TR i o \ L'-r-.-'l.'d ! . ' -. . a “'ORNL-2012, Parf 1,11, 11’ C-B4 - Reoctors-Special Features of Aircroft Reactors AEC RESEARCH AND DEVELOPMENT REPORT oy. 129A * | ol T =3 . \ ey i 3 445k 0349900 & &;i 4 S Crs c&; I &7 3 * | d‘fi! AIRCRAFT NUCLEAR PROPULSION PROJECT C—_i’« Tfl QUARTERLY PROGRESS REPORT IE ‘3."3‘{ FOR PERIOD ENDING DECEMBER 10, 1955 W 3 CrassrrIcATION (1 Py Al'l'!fl‘lé B -"flt OAK RIDGE NATIONAL LABORATORY OPERATED BY UNION CARBIDE NUCLEAR COMPANY A Division of Union Carbide Lund Carbon Corporation i b N POST OFFICE BOX P * OAK RIDGE, TENNESSEE P ERSE RS R N - MIHS E'E iri i § tl its contents n, in any L ORNL-2012, Part I, II, Il} C-84 ~ Reactors-Special Features of Aircraft Reactors This document consists of 244 pages. Copy /.,2701‘ 324 copies. Series A. Contract No, W-7405-eng-26 AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT For Period Ending December 10, 1955 W. H. Jordan, Director S. J. Cromer, Co-Director A. J. Miller, Assistant Director A. W. Savolainen, Editor DATE ISSUED FER 20 1958 OAK RIDGE MATIONAL LABORATORY Operated by UNION CARBIDE NUCLEAR COMPANY A Division of Union Carbide and Carbon Corporation Post Office Box P Oak Ridge, Tennessee mic ts cantents MARTIN MARIETTA ENERGY SYSTEMS LIBRARIE N 3 4456 03494900 b NS R e b aaa £ ol UL - -.a—-a-u_q_. in any mannef" R. G, Affel 2 R. Baldock 3. Borton 4. D, S;Wgillington 5. F. F. BWNgkenship 6. E.P. Blizdg 7. C. J. Borkow 8. G. E. Boyd 9. M. A, Bredig 10. F. R, Bruce 11. A. D, Callihan 12.+ D, W, Cardwell 13. J. V. Cathcart 14. C. E. Center (K-25) 15. R. A. Charpie 16. G, H. Clewett 17. C. E. Clifford 18. J. H. Coobs 19. W. B, Cottrell 20-21. D. D, Cowen 22. S. Cromer 23. R. S, Crouse 24. F. L. Culler 25. J. H. DeVan 26. D. A, Douglas 27. E. R, Dytko 28. L. B. Emlet (K-25) 29. M. J. Feldman 30. D. E. Ferguson 31. A.P. Fraas 32. J. H. Frye 33. W. T, Furgerson 34. H. C, Gray 35. W, R. Grimes 36. E. E. Hoffman 37. A. Hollaende 38. A. S, Housgffolder 39. J. T.Ho 40. H. K kson 41. W. HBordan 42. G. Y. Keilholtz 43, P, Keim 44. M. T. Kelley 45, F. Kertesz 46. E. M. King 47-48. J. A. Lane w ORNL-2012, Part I, 11, 1l C-84 - Reactors-Special Features of Aircraft Reactors INTERNAL DISTRIBUTION o 64. - .94, 49. R, S, Livingston 50. R.N. Lyon 51, F. C, Maienschein 52. W. D, Manly 53. E. R. Mann 54. L. A, Mann 55. W. B. McDongjd 56, F.R., McQug i 57. R. V. M rebllan 58. R. P. Mfliford 59. A. Jg lller 60. RgE. Moore o1, j _ Z, Morgan 6g¥ E. J. Murphy 3. J. P. Murray (Y-12) G. J. Nessle 65. R. B, Oliver 66. L. G. Overholser 67. P. Patriarca 68. R. W, Peelle 69. A. M. Perry 70. J. C, Pigg 71. W. G, Piper 72. H, F. Poppendiek 73. P. M. Reyling 74. H. W, Savage A. W, Savolainen RS, R. D. Schultheiss W E.D. Shipley 78. %. Simon 79. OQSisman 80, G. Wy Smith 81. A. H,%pell 82. C. D, S&ano 83. J. A. SwaRgut 84. E.H. Taylo 85. R, E. Thoma 86. D. B, Trauger % 87. E.R. VanArtsdaly 88. F. C. VonderLage ¥ 89. G. M. Watson 90. A. M. Weinberg 91. J. C, White 92. G, D. Whitman 93. E. P. Wigner (consultant) '@, C. Williams 95. J. C. Wilson 96. C. E. Winters *1*** ORNL-2012, Part I, I, Il C-84 — Reactors-Special Features of Aircraft Reactors 107-126. Laboratory Records Department 127. Laboratory Recorgs, ORNL R.C, 97-1084 ORNL —Y-12 Technical Library, 128-130. Central Researchii® ibrary Document Reference Section 31. ] 134 135. 136. 137. 138. 139. 140. 141. 142-144. 145. 146. 147. 148-153. 154, 155. 156. 157. 158. 159. 160. 161-162. 163. 164. 165. 166. 167-169. 170-172. 173. 174-177 17§ 1. 0. 81. 82. 183. 184. 185. 186. EXTERNAL DISTRIBUTION AF Plant Representative, Baltimore 2. AF Plant Representative, Burbank . AF Plant Representative, Marietta AF Plant Representative, Santa Monica F Plant Representative, Seattle Plant Representative, Wood-RidgedFf Aflateriel Area ; Air Y search and Development Copgfiiand (RDGN) Air Rqearch and Development Cglimand (RDZPA) Air TeMigical Intelligence Cenj Aircraft Mlboratory Design Boich (WADC) ANP ProjeN@ Office, Fort Wgl Argonne Naffinal Laboratg Armed ForcesNpecial Wellpons Project, Sandia Assistant SecreWixy offfle Air Force, R&D Atomic Energy Coamifsion, Washington Battelle Memorial JMtitute Bettis Plant ‘ Bureau of Aerongltics Bureau of Aerqfifiutics (ede 24) Bureau of Aeglifautics Geral Representative Chicago Opgiiitions Office\g Chicago PgiBnt Group Chief of iiifval Research Convairgieneral Dynamics Corpiation Directglfof Laboratories (WCL) \ Dire of Requirements (AFDRQ& Dirgfifor of Research and Developmefl (AFDRD-ANP) Digtorate of Systems Management (A7 -15N) e ctorate of Systems Management (R(¥ge 1SS) quipment Laboratory (WADC) eneral Electric Company (ANPD) Hartford Area Office _ Headquarters, Air Force Special Weapons Cerfg idaho Operations Office Knolls Atomic Power Laboratory Lockland Area Office Los Alamos Scientific Laboratory Materials Laboratory Plans Office (WADC) Mound Laboratory \ National Advisory Committee for Aeronautics, Clevelan¥ 187. 188. 189. 190. 191. 192. 193-195. 196-199. 200. 201. 202. 203, 204. 205. 206-208. 209-323, 324. - ORNL-2012, Part |, Il, 1] C-84 = Reactors-Special Features of Aircraft Reactors National Advisory Committee for Aeronautics, Washington Naval Air Development Center New York Operations Office North American Aviation, Inc, (Aerophysics Division) Nuclear Development Corporation Patent Branch, Washington Powerplant Laboratory (WADC) Pratt & Whitney Aircraft Division (Fox Project) San Francisco Operations Office Sandia Corporation School of Aviation Medicine Sylvania Electric Products, Inc. USAF Project RAND University of California Radiation Laboratory, Livermore Wright Air Development Center (WCOSI-3) Technical Information Extension, Oak Ridge Division of Research and Development, AEC, ORO - Reports previously issued in this series are as follows: ORNL-528 ORNL-629 ORNL-768 ORNL.-858 ORNL-219 ANP-60 ANP-65 ORNL-1154 ORNL-1170 ORNL-1227 ORNL-1294 ORNL-1375 ORNL-1439 ORNL-1515 ORNL-1556 ORNL-1609 ORNL- 1649 ORNL-1692 ORNL-1729 ORNL-1771 ORNL-1816 ORNL-1864 ORNL-1896 ORNL.