125 ANP PROJECT QUARTERLY PROGRESS REPORT and, hence, no potential corresponding to metal transport is observed. The potential of opposite sign that was observed hasnot been clearly accounted for. It would appear to be dependent upon some process involving the hy- drogen and sodium hydroxide, with the nickel electrodes serving only as carriers for the hydrogen gas and thus acting as hydrogen electrodes. The measurement of the thermo- galvaniec potential appears to offer a fundamental method of studying the sodium hydroxide system apart from any participation by the electrode metal 1tself. It 1s significant in relation to the corrosion and metal transport processes that the use of a hydrogen atmosphere cleans up traces of nickel oxide and leaves the melt free of dissolved nickel and hence incapable of developing the potential corresponding to metal transport. The work described establishes the existence of thermogalvanic potentials between nickel electrodes in fused sodium hydroxide containing dissolved nickel oxide. The potentials have been shown to be of such sign and magnitude as to be capable of accounting for transport of metallic nickel in nonisothermal nickel systems in which fused sodium hydroxide 1s circulated, Mechanism of Fluoride Corrosion (A. des Brasunas, L. S. Richardson, Metallurgy Division). A number of tests 1ntended to give a better understanding of fluoride corrosion has been completed. The postulate in the previous report(!®) that the subsurface voids encountered during fluoride corrosion are caused by the outward diffision of chromium atoms, (lglAdamaon, Brasunas, op. c¢it., OBNL-1294, p. 126. and Richsardson, 126 which leaves the metal enriched with vacant lattice sites that “precipi- tate” to form voids, has been sub- stantiated repeatedly. It has been demonstrated when chromium is re- moved from the metal lattice by either high-temperature oxidation or high- temperature vacuum treatment. Con- versely, corrosion tests made with fluoride containing small amounts of chromium powder have shown that void formation can be suppressed. The basis for this conclusion may be briefly summarized as follows: 1. Even with careful nonaqueous polishing, it has never been possible to retain particles in the porous surface region. 2. Chemical analysis of the at- tacked metal surface has revealed a substantial loss in chromium (from 15 to less than 5%).(29) 3. The chromium content of the fluoride abnormally high 1n comparison with the nickel and iron contents, which indicated a preferential solution of chromium from the alloy. was 4. Calculations of void area anticipated from change in chromium content are 1in accordwith observations. 5. Depth of void formation bears a direct relationship to the amount of chromium in the bath. 6. Attacked specimens show weight losses but no dimensional changes. 7. Chromium loss from the alloy caused by either high-temperature oxidation or high-temperature vacuum treatment has also resulted in similar void formation, (20)Brasunns and Richardson, op. cit., ORNL-1227, p. 124. FOR 8. Such voids have been observed previously(21+-22) in diffusion ex- periments in which diffusion in one direction exceeded the extent of diffusion in the opposite direction. 9. The addition of small amounts of chromium powder to saturate the fluoride bath with chromium has prevented void formation under con- ditions that otherwise favor void development. | 10. Fluorine 1s not detected below the metal surface, The porous surface region of an Inconel metal specimen before and after exposure to molten fluoride 1s shown in Fig. 52. Many of the voids, especially those not in grain bound- aries, have definite shapes that bear a relationship to the metal lattice. The inclusions shown in Fig. 52 are convenient sites for the deposition of vacancies; two such sites may be seen. The corrosion of Inconel by air 1is known to form a chromium-rich oxide layer at elevated temperatures. In a 200-hr exposure to air at 1200°C (2192°F), the region beneath the oxide layer of an Inconel specimen was found to contain voilids similar to those observed in fluoride-attacked Inconel (Fig. 53). The high vapor pressure of chromium relative to that of iron or nickel suggests that high-temperature vacuum treatment would also be a suitable means of removing chromium. Inconel and an 80% Ni-20% Cr alloy were exposed to a vacuum of 0.1 mm Hg for 42 hr at 1375°C. Both samples showed many subsurface voids; those in Inconel appeared to be spherical, whereas those i1n the 80% Ni-20% Cr alloy were L. C. €. da Silva and R. H. Meh!l, of Metals 191, 155 (1%951). (zz)uanthly Technical Progress Report for October 20 to November 20, 1951, SEP-82. Journal PERIOD ENDING SEPTEMBER 10, 1952 angular, as shown in Fig. 54. Chemical analysis of the entire Inconel speci- men indicated a drop in chromium content from about the usual 15% to 9. 2%. The mechanism of void formation in Inconel is therefore believed to occur in two stages. The first involves a loss of chromium atoms by diffusion to the metal surface and then into the fluoride bath. The resulting large number of vacant lattice sites 1s reduced to a more stable number by “precipitation” as 1llustrated 1n Fig. 55. For ease of illustration only six vacancies are shown to constitute a void. The number of vacant atoms involved in the voids shown in previous photo- micrographs is of the order of 10%2 atoms. Tripositive Nickel Compounds from Hydroxide Corrosion (J., V. Cathcart, W. H. Bridges, L. D. Dyer, B. Borie, Metallurgy Pivision). Two new tri- positive compounds of nickel have been identified as corrosion products of nickel attacked by sodium hydroxide in the presence of oxygen. The purpose of this investigation was twofold. First, in corrosion and mass transfer studies with the hydroxides, a variety of black and colored products has been obtained that do not correspond to any known nickel compounds. The occurrence of these products has been somewhat of a difficulty in the efforts to ascertain the mechanismof hydroxide corrosion. It was desired therefore to try to identify some of the more unusual corrosion products and de- termine some of their physical and chemical properties to aid in future identification and to give some clue as to how they might have been formed. The second purpose of this in- vestigation was to determine whether nickel ions of valence greater than 2 might occur in hydroxide melts, 127 ANP PROJECT QUARTERLY PROGRESS REPORT UNCLASSIFIED Y.6795 UNCLASSIFIED Y.6796 o corrosion of Inconel by Fluorides. Fig. 52. Voids BResulting from the 2000x. (b) Porous surface (a) Inconel specimen prior to fluoride attack. region of fluoride-attacked TInconel showing geometric-shaped voids and larger 2000x. irregularly shaped voids in grain boundaries and around inclusions. 128 FOR PERIOD ENDING SEPTEMBER 16, 1952 These higher valence ions are required in the very ingenious mechanism of nickel mass transfer proposed about a year ago.(?3) At that time the only nickel compounds of valence greater than 2 for which there was anything but the most tenuous evidence were four trivalent nickel oxvhydroxides and a hydrous oxide of gquadrivalent nickel. Briefly, the most important feature of the work was the preparation of two compounds whose empirical formulas are NaNiO, and LiNiO,. These compounds, which are named sodium nickelate(III) and lithium nickelate(III), have been identified by chemical analysis for nickel and for the alkali metals; by several chemical reactions, including (23)p, G, Hill and R. A. Bolomey, Aireraft ; . : : Nuclear Propulsion Project Quarterly Progress Flg" 53. Voids F"‘.”“e" Dourlng Alr Report for Period Ending September 10, 1851, Corrosion of Inconel at 1200°C. 250X, ORNL- 1154, p. 122, ‘LASSIFIED § B R o TUNCLASSIFIED § y.6826 | K y v o, . X.6827 Fig. 534. Voids in Inconel and an 80% Ni-20% Cr Alloy After Vacuum Treat- ment at 1375°C. (a) Inconel. (&) B80% Ni-20% Cr Alloy. 50X 129 ANP PROJECT QUARTERLY C CHROMIUM ATOM F IRON ATOM N NICKEL ATCM BEFORE LEACHING Fig. 585. Chromium Atoms. a quantitative determination of the oxidizing power; by a reaction that gives a known trivalent nickel com- pound; and by complete crystal struc- ture analysis. Solutions of Metals in Molten Halides (M. A, Bredig, J. W. Johnson, H. R. Bronstein, Chemistry Division). Considerable effort has been directed toward i1mproving the experimental technique so that more accurate determinations can be made of the solubility of metals tn the melts of their halides. With the improved technique, which does not appear to be subject to the previous objections, a solubility of 2.87 mole % potassium in the molten ternary eutectic LikF- NaF-KF at 880°C was obtained. The discrepancy with the value of 1.6 mole % reported previously, which 1s not too serious 1in view of the ex- perimental difficulties, 1is thought to be due mainly to removal of potassium metal 1n the earlier experiments by leaching the salt phase with alcohol. The leaching may not, as intended, have been confined to the dissolution of metal mot dissolved in the melt; the alcohol may also have attacked the dispersed in the solidified salt phase. Separation by alcohol leaching has been abandoned. Satis- factory separation of molten halide metal 130 VACANCIES CAUSED BY LEACHING PROGRESS REPORT UNCLASSIFIED DWwG. 18322 VACANCIES PRECIPITATED Sketeh Illustrating Formation of Subsurface Voids by Removal of and molten metal phases 1s now being achieved in capsules of 1mproved design that incorporate a ball check valve between the lower part containing the salt phase and the upper part containingthe metal phase and portions of the salt phase. Among other halidesystems, chlorides are recelving continued attention, In view of the possibility of chlorine isotope separation, such dystems may even be of direct interest for reactor applications. In the sodium metal- sodium chloride system, an increasing solubility of the metal from amounts of the order of 1 mole % near the melting point of sodium chloride to amounts of 20 mole % 200°C above the melting point was observed. The disagreement of such comparatively high solubility with the solubility predicted by a current theory of metal~metal halide solutions 1s being studied. Possible relationships between changes in structure and volume upon melting of the molten halides and the solubility of metal in them are beiling investigated. Fluoride Corrosion Phenomena (M. A. Bredig, H. R. Broustein, J. W, Johnson, Chemistry Division). A misleading statement appears in the previous report (ORNL-1294, p. 130, FOR PERIOD ENDING SEPTEMBER 10, fourth line from the “The heat of for- mation of ZrF,, AF = 445 + 30 kg/cal, 1s 10 that of KF, and the equilibrium concentration of FelF, at 1100°K, which is from 10°% to 1077, is one~thousandth that of KF,” 1t should read, “The free energy of right column, top); 1nstead of, times 1952 formation of ZrF,, AF ¥ -445 + 30 keal, yields for theequilibrium concentration of FeF, at 1100°K (1500°F) a value between 10°° and 10°°. The higher of these limits 1s ten times larger, and the lower one one-thousandth the Fer concentration 1in the case of corrosian by KF.” ' 11. W. D. Manly METALLURGY AND CERAMICS Jo M. Warde Metallurgy Division A systematic study of the basic welding variables associated with the