-1947 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 Period Ending June 10, 1952 Period Ending September 10, 1952 Period Ending December 10, 1952 Period Ending March 10, 1953 Period Ending June 10, 1953 Period Ending September 10, 1953 Period Ending December 10, 1953 Period Ending March 10, 1954 Period Ending June 10, 1954 Period Ending September 10, 1954 Period Ending December 10, 1954 Period Ending March 10, 1955 Period Ending June 10, 1955 Period Ending September 10, 1955 FOREWORD This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL records the technical progress of the research on circulating-fuel reactors and other ANP research ot the Laboratory under its Contract W.7405-eng-26. The report is divided into three major parts: |. Reactor Theory, Component Development, and Construction, H. Materials Research, and lll. Shielding Research, The ANP Project is comprised of about 530 technical and scientific personnel en- gaged in many phases of research directed toward the achievement of nuclear propulsion of aircraft. A considerable portion of this research is performed in support of the work 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 design, construction, and operation of the Aircraft Reactor Test (ART), with the cooperation of the Pratt & Whitney Aircraft Division, are the specific objectives of the project, The ART is to be a power plant system that will include a 60-Mw circulating- fuel reflector-moderated reactor and adequate means for heat disposal. Operation of the system will be for the purpose of determining the feasibility, and the problems associated with the design, construction, and operation, of a high-power, circulating-fuel, reflector- moderated aircraft reactor system, vii FOREWORD SUMMARY 1. ............................................... ................................................... ......................................................................................................... ......................................................................................................... PART I. REACTOR THEORY, COMPONENT DEVELOPMENT, AND CONSTRUCTION REFLECTOR-MODERATED REACTOR ..ottt ne e s rs s ART Facility Design and Construction ... st s s scne et et e enens ART DESIGN civirirriiieiirereee ittt cseste s s sa e st s bt st a e e s st e st s resaesbesaeasesesaanbastsresseneenaneaseesbantn e saentnsnenresesne s System Flow Sheets .ocevvvevvirenens ......................................................................................................... Fuel-to-NaK Heat Exchanger ..ot et et he s enens Fuel Filleand-Drain System s seee ettt abesae s st e a e s e sae e s s aeneasnnsreens ReaCtor Shield ..o e Core Flow Studies............. et are e --------------------------------------------------------------------------------------------------------- Engineering Test Unit oottt ettt e s e e st eb e e ae et et es e s sasasteesemesseeaeeenenee CoNtr OIS AN IS trUMEM O i O e coiiiieeerii s eiesceeereaesssrresarsnnsteeneesnarennnsssasesssstasasasssrssresansnsrasessssssesnsssesssessessenees Procurement of Special Reactor Materials and Components .....ccocceeverieivceiiieciecieee e, Beryllium oo Shell Fabrication .coveecerveeciireeeennene. CX-900 lnconel .ccoceoveeerccerreeeeeeeeene, ......................................................................................................... ......................................................................................................... ......................................................................................................... Main Heat Exchangers and Radiators ..ottt e Operation of ZrF , Vapor Traps in the High-Temperature Critical Experiment ...cocovvveviviiiiinnes, EXPERIMENTAL REACTOR ENGINEERING ...oooioioe et esver st e sve e sas e eb e In-Pile Loop Development and Tests ..ottt et et e st ene st esene e eanese s oo [MeP ile LLoop OPEration v eoccoeeeiieteieeveee ettt ee st eeae et et s et ee s e bt s s ab s bt ebes e eaetensenserentesessnsenes L 00D P UrGe SYSTEM ittt ettt e bbb bbb bbbk eb bt n b b s b bt a b e b ne b Forced-Circulation Corrosion and Mass Transfer Tests v eessieeeceeeeereerseeeereeeeteeeseeaeeeesaresnens Fused-Salt—lncone] Sy stems ..ottt ettt en et Liquid Metals in Multimetal Loops ........................................................................................................ P UMD DEVEIOPMENT oueitiiiiiiiriiie ettt s e et snae s et et sabs e s era e ese e ae e ses et e e e ar e seannrsnnens Bearing=and-Seal T ests «i ittt e e e et a s s raaeeare s Sodium-Pump Water Performance Tests ...t irnnenie e neee e sere s s sane e ses e rnesee s High-Temperature Tests of ART MF-2 Fuel Pump with NaK.......ccooi s High-Temperature Pump-Performan 8 TSt SHANAS tooreriiee et e et eeese e e e te e ae e e e e erraneaass Heat Exchanger Development ...ttt e s Intermediate Heat Exchanger Tests oottt st st ane e e Small Heat Exchanger Tests ittt et e s e s en Heat-Transfer and Pressure-Drop Correlations ..o U CEUIA] T @S S ettt ettt i e ete e e eiat s e tbesesarresabeaesabbaesbe e sb b s nesnsesn e ab b e s e eR b e e e an s e rabeesesararnsnrerrnneeeabreannnrs Outer-Core-Shell Thermal-Stability Test............ SO PP U OO U S PRTPI RO Inconel Strain-Cycling Tests ....... .......................................................................................................... Thermal-Cycling Test of Sodium-lnconel-Beryllium System .., Reactor Component Development ... Dump Valve crciiie e Cold Trap and Plugging Indicator .......................................................................................................... .......................................................................................................... .......................................................................................................... vii 15 15 19 19 19 20 20 23 25 25 25 25 26 26 26 26 27 27 27 29 30 30 32 33 33 35 36 41 4] 41 49 49 54 54 56 58 59 59 60 Zirconium FIuoride VAPOr TIAP v ste e s es e sasseesnes b ene st e snesbe e e ssassneesssssensesanns Water Test of Aluminum Mockup of Top of ART ..ot e b s 3. CRITICAL EXPERIMENTS ............... ........................................................................................................ Room-T emperature Reflector-Moderated-Reactor Critical Experiments....c.c.ccecermveninnrcennreneninians Power and Neutron-Flux Distributions ...t s sae e anens Neutron Production in the Fuel-to-NaK Heat Exchanger .......ccccoccoininnccveiiieeeieeenieans Radial Importance of Uranium in the Fuel Annulus. ..o e see e, Importance of Beryllium at End of Reactor ....cvecveieiceiiniicrrce et see e Axial Importance of @ Neutron SOUFCE ..ottt High-Temperature Reflector-Moderated-Reactor Critical Experiments ...ccccoccevrnnvinviniinecinininnninee. Compact-Core Reflector-Moderated-Reactor Critical Experiments ... PART Hl. MATERIALS RESEARCH 4. CHEMISTRY OF REACTOR MATERIALS cc..coiici ittt et Phase Equilibrium Studies.....c............ The System ZrF -UF The System LiF-UF , oo, The System NaF-LiF-UF ... The System KF-UF4 .......................... The System NaF-KF-Z¢F,................ The System NaF-LiF-BeF,........... The System NaF-LiF-BeF ,-UF , ... The System KF-BeF ..o, The System NcF-KF-BeF2 .............. Systems Containing Alkaline-Earth Chemical Reactions in Molten Salts.. ........................................................................................................ ........................................................................................................ ........................................................................................................ ........................................................................................................ ........................................................................................................ - ........................................................................................................ ....................................................................................................... -------------------------------------------------------------------------------------------------------- -------------------------------------------------------------------------------------------------------- ........................................................................................................ FlUOr IS eoeeeeeee oot e e et st st e v e e s s sa s bmns ........................................................................................................ Equilibrium Reduction of FeF, by H, in NaZrF ¢ e Reduction of UF4 By Structural Metals (it e st et Stability and Solubility of Chromium Fluorides in Various Molten Fluorides ..occoveeiicvececcvernnccennne. Reaction of UF3 with Alkali Fluorides ..ciiiiiienniiiciiceeiee Eeereereeesneanr et rrresb e aasenener naaaaesaraen Reduction of Alkali Fluorides by Uranium Metal......ccocoooiiiiiiie e e Experimental Preparation of Pure FIuorides .....ccciceoiiiiiiiieceees e Production of Purified Fluorides ...... ........................................................................................................ Removal of CrF, from Nc:F-ZrF“-UF4 MEXEUEES ettt e eeee e e et e m e et ene e e seneeeeneeen Laboratory-Scale Purification Operations .........ueiiririeininienc et ene e seses e e sn e as Special Preparation of NaF-ZrF ,-UF J-UF | s Evaluation of Raw Materials for Fuel Preparation ..o et Pilot-Scale Purification Operations Production-Scale Operations............ ........................................................................................................ -------------------------------------------------------------------------------------------------------- Batching and Dispensing Operations. ..ot rers s ene et e st Special Services ...oovvvceveiiceeneiiennns ........................................................................................................ Fundamental Chemistry of Fused Salts ...ttt e Relative Viscosity-Composition Studies of Nch-LiF-ZrF4 Mixtures at 600°C...........covrniviieneee. EMF Measurements in Fused Salts -------------------------------------------------------------------------------------------------------- 60 64 66 66 66 66 67 67 70 71 73 F o lUOPIAE S 1n TGO e oeee oot e et ee et e e e e e e e e e et v ereerere e e aa s aesaaaseasasaassaasssssasssnsesnsassaessenasnsesnsasessnenses 103 SOdTUM N TNCONEL vt et e e s sbesne e s s st ea st st e e e e naasresnnesteabeesrensteasens 106 Thermal-Convection STUdIies . oottt ettt s e s b e sas s e ebe s e nen e aesans 107 Effect of Difference Between Loop Wall Temperature and Fluid Temperature ......c.oooveeeerecennnnne. 107 Effect of Zirconium Hydride Additions to Fluoride Mixture......coooimeiiiiiiincin et 107 Loops Fabricated from Special Inconel-Type Alloys ..ccoveriiecimree e 110 Nickel-Molybdenum Alloy Loops ittt r et eb et e es bbb er e ees 111 Effect of Nitrogen Atmosphere ...ttt st e 111 General Corrosion STUAIES oo e s et s et e st bbese s s e eanesaesreesaesrneesbensesanabensessans 112 Thermal-Convection Loop Tests of Brazed Inconel T-Joints in NaF-ZrF -UF, s 112 Seesaw Tests of Brazed Inconel T-Joints in Sodium and in Fuel Mixtures.....cccoeeivvivccricniiniennnn., 113 Static Tests of Brazed Materials ..ottt e se st e abs et ene st 117 Static Tests of Brazed Hastelloy B T-Joints in Sodium and in NaF-ZrF ,-UF . 120 Seesaw Tests of Chromel-Alumel Thermocouple Joins to Inconel Thermocouple Wells ................ 120 Effect of Ruthenium on Physical Properties of Inconel oo, 122 Boiling Sodium in INConel Loop .ottt sasa e r et e s 122 Decarburization of Mild Steel by Sodium....cciiieii s e e e 123 Static Tests of Special Stellite Heats in Lithium it 124 Solubility of Lithium in NaK ... s e 125 Seesaw Tests of Titanium Carbide Cermets in NaF-ZrF -UF oo, 127 Thermal-Cycling Tests of Inconel Valve Disks and Seats Flame-Plated with a Mixture of Tungsten Carbide and Cobalt ..o 130 Static Tests in NaF-ZrF -UF , of Kentanium Cermet Valve Parts Brazed to Inconel ..........c..c...... 131 Effect of an Air Leak Into an Inconel~Fused-Salt Test System ..cccocevvvveiiiiiceicci e 131 Fundamental Corrosion Research ..ottt sa b s 133 Seif-Decomposition of Fused Hydroxides.......ococcvueiciniiiiiciinec e e 133 Mass Transfer and Corrosion of Various Materials in Fused.Sodium Hydroxide .......ccccocevninrinenn, 133 Chemical Studies of Corrosion ...ttt st st s bbb asteaansaeeris 138 Reaction of Inconel with Sodium and NaK .. 138 Reaction of Sodium Hydroxide with Nickel ..o 139 Study of Eutectic Mixtures by Zone Melting........c.ocooii ittt e 139 METALLURGY AND CERAMICS ..ottt ettt et et bbb ettt enes s 141 Fabrication of Test Components ..ottt e 141 NaK-to-Air High-Conductivity-Fin Radiators ....ccocoeiiiiniin e 141 Twenty-Tube Fuel-to-NaK Heat Exchangers ... 143 Intermediate Heat Exchanger No. 3 ...t st ennes 143 Intermediate Heat Exchanger Job-Sample Evaluations ..o 144 Examination of NaK-to-Air Radiators That Failed in Service ..o, 145 Brazing-Alloy Development ... 149 Mechanical-Property Studies of Nickel-Molybdenum Alloys ..o 152 Influence of Aging Heat Treatments on the Creep Properties of Hastelloy B.......ccoccoeninieii, 152 Preliminary Investigation of Creep Properties of Hastelloy W ..o 153 Investigation of the Creep Properties of Some New Nickel-Molybdenum Alloys.......oc 153 Special Materials Studies...coccovrriiiiiiience e rr et e e et b et e e et b eba st e rene s 155 Extrusion of Seamless Duplex Tubing ... 