cone-arc welding process is being made, The effects of arc-current magnitude and arc-current duration omn cone-arc welds are described 1n this report. The following welding specifications have been prepared for the joining of Inconel pipe and fittings for use with highly corrosive materials: procedure specifications for d-~c¢ inert-arc weld- ing of Inconel pipe and fittings, and the operator’s qualification specifi- cations for d-¢ inert-arc welding of Inconel pipe and fittings. The mechanical properties of high- temperature brazing alloys have been investigated, and the joint efficiency has been calculated by comparing the tensile strength of the brazed joint with the tensile strength of a speci- men of the parent metal that has undergone the same temperature cycle as that used in brazing. | The preparation of the components for use in the ARE control and safety rods has been completed., Sixty of the B,C~-Fe inserts and 29 of the Al1,0,-R,C inserts were made and canned by Nicro- brazing. The control rods and safety rods have been delivered to the ARE staff for calibration. The effect of small differences 1in test temperature on the stress-rupture life and creep rate of Inconel has been studied. The creep properties of Inconel when tested in argon and azir have been measured, and it was found that Inconel tested in an air atmos- phere has a much longer stress-rupture life, A satisfactory high-temperature ceramic coating has bheen applied to nickel sheet to be used in a ligquid metal-to~air radiator. The possibility of the development of a glass-bonded beryllium oxide material for use as a reflector moderator 1s being investi- gated, CONTROL ROD FABRICATION E. S, Bomar J. H. Coobs Metallurgy Division The preparation of components to be used 1n the ARE regulating and safety putlined pre- been completed., A on work has based (1) rods, viously, (1)E. 5. Bomar and J). H, Ceobs, Adircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending Jaurne 10, 1952, ORNL- 1294, p. 135; and Metallurgy Division Quarterly Progress Report for Period Ending April 30, 1952, ORNL-1302, p. 78. ANP PROJECT QUARTERLY PROGRESS REPORT total of 60 inserts composed of boron carbide and 1ron were pressed; 54 inserts were needed for three safety rods. Thirty-three cylinders composed of dilute mixtures of B,C and Al,0, were fabricated for use in two control rods with different B,C investments, These two rods required 29 inserts, The surplus parts were prepared for use in case of breakage, Only minor difficulties were encountered 1in producing the inserts, The first cylinders prepared cracked along their length during the cooling cycle following hot pressing. This difficulty was avoided by ejecting the graphite mandrel from the center of the slugs while they were still at temperature., Another difficulty in- volving differences 1in expansion co- efficients was encountered when the first insert was sealed brazing. in a can by The clearances originally set up were not adequate to allow for the expansion of the inner stainless steel liner when the assembly was heated to 1120°C for brazing. Consequently, the cylinder broke excesslive 1nternal loading. the inside diameter cylinders by 0,015 1in. because of Increasing of the pressed prevented this ceramic type of failure during the canning operation, The percentage of B,C in the B,C~Al,0; cylinders was 1ncreased to compensate for the decreased volume accompanying the increase in inside drameter. Densities of 80% of those of the theoretical mixtures were obtained 1in pressing full-size parts of the two types of materials., The B,C-Fe cylinders were pressed with 2500 psi of pressure at 1520 to 1540°C, The A1,0,-B,C inserts were compacted at the same pressure but were heated to 1650°C., In both instances graphite dies surrounded by bubble alumina insulation and an 1nert atmosphere of argon were used, 132 MECHANICAL TESTING OF MATERIALS R. B. Qliver D. A, Douglas G. M, Adawmson K. W. Beber Metallurgy Division J. W, Woods Tests have been initiated to deter- mine the effect of stress corrosion by molten fluoride salt mixtures on the creep rate and stress-rupture life of structural materials., The creep and stress-rupture tests 1n argon reported in the previous report(?} have con- tinued, and new data for Inconel obtained. were Physical Testing in Fiuorides (G. M. Adamson, K., W, Reber, Metallurgy Division)., In the physical testing program of the joint experimental engineering section of the ANP Divisiocn and the Metallurgy Division, NabF-KF- LiF-UF, continued to be the fluoride mixture tested. It will not be possible to test any of the zirconium-base fluorides in the present apparatus because of their high vapor pressures, When the emphasis was shifted from sodium to fluorides 1t was necessary to use the sodium-testing apparatus, since speed was a prime consideration of this program. The sodium-testing apparatus was constructed from types 316 and 347 stainless steel. 1Inconel was the material most likely to be used 1n the ARE, so 1t the first material tested. The photomicrographs from the initial tests and they show that in every case 1n which the material in the test for an appreciable was of specimens have been received, was mass-transferred This is true of both the stress-rupture and tube-burst speciwens, All work waith Inconel has been stopped until TInconel pots and sample holders are obtained from the shop. length of time a layer appeared on the specimen, (2)R. B. Oliver, D. A. Douglas, . W. Weaver, and J. M. Woods, op. cit., ORNL-1294, p. 136, FOR Jnconel stress-rupture specimens have been tested in Nab-KF-LiF-UF, at 1500°F under stresses of 6000 and 7000 psi. A plot of the rupture times of 131 and 94 hr falls on the line through previously obtained values, These tests were run in type 347 stain- less steel pots. Stress-rupture tests of type 316 stainless are now being set up. steel samples Several tube-burst tests have been run with type 316 stainless steel specimens. The rupture times obtained from these tests did not agree well and were much lower than expected. The tube-burst tests have consistently praoduced low results with both Inconel and type 316 stainless steel. Tests now being set up with better control to deterwine are temperature whether this 1s the difficulry. An apparatus was designed for carrying out self-welding tests i1n the fluorides. The final design was an which, by the present means of a stress - apparatus 1in simple adapter, rupture equipment could be used. The apparatus will allow the two tesct specimens to be held apart until the fluorides have time to condation the surfaces. The specimens can then be brought together under any compressive load Self-welding will be determined upon completion of the test by applying a tensile load while the apparatus 1s still hot., The necessary to separate the samples will be a measure of the amount of self- welding. This procedurewonld alleviate desired. stress PERIOD ENDING SEPTEMBER 10, 1952 the difficulties encountered 1n remov- ing and cleaning the samples. Creep and Stress-Rupture Tests 'in Argon. The stress vs., creep curves for fine- sheet tested at and coarse-grained Inconel 1500°F in argon pre- sented 1w a previous.report(a) have been revised and are presented in Figs, 56 and 57. It has been suspected that a small difference in the test temperature could have a noticeable effect on the rupture time and perhaps also on the the two creep rate. Table 28 summarizes data on tests now 1in progress at temperatures. The Inconel sheet specimens tested 1n argon exhibit a shorter rupture life than reported by the International Nickel Company for similar tests con- ducted 1in air. Two specimens are being tested 1n air to resolve this inconsistency. The data from these tests are summarized in Table 29, WELDING P. Patriarca G. M. Slaughter Metalluregy Division The systematic study of the cone- arc welding process is continuing with emphasis during the last quarter on (313. B, Oliver, J, W. Woods, and C. W, Weaver, Aircraft Nucleor Propulsion Project Quarterly Pragress Report for Period Ending March 10, 19592, OBNL-1227, p, 149 and 150, Figs. 46 and 47, Table 28 CREEP AND STRESS-RUPTURE TESTS ON INCONEL IN ARGON TEMPERATURE STRESS TOTAL ELONGATION AT CREEP RATE (°C) (psi) 1000 hr (%) (% per hr) 815 2000 0.5 0.0007 821 2000 2.0 0.0040 133 ANP PROJECT QUARTERLY PROGRESS REPORT UNCLASSIFIED DWG.15684 A CREEP RATE (% perhr) G 1 10 0.000 0oet oo 20,000 =g oo \ ‘ 0% | 0.5% 10% 2% 5% RUPTURE | | | % [EXTENSION AT Ti YT 10,000 T‘é %5000 g b LT L uJ o = w 2,000 ———t L bl b IS NN hooo[w“—n» ol 02 05 10 2 5 10 20 50 100 200 500 1000 2000 5000 10,000 TIME (hr) Fig. 36. Stress vs. Creep Rate for Fine-Grained Inconel Sheet Cold Rolled at 1650°F and Tested in Argon at 1500°F. Grain size: approximately 90 grains 0.105 mm in diameter. Times were measured for 0.1, per square millimeter, Extension measured 0.5, 1, 2, and 5% extension and rupture vs. Stress. optically. UNCLASSIFIED | CREER RATE (% per hr) EWG 156854 Q000! .00 00 o1 10 20000 | T | ‘ \l ! ‘ ; T C T T T | i;l o | ] ‘1’ L] | | | | Q, 1 o, o, I 1 i | 0 1%, 05% | /ni 2% D% 33 % (EXTENSION 4T TIME OF Rup*rur‘c‘i)‘ ‘ ; l i ‘ . B ! | i ( | ! 10000 gt — b - 5 RUPTURE 2’ L ll Ll . f P e B e S T e L) . 4 ! hn.!ljh m“m”-mL,=1fl_mn P~ . [ 315 5 . | & W O 5000!— [vel = o) 2000t - [~ : w 1000 J ) H | 500 1000 2000 ol TIME {br) Creep Rate for Coarse-Grained Inconel Sheet Annealed Grain size: approximately Times were measured Extension was Fig. 57. Stress vs. for 1 hr at 2050°F and Tested in Argon at 1300°F. 15 grains per square millimeter, 0.250 mm in diameter. for 0.1, 0.5, 1, 2, and 5% extension and rupture vs. stress. measured optically. 134 FOR PERIOD ENDING SEPTEMBER 10, 1952 Table 29 COMPARISON OF RUPTURE TIMES OBTAINED FOR INCONEL SHEET RUPTURE TIME (hr) _ ' STRESS ‘ - OBNL TEST IN AIR. (psi) OPNL. Test in International Nickel Co. {time to date, hr) Argon Test in Air ‘ 3500 700 - 6000 2475 4000 400 3000 1711 the effect of the initial surge on the resulting weld. Specifications have been prepared for the welding of com- ponents for the ARE, Cone-Arc Welding., The feasibility of the application of the cone-arc welding process to fabrication of tube-to-header heat exchangers was determined by experiment in previous work. (%) Tt has been recognized, how- ever, that the systematic determination of the behavior of the basic welding variables associated with the cone-~arc welding process would contribute toward a better understanding of possaible applications and limitations. The effects of arc-current magnitude and duration on the inside and outside diameters (Di and DU) of the weld and on the minimum weld penetration (P) have been determined. The initial arc current i1s generally of the order of 100% greater than the steady-state value and is the subject of another study now 1n progress. As in the previous work, the header configuration and material sizes con- sisted of 0,100-in.-0B, 0,010-3in. - wall, type 316 stainless steel tubing and 0,125-in.