155 Neutron Shielding Material for ART .o 162 D ORE - N T O UMM et eeee ettt ettt e eteseseses e atarareete s sasasaet s e sasasas sbstssbnns rsnrssnaransnsransnaransnrsrnrns 163 Gamma-Ray Shield Material for ART PuUmps . vttt e 163 X1 xii Contr Ol R od FabriCation o ieeieieieeecceeteiessetetasnnesseensenntsnsssasnensaassrsaseertrenssssnnnnen NoNdestructive TeSHING .ovvcveiicieiiiiieieieie ettt e ent bbb aae s b Ceramic Research ..ot e e s sane s Rare-Earth Cermet Fabrication............. et e re e areirreiebeeseataerseree s aennereeseeantarteearees Boron Carbide Shield Material ..ot e cre e s se e e HEAT TRANSFER AND PHYSICAL PROPERTIES ..o Fused-Salt Heat Transfer.... e s ese e e e ses e eeas ART Fuel-to-NaK Heat Exchanger ...t ART Core Hydrodynamics ...ecciiiiiiiiiciincrite e em st e ART Core Heat TransSfer .. et veeestiete e s r e resee e sresesa e sessestraeeas VYolume-Heat-Source Convection Analyses ...cocccviiiiiiiniiicnceiincncincenee Heat Removal from Fuel Dump Tank ..o esiiiee s esteee e ssrve e sssevenns Heat Transfer in Helical Pipes. ..ottt s Heat Capacity Measurements on Fluoride Mixtures .....ccocoocieriiiniienin e Heats of Fusion of Fluoride Mixtures .....cccoicieieicoieiiviniee e seeresse e Viscosity Measurements on Fluoride Mixtures ...ocooveiiiriiiiniiieiciecce e Thermal Conductivities of Liquids..cccocciiviniieiiieicirccre e ser e e s sese e s RADIATION DAMAGE ..ot e ten e raer e s sbe s r e b e ere s rserae s b e Disassembly and Examination of Irradiated Equipment......ccccovinciiricicicnnnnnenne. MTR In=Pile Lo0p ottt e b e e e b LITR Miniature 1n-Pile Loop .ttt ARE ComPonents ..ovviceeeiiieierieressieeesrts e e ee e ssre e serteeseseeeena g e sesnneesrensrneesrraeeasas Thermocouple Errors in In-Pile lLoop Temperature Measurements..........c........... Holdup of Fission Gases by Charcoal Traps .....ccceievceininccnireiere e Creep and Stress CorroSion vt re e seveee e cee e seenas e st nssaras In-Pile Tube-Burst Creep Tests .iieiiecciieiieieieeeeereeeeeteecee b e cveeesae st aens Alternate In-Pile Apparatus ...t e ebe s e s snenae ANALYTICAL CHEMISTRY OF REACTOR MATERIALS ... Determination of Oxygen in Sodium ... n-Butyl Bromide Method. ..ot Distillation Method ...ttt et e ste e e s et e e s s as e sb b e e Determination of Traces of Rare-Earth Elements in Stainless Steels................ Spectrophotometric Determination of Aluminum in Flucride Salts with Aurin Tricarboxylie AT et se st st v s es e s ne e e ere s Determination of Water in Hydrogen Fluoride Gas ....ccooceeeeiieiiciiinnecc e Determination of Oxygen in Zirconium Oxide by Bromination .....cccoevvvrvrerrinnnnne. Determination of Tantalum in Fused Fluoride Salts .....ocoooioiiiniiiincciiieane. Direct Determination of Traces of Fe(lll) in NaF-KF-LiF-UF4 .......................... Determination of Traces of Fe(lll) in Mixtures of Alkali-Metal Fluoride Salts ANP Service Laboratory .................................. ---------------------------------- .................................. .................................. ---------------------------------- ---------------------------------- .................................. ---------------------------------- .................................. ---------------------------------- ---------------------------------- ---------------------------------- .................................. ---------------------------------- ---------------------------------- .................................. ---------------------------------- .................................. .................................. ---------------------------------- .................................. .................................. ---------------------------------- ---------------------------------- .................................. .................................. ---------------------------------- .................................. ---------------------------------- .................................. .................................. ---------------------------------- ---------------------------------- .................................. ---------------------------------- .................................. ---------------------------------- ---------------------------------- 163 164 167 167 168 170 170 170 171 174 177 177 177 178 179 179 181 182 182 182 182 183 183 183 184 184 184 186 186 186 187 189 189 190 191 192 192 192 193 10. 11, 12, 13. RECOVERY AND REPROCESSING OF REACTOR FUEL ..ottt s e 194 Pilot PEAnt DeSign wocoecoiiceee ettt e ettt et b et s ie st ks e b e ke ket e bt s e b be e te st nas s s abenranens 194 Engineering Developments ..ot b s e s 194 CONTACTOF vitiiieeeeeete st st sse s saeste st s anarar e e e s be s ressetsebeese e s eset e beaaas ssssesnasaeamtaeee s snaneseneeraasbasntssat e anarassnerasas 194 Fre@ze Valves ettt cee et ra et e e s he et ae e s e e e ba e raaare s erenee 194 Resistance Heating of Transfer Pipes and Waste-Discharge Nozzle ..o 194 Process Development ..o ittt eb e a bbb eab et e bR s ae b s b be et et e bt e et aes 195 Fused-Salt FIuorination SHudies .. r e ree s ere s s e sbasr et bsebs s st 195 NaF Absorption Capacity and UF , Loss on Desorption.........occoierminieiminncincii s 196 UF6 Decontamination in NaF Absorption Step ...ttt s 197 PART Hll. SHIELDING RESEARCH SHIELDING ANALYSIS Lot ieee s st e st ae s e b s b s e e s eaesasasseseesasssnesassassesaasensnseneaes 203 Air Scattering of Co®? Gamma Rays: Theory vs EXperiment.......coooooomeieceeioniiiesoneeneiensenceseeesnieees 203 Energy Absorption Resulting from Gamma Radiation Incident on a Multiregion Shield With S1Ob GEOMEITY c.c.viitieini ettt s re e ettt s arn e sr e aesr e s 203 Integral Equations for the Flux Density near a Thin Foil and for Neutron Scattering in Air in the Presence of the Ground ... et se e ae s st s b ane s 203 SHIELD DESIGN L.ttt ettt es s s tese et st st s aa st es e s b ts e sab s e e s s st se st esseresnessoanansebanns e 204 Calculation of the Sodium Activation in the Heat Exchangers of Circulating-Fuel Reactors .......... .204 Calculation of Activation by Core Neutrons.......co.oceiiiiii ettt st et ee st 204 Calculation of Activation by Delayed Neutrons.......coevevciriincniiienc e 207 Calculated Total Activations for Several Reactors......c.cocoiverirecerinierireeriesecersssss e ssrsssssnne, 209 LID TANK SHIELDING FACILITY Lottt sttt et v b ee st st st et st reean s e emns 210 Static Source Tests with Mockups of a Reflector-Moderated Reactor and Shield .......oooevveiviveinenne. 210 Effect of Varying Lead Thicknes s ...ttt nee e 210 Study of Secondary Gamma-Ray Production ...t 213 Effect of Heavy Metals in the Reflector ..o 218 Effect of Borating the Water Shield ..ot eer et e sse s s 219 Dynamic Source Tests with Mockups of a Reflector-Moderated Reactor and Shield .....ccocccceiiicis. 219 Sodium Activation in the Heat Exchanger ...t e 220 Fission-Product Gamma-Ray SPectrum ..ottt ettt et e bena s 223 - xiii R Ty O 2 X P ANP PROJECT QUARTERLY PROGRESS REPORT SUMMARY PART I. REACTOR THEORY, COMPONENT DEVELOPMENT, AND CONSTRUCTION 1. Reflector-Moderated Reactor Construction work is now under way on the Aircraft Reactor Test facility, Package 1 con- struction, which consists in alterations to the existing building, an addition to the building, and installation of the reactor cell, was 10% complete at the end of the quarter. A major design change was made in order to provide more space for NaK system piping and equipment, as well as for a special heat-dump facility for the removal of heat from the fuel fill-and-drain tank, Drawings and specifications for a modified version of pack- age 2 work were completed. This work now con- sists in the installation of the diesel generators and facility, the electrical control center, and the spectrometer room electrical and air-conditioning The piping work for nitrogen, air, cooling water, helium, lubricating oil, and hydraulic oil drive systems was removed from the package 2 work unit and was designated as package A, Design work continued on package 3, which covers the experimental instruments, controls, process lines, and process equipment to be installed by ORNL forces, System flowsheets and instrumentation lists were prepared in an initial attempt to define the entire ART. The flowsheets show the various components of the system; the annunciator pickup and presentation locations; the operating con- ditions of temperature, pressure, and flow rates; the control stations for electrically operated components; the normal valve positions and uniform valve identification; and line sizes, The fuel-to-NaK heat exchanger design was modified to overcome interferences at the headers and to provide additional space in the region of the headers for the beryllium support struts., Di- mensional and operating data for the heat ex- changers were revised accordingly, The problem of cooling the fuel fill-and-drain tank was studied. Dual NaK systems are to be utilized for both cooling and heating the tank, Each system is to be capable of operating individu- equipment, ally and of removing the total maximum heat load of 1.75 Mw from the fuel in the tank. The basic mechanical design work on the ART lead-and-water shield and supports was completed, and the procedure for assembly and installation was studied, A comparative evaluation of rubber and aluminum containers for the borated shield water showed aluminum to be preferable on all counts, including weight. Additional core flow studies were made on both the full-scale aluminum and plastic core models, Photographs were made of the flow potterns ob- tained with various core inlet configurations. Intensive studies of the various flow patterns are under way, An equipment layout was prepared for the Engi- neering Test Unit (ETU), and work was started on design of the facilities required for operation of the unit, Design problems have delayed the issuance of firm drawings of the reactor com- ponents, and it is estimated that operation of the ETU will be about three months behind the origi- nally scheduled date of September 1, 1956. Instrumentation |ists were prepared for each ART system, and instrument designated. Five permanent instrument information centers will be used, in addition to three temporary ones, and some instruments will be located and read on the equipment, Layouts of control and instrument panels and boards are being prepared, and elementary wiring diagrams have been com- pleted. The layout of the control rod and drive mechanism is being prepared. sensor IOCO‘HOI’IS were The problems of procurement of special reactor materials and components are being investigafed. Equipment for machining the large beryllium re- flector blocks is being prepared by The Brush Beryllium Co. A newly developed Hydrospin process shows promise of being satisfactory for the fabrication of the required six sets of thin Inconel shells, which vary from 18 to 54 in, in The processing of the 100,000 Ib of special CX-900 Inconel required for tubing is under way. Welders qualified under the special ORNL procedures are being trained for several diameter, - | ANP PROJECT PROGRESS REPORT vendors who are interested in fabricating the heat exchangers and radiators. Operation of the high-temperature critical experi- ment revealed that the ZrF, vapor traps were inadequate, Therefore a program for the develop- ment of suitable equipment is under way, 2. Experimental Reactor Engineering An in-pile loop was inserted and successfully operated in the HB-3 beam hole of the Materials Testing Reactor, after attempts to operate two other loops had failed because of difficulties with electrical heating circuits, The fuel mixture circulated was NaF-ZrF4-UF4 (53.5-40-6.5 mole %); the maximum fuel temperature was 1500°F; the maximum mixed-mean temperature differential was 150°F; and the total operating time was more than 400 hr, The loop power generation was estimated to have been 20 kw at the position of maximum insertion, The power density was calculated to have been 0.6 kw per cubic centimeter of fuel mixture., The loop is now being disassembled for shipment to ORNL for examination. Operating difficulties are being studied in connection with design work on the next in-pile loop. Studies of the electrical failures which prevented startup of the first two loops have indicated that the helium atmosphere used allowed electrical breakdown and that the Savereisen cement used as insulation tended to conduct current at high temperatures, Since nitrogen is a better insulator than helium, tests are being conducted to de- termine the effectiveness of nitrogen as a purge gas for the nose section. Grade A Lava sleeves are to be substituted for Sauereisen cement at the heater terminals in future loops. Five electrically heated and three gas-heated fused-salt—Inconel forced-circulation loops were operated during the quarter. A study was made of temperature measurements and heat distribution in the gas-heated loops, and the electrical heaters of the electrical-resistance-heated loops were modi- fied so that fluid-to-wall temperature ratios could be varied. Also, six forced-circulation loops were operated with sodium and noneutectic NaK. Further evidence was obtained that the plug indicator does give a measure of the oxide contamination in the sodium, Seals of various designs are being tested for possible use in the lower position in the ART fuel pump., Several seals have been found to meet the 2 %@_ Cm t.