-thick, type 316 stain- less steel header sheets containing 19 tube holes drilled one diameter apart at the apex of an equilateral triangle, which was the basic pattern. All (4 i ' )P. Patriarca and G. M. Slaughter, op. cit ORNL-1294, p. 138, tubes were set flush with the header surface prior to welding. A series of welds were made using the conditions outlined in Figs. 58 and 59, the variables being arc current and arc time, respectively. Examination of Fig, 58 will reveal that the predominate effect of in- creasling arc current 1s to 1ncrease the value of D . This, coupled with the relatively small increase in weld penetration P and the equally small decrease 1in Di’ indicates that the major portion of additional energy introduced as increasing arc current served to increase the 1ineffective weld area. It may be noted, however, that cone-arc welds may be.made over a relatively wide range of conditions, but that operation in the higher current range should be avoided since greater header distortion becomes evident with higher energy input, The results of a similar study are presented in Fig. 59, which presents arc time as the variable. The sima- larity of Figs. 58 and 59 1s evident. Analysis of the curves in Fig., 59 will reveal that arc time 1s not a critical variable within the limits of 1.5 to 4.2 sec under the other welding con- ditions indicated, Specifications for ARE Welding. The following specifications have been prepared for use in the fabrication of 135 ¢l WELD DIMENSION (in.) 0.20 O o MINIMUM CURRENT FOR ACHIEVEMENT OF COMPLETE PERIPHERAL WELD UNCLASSIFIED DWG. 6380 |0, OVERLAP| ! AND \i (01 SENSITIVE | PERMISSIBLE OPERATING RANGE RA NGE | HEADER HOLE CLOSED 0, P 40 50 60 70 ARC CURRENT {D-C amp) Fig. 58, Effect of Arc Curvent on Cone-Arc Weld Parameters Do. Di. Material: tubes, type 316 stainless steel, 0.10 in. OD, 0.010-1q. w header, type 316 stainless steel, 0.125 in. thick tube hole separation, 0.10 in. (one tube diameter} Welding Conditions: arc time, 1.5 sec. arc distance, 0.05 in. high-thoria tungsten filament, 0.0625 in. in diameter {pointed) argon gas fiow, 30 cfh 8C and P. all LAT6Ud dNV LHOdAY SSAUs0Hd A'HALEVNI0 WELD DIMENSION (in.) LET UNCLASSIFIED DWG, {638t MINIMUM ARC TIME FOR ACHIEVEMENT OF COMPLETE Dq — 0.20 ___ PERIPHERAL WELD HEADER HOLE CLOSED | D, OVERLAP RANGE D, SENSITIVE PERMISSIBLE OPERATING RANGE _, __ RANGE { 2 3 4 5 ARC TIME (sec} Fig. 39. Effect of Arc Time on Cone-Arc Weld Parameters D_, D, and P. Material: tubes, type 316 stainless steel, 0.10 in. OD, 0.010-in. wall header, type 316 stainless steel, 0.125 in. thick tube hole separation, 0.10 in. (one tube diameter) Welding Conditions: arc current, 5.9 amp d.c. arc distance, 0.05 in, high-thoria tungsten filament, 0.0625 in. in diameter (pointed) argon gas flow, 30 cfh ‘0T HAGWALJAES INIANT aoI¥dd HOoA cSo6l the ARE: procedure specification for d-c¢ inert-arc welding of Inconel pipe and fittings for high-corrosion appli- cations; operator’s gualification test specification for d-c¢ 1inert-arc weld- ing of Inconel pipe and fittings for high-corrosion applications. TESTS OF BRAZING ALLOYS P. Patriarca G. M. Slaughter Metallurgy Divaision The tensile strength and static corrosion tests of brazed joints are continuing, (%) The corrosion tests are being extended to include static tests of brazed joints in moist air and tests of nickel joints 1n hy- droxides. Brazing efficiencies as high as 92% for butt-brazed joints have been demonstrated in tensile tests, The brazing of molybdenum 1is being i1nvestigated, An additional brazing research facility has been provided by the recent installation of a large G-E brazing furnace for dry- hydrogen bhrazing operations, Static Corrosion Tests of Brazed Joints. The static corrosion testing of brazed joints for the high-tempera- ture brazing alloy evaluation program is continuing. The 16,5% Cr-10.0% Si~2.5% Mn~71,0% Ni alloy developed by the General Electric Company 1s attacked rather severely by the fluoride mix- ture NaF-KF-LiF-UF, when brazed both on Inconel and type 316 stainless steel, Photomicrographs of these corroded joints are shown in Fig. 60, nickel has been found to be relatively unattacked in molten sodium hydroxide, static corrosion test specimens of this material brazed with several of the promising alloys have been Since A prepared. Examination of the D rbid., p. 140. 138 samples has not been completed, but several photomicrographs of the as- brazed joints, particularly the joint brazed with 60% Pd, 40% Ni, and Nicrobraz, show excellent wettability and flow characteristics. The resistance of brazed joints to oxidation at high temperature 1is an important factor, so tests are being conducted on brazed jolnts to deter- mine the amount of deterioration. Butt-brazed joints of Inconel, type 316 stainless steel, and A nickel have been prepared by using several of the brazing alloys being investigated, Many of these have bheen tested in an atmosphere of moist air at 1500°F, but metallographic examination of these jJoints 1s 1in the preliminary stage. Effects of oxidation at higher tewmper- atures will also be investigated, but the tests are presently being run to provide a source for the screening of the less desirable alloys, Tensile Strength of Brazed Joints. In order that an approximate tensile strength of joints butt-brazed with the 60% Pd-40% Ni alloy could be obtained, a few preliminary tensile tests were made on these joints prior to a more thorough i1nvestigation, pending the arrival of an appropriate high-temperature extensometer, Standard tensile bars 0,505 in. in diameter were machined from butt-brazed type 316 stainless steel tensile blanks., Three tensile bars were tested at room temperature and three were tested at 1500°F, Test bars of type 316 stain- less steel, which were subjected to heating cycles similar to the brazing cycles, were also pulled as check samples on the strength of the parent metal. The results of these tests are summarized in Table 30, The fracture of the butt-brazed joint at room temperature occurred along the diffusion zone of the joint, R 070 v-66298 & Fig. 60. Joints Brazed With 16.5% Cr~198.0% Si-~2. 9% Mn~731.9% Ni Alloy and Tested in NaF-KF-LiF-UF, (10.9-43.5-44.5-1.1 mole %) for 100 hr at 150D°F. (e¢) Type 316 stainless steel joint showing moderate attack of the brazing alloy. (b) Inconel joint showing severe attack of the alloy. Unetched. 75X, 139 ANP PROJECT QUARTERLY PROGRESS REPORT Table 30 TENSILE STRENGTH OF TYPE 316 STAINLESS STEEL BRAZED WITH 60% Pd-49% Ni ALLOY CHECK BARS BRAZED JOINT TEMPERATURE BRAZING EFFICIENCY (°F) Tensile Strength | Elongation | Tensile Strength | Elongation (%) (psi) (%) (psi) (%) Room 75,400 40.0 82,000 75.0 92 1500 22, 400 18.4 25,600 87 The fracture of the joint brazed at 1500°F resulted in a more ragged surface and occurred frequently along the braze metal-base metal interface. However, all these data suggest that the 60% Pd-40% N1 alloy is very promising, although it is sti1ll ex- tremely desirable to strive for lower- ing of the melting point by alloying. Only one brazing alloy has been investigated by using the larger 0.252-1n,-dia butt-brazed tensile bars. A high-temperature extensometer de- with this size of and work on this si1gned for specimen has arrived, use phase of the brazing alloy evaluation program should proceed much more rapidly. A series of tensile tests at room temperature and at 1500°F were completed for Inconel brazed with the 75% Ag~20% Pd-5% Mn alloy. The Inconel check bar has not yet been tested., At room temperature, the average tensile stress of the brazed joint was 55,900 ps1, and the average elongation 1in a l-in. gage length was 9%. At 1500°F the average tensile stress of the brazed joint was 19,700 psi, and the average elongation was 2.5%. These results apparently mean that the brazed joint decreases 1n strength with 1ncreasing temperature more rapidly than does the base Inconel. The elongation of the base metal is quite large at room temperature, whereas at 1500°F there is very little elongation. 140 Brazing of Molybdenum. In view of the interest shown in the fabrication of molybdenum, it was decided to conduct a limited program on the brazing of this metal with the various high-temperature alloys presently being investigated. The wettability charac- teristics of the various brazing alloys on bright-annealed molybdenum sheet in a dry-hydrogen atmosphere are listed in Table 31. A further indication of the strength of molybdenum brazed joints should be obtained from the tensile test speci- mens to be prepared in future work. If a small guantity of ductile mo- lvbdenum can be obtained, tensile tests will be made on this material, since morereliable data should result, Dry-fiydrogen Brazing Furnace. A 24-kw, 312-amp, G-E brazing furnace has been i1nstalled for use 1in dry- hydrogen brazing operations. A “Square D" stainless steel muffle, welded shut at one end, was built 1into this furnace for the purpose of ob- taining the atmospheres required for many furnace-brazing problems, The physical dimensions of this muffle are such as to accommodate a sample of the following dimensions: width at base, 6 in.; height at center, 5 in,; length to be heated at one time, approximately 4 ft (hot zone of furmace); total length, approximately 9 ft {(includes a 5-ft cooling .hamber). A stainless FOR PERIOD ENDING SEPTEMBER 10, 1952 Tablie 31 WETTABILITY CHARACTERISTICS OF VARIDUS BRAZING ALLOYS ON BRIGHT-ANNEALED MOLYBODENUM SHEET IN A DRY-HYDROGEN ATMOSPHERE BRAZING ALLOY WETTING PROPERTIES 60% Pd-40% N1 Nicrobraz 16.5% Cr-10.0% S1-2.5% Mn-71.0% Ni 50% Pd-40% N1-10% Mn 40% Pd-40% N1~20% Mn 60% Pd-37% N1~-3% Si 73.5% Ni-16.5% Cr-10,0% Si 75% Ag-20% Pd~5% Mn 64% Ag-33% Pd-3% Mn Excellent Fair Good Very good Gond Excellent Excellent Excellent Excellent steel pipe was welded into the sealed end of the muffle to permit the entry of dry hydrogen. A system, Fig, 61, for providing a dry-hydrogen atmosphere is attached to this furnace. Included in this system is a nitrogen cylinder (to purge the hydrogen lines of air), a Deoxo Duridryer,{®) and a Lectrodryer dehumidifier.¢’? The dryer converts oxygen to water and the dehumidifier removes the water., Dew points of ~120°F can be obtained, although -60°F is satisfactory for most applications. When the need arises for the furnace brazing of assemblies larger than the stainless steel muffle permits, 1t 1s convenient to use canning techniques. Such techniques consist of enclosing the assembly in a stainless steel can, which 1s then welded shut, Stainless steel entrance and exit tubes are provided in the can for the hydrogen “”Baker and Co,, Newerk, N. J. 7 ¢ )Pitcflburgh Lectrodryer Corp., Pittsburgh, Pa, flow. located entirely in the stainless can, any furnace of proper size and temper- ature limits can be used as the source of heat. The sodium-to-air heat ex- changer, Fig. 62 (approximate di- mensions, 12 by 6 by 8 in.) is typical of the assemblies that can be brazed by this canning technique. With the controlled atmwosphere Nicrobraz powder was placed on all of the joints, including approximately 2500 tube-to-baffle plate joints. Good hydrogen flow over the entire assembly was needed for proper flow of the brazing alloy, and 1t 1s expected that a series of baffle plates 1inside the can will be required to attain this flow for brazing of similar, but larger, heat exchanger units, REDUCTION OF MOLYBDENUM BISULFIDE G, P. Smith, Metallurgy Division Simple thermodynamic calculations have been made in an effort to deter- mine 1in a preliminary fashion the 141 ANP PROJECT QUARTERLY PROGRESS REPORT UMCLASSIFIED DWG. 16537 DEOXQ - - = PURIDRYER H2 FLOWMETER N2 FLOWMETER N2 H(2 H2 L_-._.l.,.