‘i Ve Ak g fiff& 1 stringent specification of ‘'leakage of oil into helium across a pressure differential of 0 to 5 psi of not more than 2 cm?3 per 24 hr."’ Full-size models of the impeller and the volute for the ART sodium pump are being tested with water. Initial tests indicated that the impeller would meet head and flow requirements at slightly lower speeds than predicted. The cavitation tests performed, to date, indicate that the pump will operate safely at inlet pressures below the pres- sure planned for the reactor sodium pump, Tests of an ART fuel pump are under way with high-temperature (1400°F) NaK as the pumped fluid. A heat-source plug was used in the initial tests to simulate gamma-ray heating, but equipment failures necessitated discontinuing this phase of the experiments., Tests with a heat-source plug will be resumed later, if more satisfactory means can be found to simulate reactor conditions, Two high-temperature loops are being fabricated for testing ART-type fuel pumps with NaK and with fuel mixtures, and loops are being designed for testing ART-type sodium pumps and ART-type NaK pumps at high temperatures. Operation of intermediate heat exchanger test stand A was continued, and test stand B was placed in operation. Heat exchangers were de- signed for future operation in these tests stands and in test stand C, which is being fabricated. York radiator units Nos. T and 2 were tested on stand A, and unit No. 3 is being installed, Ex- amination of unit No, 1, which failed after 140 hr of operation, revealed severe buckling of the side plates of the radiator core because of differential thermal expansion between the tube matrix and the The side plates of unit No. 2 were split to allow the tubes to move freely, and this unit operated for 435 hr. The diffusion cold trap installed in the loop was found to be ineffective, and it was replaced with a circulating cold trap, which, although somewhat more effective, is not satisfactory, Test stand B has completed a total of 670 hr of operation, including 247 hr of bifluid operation, in a series of radiator, heat exchanger, and circu- lating-cold-trap tests. The system is now in steady-state operation with a temperature differ- ential imposed, side plates, A small heat exchanger test stand is also oper- ating with a 20-tube fuel-to-NaK heat exchanger, a 500-kw NaK-to-air radiator, and a circulating e R 2 . :... 5 cold trap. The NaK circulating in the cold trap is cooled by an economizer located in the inlet line to the cold trap and by air circulated in the cold-trap cooling coil. With this arrangement it has been possible to operate at cold-trap temper- atures below 150°F, with system temperatures as high as 1500°F. A preliminary analysis of the data indicates that cold-trap performance is repro- ducible and that the oxide concentration of the NaK can be repeatedly reduced, The test data obtained with the intermediate and small heat exchanger test stands have been correlated. The seven radiators tested, which were built by three fabricators from the same specifications, have given a wide range of per- formance. An intensive investigation is under way to determine the reasons for the differences, The correlation of the heat exchanger data revealed an apparent rise in the NaK pressure drop during operation of the system with a temperature differ- ential imposed, which is also being investigated, Work has continued on the fabrication of a one- fourth-scale outer core shell model that is to be subjected to cyclic thermal and pressure tests. Two anvil bending-test assemblies were put in operation to obtain basic information on the be- havior of Inconel under strain cycling at elevated temperatures in both inert and corrosive atmos- pheres. The preliminary tests have indicated that temperature has a major effect on the initia- tion of cracks. The number of cracks produced for a given number of cycles was greater at 1200°F than at 1400°F and was greater at 1400°F than at 1600°F. A test assembly for use in the development and testing of cold traps and plugging indicators is ‘being fabricated, and a test assembly for use in the development of a ZrF, vapor trap was placed in operation. Preliminary results indicate that there must be very close control of the temperature of the inlet line to the trap. The ZrF, vapor precipitated and plugged the line when there was a temperature differential of 25°F or more on the 0.5-in.-0D, 0.025-in.-wall inlet line, A full-scale aluminum model of the top portion of the ART is being fabricated, Problems of design and operation will be investigated with this assembly, which will be equipped with Inconel fuel and sodium pumps and the external piping needed to permit the pumping of water under simu- lated reactor flow conditions. PERIOD ENDING DECEMBER 10, 1955 3. Critical Experiments room-temperature experiments on the reflector-moderated-reactor critical assembly have been completed. Power-distribution measurements across the fuel section at the equatorial plane of the reactor show an edge-to-center ratio of 4.5, with only about 10% of the fissions at the center being produced by thermal neutrons; adjacent to the beryllium, this fraction is 70%. Neutron flux measurements, with gold foils, show a similar depression and energy distribution in the fuel. In a mockup of the fuel-to-NaK heat exchanger, shielded from the reactor by a ‘é-in. thickness of boral, the fissioning rate was about 0.1% of that in the reactor fuel region. The contribution to the over-all critical assembly reactivity by a sample of Several vranium in the fuel region was shown to vary by a factor of 2 as the sample was moved from the edge to the center of the annulus. The effect of the berytlium at the end of the assembly on the over-all reactivity was found to be about 20 cents per linear inch. Experiments with the reflector-moderated