--.J ST TSBURG TO MOLYBDENUM LECTRODRYER FURNACE WINDINGS TYPE BAC—0 BOTTLED GAS MANIFOLD (mxn : OUTLET FOR DEW-POINT MEASUREMENTS CEE) = TWO-STAGE REGUL ATOR TG FURNACE MUFFLE Fig. 61. HManifold System for Dry-Hydrogen Furnace Brazing. SR R R R AR, S : 2 S . . UNHA%gIlED:, SR e possibility of reducing molybdenum disulfide with hydrogen or zinc. These calculations show that flowing hydrogen should be capable of reducing molybdenum disulfide at temperatures 1in the neighborhood of 1000°C, Solid zinc is thermodynamically capable of reducing molybdenum disulfide and would probably work as well as molten zinc, The reduction of molybdenum di- sulfide by hydrogen would be ac- complished by the following reaction: MoS, + 2H, = Mo + 2H,S , from which Fig. 62. Heat Exchanger Test Unit . Nicrobrazed by Camning Techmigque. 142 FOR PERIOD ENDING SEPTEMBER 10, 1952 where p 1s the equilibrium pressure and K 1s the appropriate equilibrium constant, as implicitly defined by the above equation, Values of K are given by Kelley. (%) For a hydrogen pressure of 760 mm the equilibrium pressure of H,S at 910°C 1s 6.65 mm, and at 1100°C 1t is 14.5 mm, These are guite appreciable pressures, and i1t should be easy to reduce molybdenum disulfide at these temperatures if the reaction rate is at all reasonable. Tt would be neces- sary to remove H,S from the reacting mixture, but this could be accomplished with a current of hydrogen. The reduction of molybdenum di- sulfide by zinc can be represented by the following equation: ' MoS, + 2Zn - > Mo + 2ZnS . The standard free energy change for this reaction in the case of solid zinc may be obtained from the data given by Kelley.,(®) AFC = 28,340 - 2,567 log T + 2.40 x 107312 +92.198 x 10737 + 9,121 . Thus at 25°C, AF° = ~26.6 kcal/g-atom of MoS,. The standard free energy equation does not exist, and calcu- lations of AF° at elevated temperatures would guantitatively be somewhat deceiving., However, the reduction 1s thermodynamically possible at 25°C, and the AF® is sufficiently sub- stantial to warrant consideration of this methed of reduction at elevated temperatures. (8)K. K. Kelley, Contributions to the Data on Theoretical Metaliurgy, Bureau of Mines Bulletin 406 (1937), p. 53, ' (9 ybid., p. 54, 66. Eq. 305 onm p. 54 is incorrect multiplied by T. The I factor as given in and should be CERAMICS RESEARCH J. M. Warde, Metallurgy Division A ceramic coating to inhibit oxidation at elevated temperatures has been successfully applied to nickel to be used in the construction of a radiator for the ANP. Beryllium oxide shapes for use in ANP reactor experi- ments have been made and fired at the U, 8. Bureau of Mines Electrotechnical Laboratory at Norris, Tennessee., Pre- liminary work on the development of a glass-bonded beryllium oxide material for use as a reflector-moderator for & liguid-cycle aircraft completed, reactor was Ceramic Coatings forMetalis. Further work was carried out on the application of a ceramic coating to nickel to provide oxidation resistance at elevated temperatures. Successful application of such a coating would make it possible to use nickel in a radiator. It was found that by annealing specimens of pyrometallurgi- cal nickel, 10 mils thick, 3In an atmosphere of wet hydrogen for 65 hr at 1000°C, good adherence was obtained with National Bureaun of Standards ceramic coating A-418, 1 to 2 mils thick, applied at 1700°F., The frig composition of this coating has been described in a previous report,(!?) Specimens coated with the A-418 com- position showed no signs of oxidation in a 9-day oxidation test at 1500°F, It i1s proposed that in the near future this coating be applied to actual radiator parts fabricated from nickel. Ceramic Reftlector. To satisfy the requirements for an efficient and economical material for use as a reflector-moderator for a ligquid-cycle aircraft reactor, a glass-bounded beryllium oxide material has been proposed, This material would have 143 ANP PROJECT QUARTERLY PROGRESS REPORT nearly zero porosity and would contain a minimumof 75 mole % beryllium oxide. The glass forming the bond would also contain a small amount of beryllium consistent with viscosity re- Simulated mixes using soda-lime-silica glass and silica sand were prepared on a small scale, and it was determined that such a material could be made successfully, The density of a selected, glass-bonded beryllium oxide, reflector-moderator material was determined as 2.6 g/cc. It is proposed that the reflector be cooled by metal pipes embedded in the oxide, quirements for fabrication. glass~bonded beryllium oxide. The pipes would contain flowing sodium as a coolant, and once the glass-bonded beryllium oxide had been ponrred into the shell and had flowed around the cooling pipes, the reactor shell would not be allowed to cool below a temper- ature just above the annealing range of the glass bond. Further refinements and engineering tests will be made, should this material be of interest, Among these would be the determination of the devitrification tendency of the glassy matrix and the corrosion effect of the glass with selected metals, 12. HEAT TRANSFER AND PHYSICAL PROPERTIES RESEARCH H. F. Poppendiek, Reactor Experimental Engineering Division Viscosity measurements have been made for two zirconium-bearing fluoride fuel mixtures. For the probable ARE fuel NaF-ZrF,-UF, (46-50-4 mole %) the viscosities range from 20 centipoises at about 580°C to 7 centipoises at about 830°C, and for the fuel Nalk- ZrF,-UF, (50-46-4 mole %) the vis- cosities range from 17 centipoises at about 580°C to 6 centipoises at about 830°C. A capillary viscometer is being assembled and will be used to obtain data to supplement the data from the existing viscometers, Some preliminary thermal conductivity experiments on the fuel NaF-KF-LiF-UF, (10.9-43.5-44.5-1.1 mole %) have vielded the conductivity value 2.2 £ 0.2 Bru/hr:ft? (°F/ft) over the temper- ature range 500 to 750°C, The vapor pressures of Zr¥F, and ZrF, -bearing fuel mixtures are being determined. Zirconium tetrafluoride is the most prevalent vapor phase above Zrl,- bearing fuel to the extent of about 10 mm Hg at 800°C, Enthalpy and heat capacity have been determined for lithium hydroxide. 144 Experimental, forced-convection, heat transfer data for moelten sodium hydroxide flowing turbulently in circular tubes were represented by the equation Nu = 0,021Re®:8pyp0.4 over the Beynolds number range of 6,000 to 12,000, This equation 1s within 9% of the one normally used to describe convective heat transfer of Information on lengths was also ordinary fluids. thermal entrance determined, Some experimental, pool-boiling data for a mercury system have been obtained for several pressure levels and two different kinds of heat- trans fer-surface materials, Heat and momentum transfer analyses have been made for the ANP thermal- convection harps for predicting the liquid circulation velocities. This information is needed in the interpre- tation of the harp corrosion results. In the one case examined, the circu- lation velocity was approximately 0.1 ft/sec. FOR PERIOD ENDING SEPTEMBER 10, VISCOSITY OF FUEL MIXTURES The viscosities of two ZrF,-bearing fuel mixtures have been determined on the rotational viscometer. Some viscosity measurements are now being made for NaOH and NaNQO; by using the efflux viscometer., In addition, the importance of viscosity of fuel and moderator-coolant mixtures in design of the ARE components has prompted study of additional methods for de- termination of this property at high temperatures, An improved capillary viscometer has been constructed. Also, a new structure has been completed that will permit the operation of the Brookfield viscometer, the efflux viscometer, and the density device when working with toxic materials, NaF-ZrF,-UF, Fuels (R. F. Redmond, D. F., Smith, T. N. Jones, Reactor Experimental Engineering Division). Viscosity measurements have been made with the Brookfield rotational vis- cometer for the zirconium-bearing fuel mixtures NaF-ZrF, -UF, (46-50-4 mole %) and NaF-ZrF,-UF, (50-46-4 mole %), and they have been plotted in Fig. 63. The viscosities of both mixtures are SECRET DWG 16366 -ZrF, */UFq NaF-7rE, -UF, (46-50-4 mole %) (50~-46-4 mole VISCOSITY (centipoises) Naof ~KF -LiF-UF, (10.2-42.5-44 5-1.1 mole %) 750 BOO 850 900 950 1000 1050 1100 TEMPERATURE ' (°K) Fig. 63. viscosity of Na¥F-ZrF,-UF, Salt Mixtures. 1952 approximately twice as high as that of the NaF-K¥-LiF-UF,. Capillary Viscometer (F. A. Knox, N. V. Smith, F. Kertesz, Materials Chemistry Division). The method employed in the capillary viscometer(?) consists of forcing the liquid through the measuring tube with the help of a vacuum pump, which establishes a pressure differential between the outside and inside liquid surfaces. The pressure difference is automati- cally stabilized at predetermined levels through a circuit that ‘has leads into a mercury manometer, - The ligquid travels upward through the calibrated tube and reaches a wider tube in which three electrically in- sulated metal probes are immersed. The time elapsed between reaching a lower and a higher level 1s measured with an electric timer connected to these leads. Upon completion of the time measurement, a relay closes the valve to the vacuuw pump and opens a valve to compressed argon, the pressure of which can also be regulated, The liguid is then forced back to 1its original position and the apparatus is ready for another measurement., The temperature 1s regulated and recorded during the experiment with a Micromax recorder, A 5-gal container 1is used te reduce the pressure fluctuations of the vacuum pump. During the whole operation the surface of the melt 1s protected with a blanket of argon from contamination by the air, Preliminary trials with the apparatus have indicated satisfactory performance. Calibration studies utilizing a number of fused salts are now 1n progress. In the only experiment to date on material of interest, the viscosity of the mixture containing 50 mole % NaF, 46 mole % ZrF,, and 4 mole % UF, was found to be about 13 centipoises at (l)H. Proc. Roy. Soc. Bloom, B. 5. Harrap, and E. Heywmann, {London) 194A, 237 (1948}, 145 ANP PROJECT QUARTERLY PROGRESS REPORT 600°C and about 4 centipoises at 800°C, These data must be considered tentative until the calibration studies completed and evaluated. are THERMAL CONDUCTIVITY OF LIQUIDS L. Cooper S. J. Claiborne W. D. Powers R. M. Burnett Reactor Experimental Engineeraing Division Preliminary thermal conductivity values have been obtained for the fuel NaF-KF-LiF-UF, by using apparatus described previously,(2?) The values obtained and some of the data on the fused salts studied previously are given in Table 32, From the data of Table 32 1t appears that the thermal conductivity increases as the per- centage concentration of the heavy UF, compound 1s decreased in the salt mixture. The ZrF,-bearing fuels are currently being studied. A thermal conductivity device (a modification of apparatus previously described¢3) for use with solids) that 1s particularly useful for studying liguid metals has been used to check values for liguid sodium; the measure- ments obtained were within 20% of the (Z)L. F. Basel and M. Tobias, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending December 10, 1950, ORNL-919, p. 196, (B)M. Tobias, Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending March 10, 1551, ANP-60, p. 243, values given in the literature. Pre- liminary data for the lead-bismuth eutectic have been obtained. is currently being studied. Lithium A transient hot-wire method 1s be- ing studied in connection with thermal conductivity measurements, This method, which i1nvolves 1imposing a transient temperature daistribution on a wire surrounded by the fluid to be investigated, appears tobe well suited to the study of the molten fluorides and hydroxides. HEAT CAPACITY W, D. Powers C. G, Blalock Reactor Experimental Engineering Division The enthalpy and the heat capacity of lithium hydroxide have been deter- mined by Bunsen ice calorimeters for the temperature range of 500 to 900°C and are given by the equations H (liquid) - Hyo (solid) = 0.85T + 110 , (1) c, = 0.85 % 0.06 , (2) where H i1s in cal/g, T in °C, and ¢ . P in cal/g°C. Some of the 12-1in, for the calorimeters have been re- placed by 24-in. furnaces that have yielded more uniform capsule tempera- tures; this modification has reduced heating furnaces Table 32 THERMAL CONDUCTIVITY VALUES FOR VARIOUS FLUORIDE SALT MIXTURES TLUORIDE SALT MIXTURE COMPOSITION (mole %) THERMAL CONDUCTIVITY, Btu/hr- ft? (°F/ft) NaF-KF-UF, 46.5-26.0-27.5 NaF-KF-LiF-UF, 10.9-43. 5-44.5-1.1 Nat'-KF-LiF 11.5-42.0-46.5 0.53 at 550°C < ¢t < 750°C 2.0 to 2.5 at 500°C < ¢t < 750°C 2.6 at 500°C < t < 750°C 146 FOR PERIOD ENDING SEPTEMBER 10, the scatter of the enthalpy data. Several fluoride fuels with Sr(OH), added and the LiCl-KCl eutectic are currently being studied. DENSITY D, F. Smith Reactor Experimental Engineering Division Preliminary density data on Nal- Zr¥,-UF, (46-50-4 mole %), NoF-ZrF, - UF, (50-46-4 mole %), NaNO;, and molten NaOH are being obtained. A mean value for the density of the 50-46-4 mole % mixture at 640°C < ¢ < 870°C is 3.3 g/ce. The BeF,-bearing mixtures are to be studied next. VAPOR PRESSURE OF FUEL CONSTITUENTS R. E. Moore Materials Chemistry Division The vapor-pressure data for solad Zr¥, and for two typical ZrF,-bearing fuel mixtures have been obtained during the past guarter by use of techniques 4 e (4,5,6) and apparatus previously described. Zirconpium Tetrafluoride. Since salt mixtures containing ZrF, are under consideration as fuels, and since ZrF4 is the only important component of the vapor phase above these molten salt mixtures, a determi- nation of the vapor pressure of solid zirconium tetrafluoride was made 1n the temperature range 722 to 854°C, The (4)R. E. Moore and (. J. Bartom, Aircraft Nueclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1931, OBNL-1134, p. 136, (S)R. E. Moore, Aircreft Nuclear Propulsion Project Quarterly Progress Report for Perieod Ending December 10, 1951, ORNL-~1170, p. 126, (6)H. E. Moore, Aircra)ft Nuclear Propulsisn Projeet Quarterly Progress Report for Period Ending June 10, 1852, OBNL-1294, p. 158. 1952 data, given in Table 33, represented by the equation may be 11320 T(°K) + 12.46 , log P(mm Hg) = -~ The heat of sublimation is 52 kcal/mole and the sublimation point 1s 908°C, as calculated from the equation. The value for the sublimation point 1s 1n fair agreement with the estimate (927°C) found in the literature,’’’ Table 33 YAPOR PRESSURE OF ZIRCONIUM TETRAFLUORIDE TEMP ERATURE OBSERVED PRESSURE (°c) {mm Hg) 729 13.5 741 21 743 22 754 27 776 43 777 48 778 45.5 854 278 Zirconium-Bearing Fluoride Mixtures, In a previous report %) the vapor pressure of a mixture containing 42 mole % ZrF,, 51 mole % KF, 5 mole % NaF, and 2 mole % UF, was reported, Since then interest has centered on the system ZrF, -NaF-UF,, and vapor pressure data for two compositions in this system have been obtained. The data for the solution containing 50 mole % Zr¥F,, 46 mole % Nal, and 4 mole % UF, are given in Table 34. The efquation 8095 10g P{mm Hg) = - ?zgfii 7). . ( )La L. Quill {ed.), The Chemistry and Metallurgy of Miscellanecus Matertials: Thermo- dynamics, McGraw-Hill, New York, 1956, 147 ANP PROJECT QUARTERLY PROGRESS REPORT best represents the data. The calcu- lated heat of vaporization 1s 37 kcal/mole. Table 34 VAPOR PRESSURE OF MIXTURE CONTAINING 530 mole % ZrF4, 46 mole % NaF, and 4 mole % UF, TEMPERATURE OBSERVED PRESSURE (°C) (mm Hg) 807 i2 842 21 876 31 907 45 936 72 940 84 Vapor pressure data for the solution containing 46 mole % ZrF,, 50 mole % NaF, 4 mole % UF, are presented in Table 35. The eguation 7630 log P(mm Hg) = - — + 7.988 T{°K) The heat of vaporization as calculated from the equation is 35 kcal/mole. An increase was obtained from the data, of 4 mole % ZrF, content results 1in a considerable increase in vapor pressure, It seems likely that, since ZrF, 1s by far the most volatile component of the mixtures being considered, the vapor pressure will, in general, increase with increasing ZrF, content, although probably not in a regular fashion. Table 35 VAPOR PRESSURE OF MIXTURE CONTAINING 46 mole & ZrF4, 50 mole % NaF, and 4 wole % UF, TEMPERBATURE OBSERVED PRESSURE (°C) (mm Hg) 807 8 858 17 8§72 22 906 33 148 CONVECTIVE HEAT TRANSFER IN MOLTEN SOPIUM HYDROXIDE H. W. Hoffman J. Lones Reactor Experimental Engineering Division Final results have been obtained from the turbulent forced-convection heat transfer experiment in which molten sodium hydroxide flowed through tube. The apparatus was in detail 1n previous re- a circular described ports.(®) The experimental data for the region of fully established flow are correlated by the equation Nu = 0.021Re® 8pyp0-* for a Reynolds modulus range of 6,000 to 12,000 and temperatures between 700 and 900°F, This equation is compared with the McAdams’' equation for the heating and cooling of ordinary fluids (fluids other than the liquid metals) in Fig, 64. The curve lies 9% below the McAdams’ correlation. (B)H. ¥. Hoffman and J. Lones, op. cit., ORNL-1294, p. 152, and Aircraft Nuclear Propulsion Project Quarterly Progress Report for Pertod Ending March 10, 1952, ORNL-1227, p, 161, UNCILASSIFIED 100~ — e g — DWG. 16361 m”__L;j%_mr_rmm{ e s80+———— A4 R - | — - McADAMS B a - THIS WORK } 60 e 1 - s - | = S 40F-—— - 44/ s j ;%V £ /gj g . fi ) - 20— | ‘ | 10, b | 2,000 10,000 30,000 REYNOLDS NUMBER Fig. 64. Experimental Heat Transfer Coefficients for Sodium Hydroxide. FOR PERIOD ENDING SEPTEMBER 160, Thermal entrance lengths were obtained for each run. This length is defined as the distance from the entrance of the test section, in terms of tube diameters, at which the local heat transfer coefficlent, or con- ductance, reaches a value within a given percentage of the established value. A 10% value was used here, Figure 65 presents the thermal entrance length as a function of the Peclet modulus (Beynolds modulus times Prandtl modulus), UNCI_ASS]IFIED 50, DWG, 15262 - I B b ek L L LA b 20 'L'""“’"“" ’C-:;fl) o | {Q frrmmmmrememem s d e L — - 9 i L { I e - 8 e e 38 wma .| 7 e s e e a oL e Bfmmree s e i ] o L e 0,000 100,000 PECLET MODULUS Fig. 5. Thermal Entrance Length for Molten Sodium Hydrexide. A new apparatus to be used to determine the convective conductances for the NaF-KF-LiF mixture is now be- ing assembled, This system is similar to the one used in the sodium hydroxide experiment, Owing to the corrosive nature of the NaF-K¥-LiF mixture, all system components except the test section are constructed of Incomnel. The test section will be made of nickel tubing because the proper size of Inconel tubing 1s not available. . temperature incremexls, 1952 BOTLING HEAT TRANSFER IN MERCURY W. 5. Farmer Reactor Experimental Engineering Division Further measurements have been obtained on free-convection hoiling of mercury from horizontal surfaces. The experimental heat transfer vs. temper- ature difference data are presented in Fig., 66. ¥From this plet the mean heat transfer coefficients shown in Fig. 67 were computed. The increase 1n pressure accompanying a change in boiling temperature from 350 to 550°F produced an approximately threefold increase in the heat transfer co- efficient for any given heat flux level., On the basis of these results, beiling and condensing mercury should prove to be a good medium for heat transfer in high-temperature power cycles, A large difference between the heat transfer coefficient data for the mercury-copper {(wetted)} and mercury-chromium (nonwetted) systems may be observed. There are insufficient data at this tiwe to establish accu- rately the specific interface~-fluid temperature difference at which boiling is initiated, as well as other boiling However, 1t boiling co- curve characteristics. does appear that the efficients rise rapidly at the higher as in the case of boiling water. NATURAL CONVECTION IN CONFINED SPACES WITH VOLUME HEAT GENERATION D. C. Hamilton F. E. Lynch Reactor Experimental Engineering Division Preliminary data obtained with the flat-plate apparatus described in 149 ANP PROJECT QUARTERLY PROGRESS REPORT 100,000 80,000 60,000 40,000 20,000 HEAT FLUX, @/4 (Btu/hr-{) 10,000 |- 8,000 6,000 4,000 3 4 6 8 10 UNGLASSIFIED CWG. 16363 COPPER SURFACE -——— BOILING TEMPERATURE 550°F COPPER SURFACE BOILING TEMPERATURE 350°F CHROMIUM SURFACE —— v BOLING TEMPERATURE 350°F 20 40 60 80 100 200 SURFACE~FLUID TEMPERATURE DIFFERENCE (°F) Fig. 66. Horizontal Surfaces to Mercury. previous reports‘®) appear to be in general agreement with the temperature profile predicted by the laminar flow analysis, (%) Repeated short circuits from the electrolyte through the delicate, quartz, traversing-thermo- couple assembly to the thermocouple have prevented obtaining other than preliminary data. A new technique for joining the assembly appears to have alleviated this problem. (Q)D. C. Hamilton, F. E. Lynch, L. Palmer, and R. F. Redmond, op. cit.,, ORNL-1227, p. 156; D. C. Hamilton and F. E. Lynch, op. c1it., ORNL- 1294, p. 158, (IO)D. C. Hamilton, H. F. Poppendiek, and L. D. Palmer, Theoretical and Experimental Analyses of Natural Convection Within Fluids in which Heat Is Being Generated, ORNL CF-51-12-70 (Dec. 18, 1951). 150 Experimental Data for Free-Convection Boiling Heat Tramsfer from The circular annulus apparatus has been constructed and assembled. The cooling systems for the upper electrode and the test section have been tested for leaks, and the system will be ready to operate when the flat-plate data have been obtained. This apparatus was designed to operate in the high- laminar and low-turbulent flow regions. The upper limit is set by the 10-kw maximum power now available. The validity of all previous data was 1n question because it was not possible to directly measure the wall tempera- ture. Since the temperature gradient is maximum at the wall, 1t is diffa- cult to determine the wall temperature from “near-the-wall” measurements. FOR PERIOD ENDING SEPTEMBER 10, 10,000 8,000 6,000 4,000 1952 UNGLASSIFIED OWG. 16364 40 60 80 100 200 SURFACE~FLUID TEMPERATURE DIFFERENCE (°F) 2,000 [ o ~, £ . 2 @ < 1,000 80O 600 e COPPER SURFACE BOILING TEMPERATURE 550°F 400 | ® COPPER SURFACE BOILING TEMPERATURE 350°F A CHROMIUM SURFACE _ BOILING TEMPERATURE 350°F 200 5 6 B 10 20 Fig. 67. The present apparatus consists es- sentially of a 1/2-in.