critical assembly operated with liquid fuel (NoF-ZrF4-UF4) at 1200°F were also completed, The mid-plane power-production distribution in the liquid fuel showed an edge-to-center ratio of about 2, The value was determined from the fission-product activity induced in uranium disks which remained in the annulus throughout the test, A preliminary loading has been made of a mockup of a solid-fuel modification of the refiector- moderated reactor proposed by the Nuclear De- velopment Corporation of America (NDA). In this reactor design the solid fuel is cooled with sodium. It was found that the critical mass of the mockup had been underestimated by NDA, A 35% increase in the uranium loading and a 23% decrease in the steel content of the core were necessary in order to make the assembly critical, The core now con- tains 31 kg of U235, PART Il. MATERIALS RESEARCH 4, Chemistry of Reactor Materials Phase equilibrium studies of various fluoride systems were continued in order to obtain a better understanding of their structures and to devise and detect improved fuels. In the ZrF ,-UF, system, in which solid solutions are the only crystalline phases that have been observed, it was ANP PROJECT PROGRESS REPORT shown that there is no major miscibility gap. Diagrams of the LiF-UF, and KF-UF, systems, which had been based almost entirely on thermal- analysis data, were revised on the basis of ex- aminations carried out on quenched specimens. The study of mixtures in the NaF-LiF-'-UF4 system gave strong indications that no fuel mixture with a suitably low melting point was available. Thermal-analysis experiments and examinations of slowly cooled melts in the NcF-KF-Zer system were continued, and o series of quenching experi- ments will be carried out to obtain additional information, Work continued on determining the phase re- lationships in the NaF-LiF-BeF,-UF, system. Fuel mixtures with low melting points appear to be available, The viscosity of the mixtures was improved by lowering the BeF, content, but it appears to be likely that only very little additional lowering can be achieved by reducing the BeF, content to below about 20 mole %. Additional clarification of the phase relationships in the KF-BeF , and NaF-KF-BeF, systems was achieved. As a matter of long-range interest in materials for producing higher reactor temperatures, some phase studies were carried out on alkali-metal- fluoride~MgF , and —CaF, systems, Improved equilibration methods gave values of K, =2.5and 0.64 at 800 and 600°C for the reaction FeF,(d) + Hy(g) & Fe(s) + 2HF(g), where the concentrations in solution are expressed as mole fractions and the gas pressures are in atmos- pheres. Studies of the stability of vanadium and nicbium in NaF-LiF-KF-UF, at 600 and 800°C indicated that vanadium was quite stable and that niobium was not, Preliminary studies indicated that the behavior of tungsten was similar to that of niobium, Analytical data from equilibrations with tantalum are not yet available, Additional data were obtained on the solubility and stability of CrF, in NaF-ZrF, and of CrF, in NaF-LiF-KF. In each case the major portion of the dissolved chromium retained its original valence., The solubility of CrF2 in NaF-ZrF, increased as the available CrF, was increased and thus confirmed previous indications that solid CrF,.UF , is formed. Experiments were carried out in which UF_ was added to molten fluorides at 900°C in copper containers. No reduction occurred in the case of LiF, but, with NaF and, to an even greater extent, with KF, alkali metal was formed and partially vaporized from the melt, A study was made of the reactions of uranium metal with LiF and KF., In the case of KF, metallic potassium was produced, but LiF was stable, Since NaF-LiF is unstable to uranium metal at 750°C, it can be concluded that NaF was the unstable constituent. Preliminary experiments indicated that used fuel material can be freed from chromium by reduction with zirconium followed by standard hydrofluori- nation-hydrogenation treatment, if its oxide content is kept low by careful handling. A satisfactory commercial supply of ZrF4 has been located for use in fuel preparations. The hafnium content of the material limits it to use in tests that do not involve nuclear activity. Fuel samples for small-scale corrosion and physical-property tests continued to be prepared in laboratory- and pilot-scale facilities. The stockpile of fuel from the large-scale production operation reached an all-time high, and a temporary shutdown will occur if usage continues to be less than the predicted requirements, Low-carbon nickel reactor cans are on order to replace the A-nickel cans which have experienced frequent failures. The fuel mixture was removed from the high- temperature ART critical experiment, with the exception of the enricher system. Almost all the fuel is still in this system, and, since it is a prototype of the ART enricher, it will be used for additional experiments. Two in-pile loops were filled with enriched fuels, Relative viscosity-composition studies were made in the NaF-LiF-ZrF, system. Additional emf measurements were carried out in the fused salts, and additional optical and x-ray data were obtained on various compounds in the fluoride systems, 5. Corrosion Research Examinations were completed of several Inconel forced-circulation loops in which fluoride fuel mixtures were circulated. Additional data were obtained which confirmed previous observations that, for a constant temperature, increases in wall temperature signifi- cantly increase the amount of corrosion, The bulk-fluoride-fuel-mixture attack in the loop which operated with a maximum wall temperature of about 1710°F was 8 mils after 332 hr, in comparison with attack to a depth of 5 mils in 681 hr in a loop which had a maximum wall temperature of about 1582°F, These data were obtained with NaF-ZrF .UF, as the circu- lated fuel mixture. The alkali-metal-base mixture NaF-KF-LiF (11.5-42-46.5 mole %), with sufficient UF, and UF, added to give 11.9 wt %, was circu- lated in another Inconel forced-circulation loop, and, as previously, the ottack was negligible, but there was o continuous metal deposit in the cold leg. Several Inconel forced-circulation loops were operated in a study of the effect of the oxide content of the sodium being circulated. Mass- transferred deposits were found in the cold legs of loops operated with and without added oxygen, as Na,0,, and with and without