-ID, thick- walled, copper tube that is electri- cally insulated on the inner surface by a baked enamel coating, In this design it is possible to install a thermocouple in the wall to measure the wall temperature, This represents Free-Convection Boiling Heat Transfer Coefficients for Mercury. a great improvement over previous designs. The cross.section view in Fig. 08 shows the method of installation of the center and wall thermocouples. The center thermocouple is located 1in a gquartz capillary tube. 151 ANP UNCLASSIFIED CWG. 16365 --UPPER ELECTRODE = LIQUID MERCURY LEVEL ——--COPPER TEST SECTION — COPPER COOLANT JACKET ~— COOLANT OUT < =i =M P T ) S : — = ’ SRR s o —--WALL THERMOCQUPLE ;g f—— ——----CENTER THERMOCOUPLE 4——QUARTZ TUSE 1%in e -------- CERAMIC COATING ——-=COOLANT IN 4~ LOWER =) ELECTRODE Fig. §68. Annulas Convection Apparatus. Type of Free- PROJECT QUARTERLY PROGRESS REPORT HEAT AND MOMENTUM TRANSFER ANALYSIS OF THE THERMAL CONVECTION LOOPS H. F. Poppendiek [.. Palmer Reactor Experimental Engineering Division Mean circulation velocities should be knowo i1n order to properly interpret corrosion results in thermal convection loops. At present, ANP convection harps are instrumented sothat approxi- mate wall-temperature distributions of the cold legs and power inputs to electrical heaters on the hot legs are measured. This information can be used to estimate some maxlimum and minimum circulating velocities that How - to obtain more exact velocities represent limiting situations, ever, it 1s necessary to solve the combined heat and momentum transfer problem numerically. Sowe incomplete numerical analyses indicate that the calculated circulation velocities fall within the estimated range close to the minimum value., Thus 1t appears that the circulating velocities 1n the con- vection harps are approximately 0,1 ft/sec when the fluid is NaF-KF-1iF (11.5-42.0-46.5 mole %). It should be possible to i1ncrease these velocities by heating the hot legs and cooling the cold legs 1n such a way that uni- form wall-temperature distributions exist in those legs. 13. RADIATION DAMAGE D. S. Billington, Solid State Division A, J. Miller, ANP Division The radiation damage studies of materials exposed in the ORNL graphite reactor, the LITR, and the 86-1in. cyclotron have continued., Further experiments have been carried out on the effects of 1rradiation on the fused fluoride fuels, the performance of liquid metal loops, the creep of metals, and the thermal conductivity of metals. 152 As 1n the case of the beryllium- bearing fuels previously reported, the evidence to date indicates that there 1s no significant radiation-induced corrosion with zirconium-bearing fuels undergoing fission at the rates that will occur during operation of the ARE. Apparatus is currently being assembled and inserted 1in the MTR to provide an exposure facility for the FOR fused fluoride fuels in which fission rates comparable to those that would occur in an aircraft reactor can be achieved. Measurements made in the graphite reactor on the escape of Xe!?5 froma capsule of molten fluoride fuel showed that only a small fraction of the xenon diffused out of the liquid. Examination of an Inconel loop through which sodium had been circu- lated at a peak temperature of 1500°F in the graphite reactor disclosed no positive evidence of radiation-induced corrosion. ' In-reactor cantilever creep measure- ments were made on Inconel specimens subjected to the 1700°F anneal proposed for the ARE fuel tubes. Under some conditions of stress and flux there was no increase 1n the creep rate due to the irradiation, such as was previously reported for specimens annealed at Jlower temperatures. Tt was also found that the thermal con- ductivity of Inconel given the ARE- specified heat treatment was un- affected by irradiation in the LITR. Additional details on radiation damage studies are contained 1in the quarterly reports of the Solid Stat Division. ' IRRARIATION OF FUSED MATERIALS G. W. Keilholtz D. F. Weeks J. G. Mergan M. T. Robinson H. E. BRobertson D. D. Davies C. C. Webster A. Richt P, R, Klein W. J. Sturm M. J. Feldman Solid State Division R, J. Jones R. L. Knight Electromagnetic Research Division B, W. Kinyon ORNL. Engineering Division Three samples of the zirconium- bearing reactor fuel, NaF-K¥F-ZrF,-UF, PERIOD ENDING SEPTEMBER 106, 1952 (4.,8-50,1-41.3-3.8 mole %), were irradiated at 1500°F for 140 hr in Inconel capsules in the LITR. Energy was dissipated in the fuel at a rate of 125 watts per cubic centimeter of fuel, which is 1n the range of the peak power dissipation expected during operation of the ARE. As with the beryllium-bearing fuels previously irradiated, chemical analyvses of the zirconium-bearing fuels and metal- Jographic examinations of the capsules showed no evidence of radiation- induced corrosion when compared with two out-of-reactor control runs. Bombardment of zircomium-bearing fuel in an Inconel capsule for 8 hr in the cyclotron with a beam of 22.5-Mev protons dissipating 2500 watts per cubic centimeter of fuel caused a in the corrosion of the Tncenel. This 1s similar to the result reported previously for a lithium-bearing fuel.,(?) slight increase An experiment was performed:to determine the rate of diffusion of Xe!3% from the fused flueride fuels under essentially stagnant or turbulent conditions, made 1in the OBRNL graphite reactor at 1300°F with NaF-BeF,-UF, (25-60-5 mole %) fuel containing natural uranium in Inconel capsules. In the first run xenon was flushed out by bubbling helium through the melt, whereas in the other the gas was simply swept over the surface. The xenon removed non- Two runs were in the sweeping experiment was eqguiva- lent to only 2.5% of that removed with the flushing procedure. Additional woerk 1n progress on irradiation of the fused fluorides in- cludes melting-point determinations of irradiated fuels, rocking-capsule tests to simulate the ARE cycling, a (1){;. ¥. Keilholtz, dircraft Nuclear Propulsion Projeet Quarterly Progress FReport for Period Ending June 10, 1952, ORNL-1294, p. 160, 153 ANP PROJECT QUARTERLY PROGRESS REPORT study of the relative importance of ionization effects in the proton beam, and longer cyclotron beam exposures, The MTR exposure facility for irradiation of fuel capsules at 2,500 to 10,000 watts per cubic centimeter of fuel has been completed. The 1initial experiment at 2500 watts 1s expected to be run duraing September, “knock-ons"” vs. IN-REACTOR CIRCULATING LOOPS 0. Sisman R. M. Carroll W. W, Parkinson C. D, Bauman J. B. Trice C. Ellis A. S. Olson W. E. Brundage M. T. Morgan D, T, James F. M. Blacksher Solid State Division Metallographic examination of the in-reactor and out-of-reactor portions of the 1500°F socdium loop described previously(?) showed no evidence of radiation-induced corrosion of the Inconel. A loop with i1mproved welds and a by-pass circuit for filtering the sodium 1is being constructed for additional experiments i1in the graphite reactor. Considerable progress has been made on fabrication of the stress- corrosion sodium loop for operation in the LITR. In addition, an in-reactor system for circulating fluoride fuel 1s being designed. CREEP UNDER IRRADIATION W, W. Davis J. C. J. C. Zukas Solid State Division Wilson Increasing the annealing tempera- ture from 1650 to 1700°F appears to (2) 163. 0. Sisman et al., op, cit., ORNL-1294, p. 154 reduce the adverse effect of neutron bombardment on the creep of Inconel at 1500°F and 1500 psi reported previ- ously. ®> The bench result and an in- reactor result after annealing at 1700°F are shown in the two lower curves in Fig. 69; the upper four curves, reported previously, () are shown for comparison. Although the difference in behavior is attributed to the heat treatment, there was a difference i1n the manner of annealing that could have been partly responsible. The 1650°F anneals were carried out in sttu and only the gage length was exposed to the desired temperature, so some long-time metallurgical changes in and beyond the fillets could have occurred. During the 1700°F anneals the full length of the specimen was heated in a large furnace, The possibility of some metallurgi- such as grain growth, increased creep rate following the 1650°F annecal unlikely in Inconel. The sluggishness of the recrystallization process 1in Inconel when only small amounts of cold reduction have preceded the anneal 1s well known. Although the metallurgical history of the plate from which these specimens were made cal change, causing the 1s not is unknown, hardness measurements indicate that it received about a 5% cold reduction following the penultimate anneal, Evidence of structural changes over long periods was found in the marked increase 1n intensity of all x-ray diffraction lines following a 1000-hr soak at 1500°F after the usual anneal of 2 hr at 1700°F, Diffraction patterns of the outer fibers of both compression and tension sides of all the Inconel bench test specimens have been made, More data are required for (35, €. Wilson, J. C. Zukas, and ¥. W. Davis, Solid State Division Quarterly Progress Report for Period Ending May 10, 1852, ORNL-1301 {in press). FOR PERIOD ENDING SEPTEMBER 10, 1952 SSD-4-472 DWG. 162754 Q.15 l ’ INDICATES REACTOR SHUTOOWN — 1 Q 3 TN EXTENSION (%} SPECIMENS A, B,C, AND D ANNEALED 2hr AT {650°F - SPECIMENS £ AND £ ANNEALED 2hr AT 1700°F 0.05 / | A, B, AND F ORNL GRAPHITE REACTOR TESTS AT A FAST FLUX oF 4x10'¢ n/em? - sec | C,D, AND £ BENCH TESTS C I e —— . F T 1 H o . 0 200 400 600 TIME (hr) Fig. 69. (Cantilever Creep Tests of Inconel at 1500°F and 1500 psi. a detailed analysis, but it has been noted in all cases that the (111) line is much more intense on the compression than on the tension side. No change in the lattice constant of Inconel was found within the accuracy of measure- ment, approximately 1 part in 3000, Strain-time curves for Inconel - annealed at 1700°F for 2 hr have been obtained in the graphite reactor and the LITR at 2000 psi and 1500°F., The data indicate that a number of ad- ditional experiments are necessary in order to fully explore the variables, 155 RADIATION EFFECTS ON THERMAL CONDUCTIVITY A. Foner Cohen L. C. Templeton Solid State Division Absolute thermal conductivity measurements were made on an ARE heat- treated specimen on Inconel by using a radial-heat-flow method at tempera- tures up to 1240°F before irradiation, at temperatures up to 1500°F during irradiation in hole HB-3 of the LITR, and at various temperatures immediately 156 following irradiation. The fast flux in the HB-3 hole is of the order of 10'? neutrons/cm?*sec. The thermal conductivity values calculated, so far, from the results of this experi- ment show that before and after irradiation the thermal conductivity values are the same within the accuracy of the measurements, which is of the order of a few per cent. The ‘during- irradiation’ data show no change in the conductivity as a function of time at constant temperature, and they agree with the out-of-reactor measure- ments, SUMMARY AND The analytical chemistry program in support of the materials research and equipment development programs included the routine analysis of 1179 individual samples, as well as the development of new analytical procedures (sec. 14). 1In particular, new methods have been developed for the quantitative determaination of the constituents and impurities in fluoride fuels. The INTROBUCTION work is concerned with the improvement and simplification of these analytical techniques. current The list of reports that have been issued by the project during the last quarter includes 17 formal reports and 56 informal documents (not including internal documents) on all phases of ANP research at OBNL {sec. 15). 14. C. D. ANALYTICAL CHEMISTRY Susano ‘Analytical Chemistry Division Progress has been made in improving and simplifiing the methods of analyses of reactor fuels. A volumetric method has been developed for the determination of alkali metals that depends upon separation of the elements by ion exchange and their indirect determi- nation by titration of an equivalent quantity of chloride ion. The in- vestigation of a volumetric method for the determination of zirconium in fuels has continued., Zirconium in fuels canbe determined by a volumetric method, 1n the absence of uranium, with an accuracy of about 1%. Tt appears that this determination can also be performed in the presence of uranium, and this phase of the problem is still being investigated. The use of silver peroxide as an oxidant in the determination of chromiuw by means of diphenylcarbazide has been studied, and a new and improved method, based on this study, has been adapted for the determination of chromium. Attempts to determine nickel i1n reactor fuels by means of nitroso salicylic acrd were not successful,. ' A method similar to: the Dean and Stark method¢!? has been developed for the determination of trace amounts of water 1n finely divided solzid materials. This method 1s particularly valuable for use with some fluorides with which other methods fail. The products of the reaction between NaK and reactor fuels have been studied. Tt was found that uranium{IV) is reduced to UF,, and at least one other constituent of the fuel, be zirconium{1V), which must 15 also reduced. A review of the service analyses performed during the presented. quarter 1is ANALYTICAL STUDIES OF COMPONENTS OF FLUGRIDE MIXTURES ‘ R. Rowan, Jr. €. M. J. C. White W. C. K. Talbott Analytical Chemistry Division Boyd J. Boss Alkali Metals. The emphasis on ternary mixtures of fluoride salts containing only a single alkali metal fluoride has greatly simplified the problem of the determination of alkali (D oy, Chen, Dean and D. D. Stark, J., Ind. 12, 486 (1920). ' Eng. 159 ANP PROJECT QUARTERLY PROGRESS REPCORT metals and has reduced the need for methods of separation of alkali metals from each other. An indirect method was, developed for the separation and determination of sodium and potassium that involved the titration of the chloride equivalent tothe alkali metal 1on, The separation 1s accomplished by means of 10n- exchange resins. Although the sepa- ration is time-consuming, by using several columns the time for each determination 1s far less than 1s required by the conventional techniques for this separation, however, Zirconium. A volumetric method for the determination of zirconium that makes use of the fluoride complex was suggested by Sawaya and Yamashita.(2) In essence, the procedure used in this study 1s to precipitate zirconium hydroxide by the addition of sodium hydroxide, adjust the pH of the slurry to 8.5, add a 15- to 20-fold excess of potassium fluoride, neutralize the potassium hydroxide formed with excess standard nitric acid solution, and titrate the excess acid to a pH of 7.5 with standard sodium hydroxide so- lution. The reactions involved are not stoichiemetric, and as a result the method must be standardized empirically. Data gathered thus far have shown that by standardization against a known quantity of zirconium 1t 1is possible to titrate the element with an error of less than 1%, under the conditions described, for a range of zirconium concentration of 15 to 60 mg in a volume of 200 ml. The error for larger amounts has not been determined with any degree of certainty, as yet. The additionof uranium to zirconium sulfate solutions results 1n some inter ference, since uranium forms a slightly soluble hydrous oxide. {Q)T. Sawaya and M. Yamashita, J. Chem. Soc. (Japen) 72, 414-16 (1951}); C. A. 46 1911 (1972}, 160 Compensation for this interference was made on the basis that 2.22 moles of sodium hydroxide(3?’ 1is consumed to precipitate each mole of uranium, The results on samples containing 3, 6, and 12 mg of uranium per 60 mg of zirconium agreed within 1% after the application of the necessary correction factor for uranium. The application of this method to sulfate solutions of reactor fuels showed that the results for zirconium by the volumetric procedure (with uranium correction) were generally 3 to 5% higher than those obtained gravimetri- cally by the phenylarsonic acid precipitation method, The study of the application of this method 1s continuing. ANALYTICAL STUDIES OF IMPURITIES IN FLUORIDE MIXTURES Chrominm, The use of silver peroxide¢*) as an oxidant for chromi- um(III) in the colorimetric determi- nation of chromium by diphenylcarbazide has been investigated, and a procedure in which silver nitrate and potassium persulfate solutions are employed in the oxidation step was adapted for the determination of chromium 1n reactor fuels. 'The precision of the method 1s equal to that obtained when any other oxidant i1s used, and the method is probably the most rapid of any currently 1n use. An additional feature is that the chromium(VI) formed 1is stable for at least 24 hr after oxidation, which 1s not true in the case of a number of the other oxidants commonly employed in this determination. This work is described elsewhere(5) in greater detail. (S)B. Kunin, The Potentiometric Titretion of Tuballyl Ion with Alkali, A-3294 (May B, 194°7). (4)M. Tanaka, Bull. Chem. Seec. Japan 23, 165~8 (19°0). (S)J. C. Yhite, C. M. Boyd, W. J. Ross, and C. K. Talbott, " Analytical Studies of Reactor Fuels and Their Components,” Analyticael Chemistry Division Quarterly Progress Report for Period Ending June 26, 1952, ORNL-1361 (in press). FOR Nickel. The commonly used di- methylglyoxime method, with certain modifications in procedure to adapt it to use with fluoride fuels, 1s employed for the determination of nickel. Because the quantity of nickel appearing as a corrosion product in the fuels currently being tested i1s generally smaller than has been the case in the past, the lower limit of determination of nickel has become more critical.. A practical limitation is imposed on the sample size used because of the presence of zirconium and the high acid concen- tration necessary to prevent its hydrolysis. The lower limit of de-~ termination of the present method is 10 ppm, and the use of absorption cells of 5~cm path length is necessary to achieve this figure, A brief investigation was made of a method for the determination of nickel that involves extraction of nickel at a pH of 5 to 6 with nitroso salicylic acid.{®) The obvious advantage of this method is that the color can be developed in slightly acidic solution instead of the strongly basic lution requiredin the dimethylglyoxime procedure. Unfortunately, the zirconium concentration is so high that the large quantity of complexing agent (oxalate, tartrate, or fluoride) retards the formation of the colored complex. It appears that the use of 5-cm absorption cells will be necessary S0 - in most cases to obtain satisfactory absorbancy measurements for nickel 1in the range of less than 50 ppm. 0xygen, The determination of oxygen in reactor fuels 1s a problem of considerable importance, since the presence of oxygen 1is generally con- sidered to be a contributing facter in the corrosionof container materials. Bromine trifluoride i1s known to react, (6)y. H. Perry and E. J. Serfass, Anal. Chen. 22, 565-67 (19%50). ' PERIOD ENDING SEPTEMBER 10, 1952 with the liberation of oxygen, with virtually all the oxides that can possibly be present in these reactor fuels (U0,, ZrO,, Fe,0;, Cr,0;, Ni0), and 1t 1s believed that this reaction can be made to serve as a basis for a determination of oxygen. An apparatus in which BrF, will be reacted with the fuel at elevated temperatures of the order of 325°C is being fabricated at the present time. The reaction bomb and the high- pressure system are constructed of nickel. The oxvgen liberated from the reaction will be determined by a simple gasometric measurement of the oxygen pressure 1n an evacuated system at a reduced temperature. This method appears favorable because of 1its simplicity and the relatively short time required for each determination. A pyrex unit to perform this operation has been fabricated and leak-tested, If unexpected difficulties arise, however, other methods for the de- termination of the oxygen evolved in the reaction are available. Chiloride. The zirconium tetra- fluoride used in the preparation of reactor fuels 1s prepared by the hydrofluorination of zirconium tetra- chleride. Since chlorine 1s an undesirable impurity in fuels because of 1ts rather high neutron-capture cross section, it 1s necessary to determine chloride 1in tetrafluoride to concentration 1s limits. zirconium assure that 1ts beleow tolerable Several attacks were made on this problem. A semiguaptitative test was devised in which a turbidimetric comparison was made between a hydro- fluoric acid, silver nitrate solution of the unknown sample of zirconium tetrafluoride, and standard quantities of chloride in similar solutions, As little as 0.01% of chloride imparted an easlly visible cloudiness to the 161 ANP PROJECT QUARTERLY PROGRESS REPORT solution. It 1is significant that none of the samples of zirconium tetra- fluoride tested so far have produced a visible cloudiness, which indicates a chloride content less than 0.01% for these samples. A second method of attack was to oxidize the chloride present to chlorine and determine it iodometri- cally. By using a sample of suitable size, 1t was possible to determine as little as 0.05% chlorine. Potassium permanganate, ceric sulfate, and lead dioxide were tested as oxidants, Permangante 1s generally unsatis- factory; 1t yields results averaging 20% haigh. This 1s believed to be caused principally by decomposition of permanganic acid during refluxing. Ceric sulfate gave low results 1in most cases, as did lead dioxide, which was only briefly tested. The semiquantitative test yields results of sufficient accuracy for the present purposes. Water. A method for the determi- nation ol trace amounts of water 1n fluorides, oxides, and other solids in a finely divided state has been developed., Such a method was princi- pally required for the analysis of the mixture to be used for cold critical experiments and 1ts components, Water 1s extracted from the solids by refluxing with xylene by an adaptation of the Dean-Stark method.(!) The xylene-water azerotrope is distilled into methyl alcohol and the water 1is titrated with Karl Fischer reagent. This method was adopted in preference to the direct titration of the solid with Karl Fischer reagent partly because of sample size limitations and partly because investigation showed that the completeness af the reaction was uncertain in the direct method. Tgnition methods for the determination of water were unsuccessful due to the hydrolysis of fluoride 162 salts that takes place at elevated temperatures. The extraction method has proved to be completely satisfactory. Samples of the order of 20 g can be handled, 1f necessary. The blank, although accounting for 50 to 75% of the total titration 1n some samples, has been reproducible to within less than 1%. The method has been tested against standard sodium tartrate, with the recovery of water averaging mnearly 90%. In addition to NakF, Zr0,, C, UF,, and KF, the method has been used successfully on fluorides such as NH,F, NH,HF,, NiF,, KHF,, and K,FeF,. DETEEMINATION OF CARBON IN ZrF, AND 270 ,-NaF-C MIXTURES In determining carbon in zirconium fluoride and i1n Zr0,-NaF-C mixtures by combustion, the fluorides interfere. This difficulty has been overcome by removing the fluorides by dis- solution. Zirconium fluoride 1s dissolved by oxalic acid and the alkali flvorides by warm water. The residue 1s collected 1n a swmall, gunartz, filtering crucible and the carbon content i1s determined by combustion. Since large samples may be used, the accuracy of the carbon determination 1s increased. The fluoride in the filtrate from the ZrO,-NaF-C mixturesis precipitated as lead chlorofluoride, dissolved, and titrated with mercuric nitrate. The fluoride and carbon values, which can be determined rapidly by the method outlined, give an indication of the degree of homogeneity attained in the preparation of large batches by mechanical mixing. COMPATIBILITY OF REACTOR FUELS AND Nak A number of experiments has been conducted to determine the reaction FOR PERIOD ENDING SEPTEMBER 10, 1952 products of reactor fuels (NaF~ZrF¢- UF,) and NaK under static and dynamic conditions., The conclusions drawn from the chemical analyses were: (1) NaK reduces U{IV) to UF,, which separates from the molten mass and settles to the bottom of the container; {2) the greater the concentration of NaK, the more complete the reduction; (3) a nearly linear relationship exists between the hydrogen evolved from the fuel upon acidification and the amount of NaK added to the fuel: (4) NaK, 1in large excess over that required to reduce quantitatively the UF, to UF,, reduces other components of the fuel, which form compounds that alse evolve hydrogen from acidic solutions; (5) less than 50% of the reducing power of the NaK added was recovered in any test, ' SERVICE ANALYSIS H. P. House L. J. Brady W. F. Vaughn Analytical Chemistry Division The analyses of zirconium fuels and of zirconium fluoride have received major attention during the quarter, Application was made of the methods adaptedby the research and development group for determining sulfide, sulfur, carbon oxide, water, and chloride in fused zirconium-bearing fluoride salts. In addition to zirconium samples, many samples of alkali fluorides 'and quaternary eutectics composed of alkali fluorides and uranium tetra- fluoride were analyzed. Unrelated miscellaneous samples were tested: a number of metals and alloys; fluorides of iron, nickel; 1nert gases; bonding cement. complex chromium, and ceramics; and 0Of a total of 1179 samples analyzed during this quarter, 80% was for the reactor chemistry group, 15% was for the experimental engineering group, and the remaining 5% was for the ANP critical experiments group, the electromagnetic research group, and the maintenance shops. A summary of the service analyses for ANP follows: Semples on hand May 10, 1952 198 Number of samples received 1192 Total number of samples 1390 Number of samples reported 1179 Backlog as of August 10, 1952 211 163 o ANP PROJECT QUARTERLY PROGRESS REPORT REPORT NO. CF-52-4-191 ORNL- 1234 Y-F12-7 Y-F15-10 ORNL~ 1241 Y-F17-16 Y-F17-17 Y-F17-18 Y-F17-20 Y-F17-23 Y-Fl0-101 Y-Fl0-102 Y-F10-103 GRNL- 1320 Y-FI10-106 Y-F10-109 164 15. LIST OF REPORTS ISSUED TITLE OF BREPORT AUTHOR(S) I. GENERAL DESIGN Hydrodynamics of Homogenecous Reactors G. F. Wislicenus (Presented at Second Fluid Fuels Con- ference, Osk Ridge, Tenn., April 1952) Reactor Program of the Aireraft Nuclszar Propulsion Project Proposed ARE Design Three Reactor~Heat Exchanger~Shield Arrangements for Use with Fused Fluoride Circulating Fuel I1X. EXPERIMENTAL ENGINEERING Fuel Flow Studies in ARE Fuel System Mockup Study of Fluid Flow Measuring Devices Moore Nullmatic Pressure Transmitter Test Line Plug Test No. 1 - Fuel at 1500°F Tank Car Helium as a High Purity Helium Scource Production of Fluorides in Buildings 9211 and 9201-3 III. REACTOR PHYSICS Values of Resonance Integrals An QOutline of the General Methods of Re- action Analyses Used by the ANP Physics Group Statics of the ARE Reactor, Summary Report The Effect of Gaps on Pile Reactivity Reacteor Theory Terms Note on the Non-Linear Kinetics of Circulating-Fuel Reactors ™ B. Cottrell F. Haines . P. Fraas H. Ward A. Anderson W. Taylor W. Taylor A. Maon W. Savage S, Wilson B. Mills B. Mills Tamar B. Mills B. Arfken Tamor DATE ISSUED 4-30-52 6=-2-52 6-10-52 6-30-52 1-9-52 71-7=-52 71-10-52 1-10-52 B-4-52 8-6-52 4~30-52 5-16-52 5-8-52 8-14-52 7-16=-52 8-15-52 REPORT NO. ORNL- 1304 Ck-52-6~99 ORNL-~1142 OBNL- 1312 Y-895 Y-B23-2 Y-B23-5 Y-B23-7 Y-F10-108 Y-B23-9 CF-52~3-125 CF-52-3-146 CF-~52-4-99 ORNL- 1147 CF-52-6~54 FOR PERIOD ENDING SEPTEMBER 10, 1952 TITLE OF REPORT IV. NUCLEAR MEASUREMENTS Thermal Neutron Capture Cross Sections of Isotopes Energy Absorption of Capture Gammas Multiple-Crystal Gamma-Ray Spectrometer The Fission Cross Section Qf Uranium- 234 and Uranium~ 236 for Incident Neutron Energies up to 4 Mev : Crosa Sections for Carbon and Water in the Energy Range from 2.3 Mev to .025 ev. A Literature Search V. CRITICAL EXPERIMENTS Preliminary Direct Cycle Reactor Assembly- Part 11 Preliminary Direct Cycle Reactor Assembly- Part Y111 Preliminary Direct Cycle Reactor Assembly- Part [V The ARE Critical Experiment Loading of ARE Critical Experiment Fuel Tubes VI. SHIELDING OBNL “Tower Shielding Facility” Preliminary Proposal Health Physics Instruments Recommended for ASTF Some Ground Scattering Experiments Performed at the Bulk Shielding Facility The Unit Shield Experiments ‘at the Bulk Shielding Facility Data on ¥fnriched Fuel E]ements for the Bulk Shielding Facility AUTHOR(S) H. Pomerance H. L. F. Enlund F. C. Maienschein R. W. Lamphere Frances Sachs A. D. Callihan A, D. Callihan A. D. Callihan and assoclates C. B. Mills ‘D, Scett Dunlap Scott T. H. J. Burnette H. E. Hungerford . Meem H. E. Hungerford Cunningh am DATE ISSUED 6-6~52 6-11-52 7=3=52 7-15-52 8-4-52 5-21-52 6~-18-52 5-30-52 B-8~52 8-12-52 3-14-52 3-20-52 4-16-52 5= 14-52 6~10-52 165 ANP PROJECT QUARTERLY PROGRESS REPORT REPORT NO. CF-52-7-150 CF-52-8-38 Ck-52-6-145 ORNL.-1283 ORNL=-1273 CF-52-6-158 ORNL~ 1133 Part 2 CF-52-6-T4 CF-52-6-158 CF-52-6-165 CF-52-7T-1 CF-52-7-37 CF-52-7-71 CF-52-7-83 CF-52-5-40 CF-52-5-41 CF-52-5-163 CF-52-5-183 166 TITLE OF REPORT Uranium Penetration of Graphite Gamma Ray Spectral Measurements with Divided Shield Mockup. Part TII Basic Research in Shielding General Principles of a Proton Recoil-Fast Neutron Spectrometer Background Calcul ations for the Proposed Tower Shielding Facility Neutron and Gamma Dose Distribution in Water Surrounding the GE Outlet Air Duct The Shielding of Mobile Beactors-II Re: Design of Inexpensive Shield for Small Reactor Neutron and Gamma Dose Distribution in Water Phenomenological Theory of the Attenuation of Neutrons by Air Ducts in Shields Preliminary Data on GE Qutlet Air Duct Air Scattering Experiment at the Bulk Shielding Facility Gamma Ray Spectral Measurements with the Divided Shield Mockup, Part II Gamma Measurements on the GE Qutlet Air Duct as a Function of Length of Transition Section Gamma Dose Behind Iron-Water Thermal Shield of Various Thicknesses as a Function of Water Reflector Thickness. for NDA) {Investigation Gamma Dose Behind Iron-Borated Water Thermal Shield with 18-, 20-, and 30-cm Water Reflector (zamma Attenuation in [.ead-Water Shields Neutron and Gamma [Mstribution in Water Containing Lead, Steel and Air AUTHOR{(S) . K. Hullings C. Maienschein . P. Blizard R. Gossick Simon H. Ritchie E. Clifford I.. Storrs, Jr. . P. Blizard A. Wel ton Ralph Balent C. E. Clifford Simon E. Clifford E. Clifford — T . Meem E. Hungerford C. Maienschein E. Clifford E. Clifford E. Clifford E. Clifford E. Clifford DATE ISSUED 7-31-52 8-8-52 6-25+52 8-14-52 9/1952 6-26-52 6-30-52 6=-26-52 6-25-52 7-1-52 71-8-52 7-8-52 7-16-52 5-7-52 5=T7-52 5-20-52 5-26-52 REPORT NO, CF-52=-5-1 Part 8 CF-52-6~120 Part 1 CFa52-6-120 CF-52-8-166 ORNL~ 1370 CF-~52-6-148 CF-52-7-138 ORNE-1291 CF-52-6-76 ORNL-1163 CF-52-6-~127 ORNL.- 1286 Y-B31-354 FOR PERIOD ENDING SEPTEMBER 10, 1952 TITLE OF REPORT Nentron and Gamma Dose Distribution Beyond Beryllium Slab {28.5 cm, Density 1.23 gm/ec) Sources of Radiation, Chapter 2.1 of Reactor Handbook : Geometry, Chapter 2.7 of Reacitor Handbook VII. A Review of the Literature on Heat Transfer in Noncircular Ducts and Annuli for Ordinary Fluids and Liquid Metals Turbulent Forced Convection Heat Transfer in Circular Tubes Containing Molten Sodium Hydroxide Liquid Metals Handbook, Chapter 6, ‘Liquid Metal Heat Transfer” Viscosity of Fulinak Viscosity of Fuel Salt Mixtures No. 27 and No. 30 VITI. MATERIALS CHEMISTRY General Information Concerning Hydroxides Complex Fluoride Fuel Studies - X-ray The Preparation of Thin and of Thick Targets to b: Bombarded by Positive Particles Petrographic Examination of Fluoride Fuels IX. ANALYTICAL CHEMISTRY The Determination of Oxygen in Sodium Analytical Chemistry - ANP Program Quarterly Progress Report for Period Ending May 26, 1952 AUTHOR(S) C. E. Clifford E. P. Blizard F. C. Maienschein HEAT TRANSFER AND PHYSICAL PROPERTIES H. C. H., W R, N H. F B. F R. F Mary Paul R. A T. N J. C. w. J R. Rowan, M. T. Blizard Claiborne . Hoffman . Lyon . Poppendiek . Redmond: Redmond E. Lee Agron Bolomey . McVay White Ross Jr. Kelley DATE ISSUED f-6=52 6-23~52 6+ 2452 to be i1ssued to be 1ssued 9-1-52 6~19-52 7-23-52 4+21-52 6~ 16-52 6~3-52 6~20-52 4-30-52 5~26-52 167 REPORT NO. OBRNL- 1307 MM 2 MM=-6 MM-T7 Cr-52-7-124 CF-52-3~123 Y-872 OBRNL- 1114 Y-889 OBRNL-1294 Y-F26-40 168 TITILE OF REPORT AUTHOR(S) L. McCutcheon D. Susano Effect of Fluoride on the Gravimetric R. Determination of Zirconium C. X. METALLURGY Cone Arc Welding P. Patriarca MetalJographic Examination of 347 Stainless E, E, Hoffman Steel Heat Exchanger from Forced Convection l.oop which Operated for 3,000 Hours in NaK Metallographic Examination of Nicrobrazed E. E. Hoffman Stainless Steel Sodium-Air Heat Exchanger Following Failure During Test Beta Treatment of Alpha Uranium R. J. Gray A Simplified Apparatus for Making Thermal A. DeS. Brasunas Gradient Dynamic Corrosion Tests (See-Saw Tests) Selected Physical Properties of Lead in the Frances Sachs Temperature Range 100-1000°C, Search A Literature Stress+Strain-Time Phenomena in Mechanical A. G. H. Andersen Testing Selected Physical Properties of Mercury in Frances Sachs the Temperature Range 100-1000°C. A literature Search, XI. MISCELLANEOUS Aircraft Nuclear Propulsion Project Quarterly W, B. Cottrell Progress Report for Period Ending June 10, 1552 W. B. Cottrell ANP Information Meeting of August 20, 1952 DATE ISSUED 6-17-52 7-15-52 7-16-52 T-24-52 3-14-52 5-7-52 5-9-52 7-10-52 - 8-5-52 8-29-52