cold traps. The addition of barium as a deoxidant did not measur- ably decrease the mass-transferred deposit. A loop operated with a maximum sodium temperature of 1000°F demonstrated the significant effect of temperature on mass transfer; the loop was free of deposited material, and there was no hot-leg attack, A special thermal-convection loop was operated with thermocouples attached to the hot-leg wall, and it was found that a temperature difference of 160°F existed between the maximum fluid temper- ature and the hottest section of the wail, This temperature difference is higher than that found for the forced-convection loops and accounts, in part, for the attack in thermal-convection loops being deeper than that in forced-circulation loops with similar bulk-fluid temperatures. A series of six thermal-convection loops was operated with NaF-ZrF -UF , (50-46-4 mole %) to which various amounts of ZrH, had been added to reduce the UF, present to UF,. The total uranium content of the fuel mixture decreased during operation of the loop in each case, and the decrease was apparently associated with the appearance of a metallic layer on the loop walls. However, the corrosive attack was negligible (0.5 mil) compared with that in a standard loop operated without ZrH, added to the fuel mixture, A group of thermal-convection loops was fabri- cated from special Inconel-type alloys in which the chromium content was the primary variable. When the loops were operated under standard conditions with NaF-ZrF -UF , (50-46-4 mole %) as the circu- lated fluid, lf was found that the depth of attack PERIOD ENDING DECEMBER 10, 1955 in the loops that contained less than 10% chromium was appreciably lower than it was in a standard Inconel loop; that is, 2 to 3 mils in comparison with about 13 mils in 1000 hr. Loops fabricated from several nickel-molybdenum alloys in which NaF-ZrF 4+ UF, (50-46-4 mole %) was circulated also showed little attack, 1 to 2 mils, in 1000 hr, A special thermal-convection loop was fabri- cated for testing brazed Inconel segments in the hot-leg section. This loop was operated for 1000 hr at a hot-leg temperature of 1500°F and a cold-leg temperature of 1100°F with NcF-ZrF4-UF4 (50-46-4 mole %) as the circulated fluid. The Inconel segments were brazed with Coast Metals alloy No. 52 (89% Ni-5% Si-4% B-2% Fe). No evidence of mass transfer was found, but the brazed joints were porous and shrinkage voids were present, In a series of seesaw tests of Inconel T-joints in sodium and in fuel mixtures, a 75% Ni—25% Ge mixture was found to have fair resistance to both mediums, Buttons of six differ- ent braze materials were also tested in static sodium and in static NGF'ZFF4-UF4 (50-46-4 mole %), and two palladium-rich buttons were tested in NaOH and in the fluoride mixture. Coast Metals alloy No. 52 and General Electric alloy No. 81 showed fair corrosion resistance to sodium, and the 60% Pd-40% Ni alloy showed good re- sistance to both the fluoride mixture and NaOH. Thermocouple assemblies were exposed to sodium and to NaF-ZrF -UF, (50-46-4 mole %) i seesaw apparatus so fhat 1119 effect of various amounts of Chromel-Alumel in the weld nugget could be studied, The nuggets with low Chromel- Alumel content were unattacked after 100 hr ot 1500°F in sodium and in the fuel mixture, and those with high Chromel-Alumel content were attacked to a depth of about 0.5 mil in the fluoride mixture, The Inconel tubes on which the welds with high Chromel-Alumel content were made were more heavily attacked in the nonweld areas than were the tubes with the welds of low Chromel- Alumel content, A second boiling-sodium—-Inconel loop was operated, and the scheduled 1000-hr test was completed, in contrast to the 400 hr of operation for the first such loop. After the 400-hr test no mass-transferred crystals were found in the cold trap, but heavy intergranular cracking to a depth of 50 mils was found. After the 1000-hr test, mass-transferred crystals were visible in the cold ANP PROJECT PROGRESS REPORT trap, and there were cracks to a depth of 25 mils, The temperatures in areas where cracks were found varied from 1150 to 1325°F. A loop fabricated from type 348 stainless steel is now in operation, In a test of type 1043 mild steel (0.433% C) in static sodium at 1830°F for 100 hr, significant decarburization occurred, The carbon content of the mild-steel decreased, and the carbon content of the capsule walls in contact with the increased; the two types of capsules used were Armco iron {(0.019% C) and type 304 ELC stainless steel (0.022% C). Two special heats of Stellite were exposed to static lithium at 1500°F for 100 hr. They were attacked to a depth of 2 to 3 mils, in general, but there were isolated areas that suffered very heavy attack, The variations in susceptibility to attack may be due to composition variations in the specimens, Titanium carbide cermets with cobalt- and nickel- base alloys as binding material were exposed to NGF-Zer-UF4 (53.5-40-6.5 mole %) for 200 hr in seesaw apparatus, The hot and cold zones of the apparatus 1500 and 1200°F, re- spectively, Ten specimens were tested and only two showed attack; the other specimens remained unattacked, The compositions and methods of fabrication of these materials, which were sub- mitted by the Sintercast Corp. of America, are not specimens sodium were at known, Inconel valve disks and seats were flame-plated with a mixture of tungsten carbide and cobalt for testing to determine the suitability of the coating as a hard-facing material, To be suitable, the coating must be resistant to solid-phase bonding, galfling, and wearing. In an initial thermal-cycling test from room temperature to 1500°F, the coating did not crack, but, under conditions similar to those used in brazing, the coatings cracked. Kentanium cermet valve parts were held in contact at a pressure of 10,000 psi and exposed to static NQF-ZTF4-UF4 (53.5-40-6.5 mole %) for 150 hr at 1500°F. There was no indication of bonding and the seating was uniform., The joint between the cermet and the Inconel showed no signs of attack; there was no distortion of the z-in.-thick nickel shim; and the cermet pieces did not show any signs of cracking. Several experiments were performed to determine the effect of a small air leak into a fused-salt— Inconel system. The test capsules were heavily attacked by the air-contaminated fused salt, thermal Measurements were made of the self-decomposition of fused sodium hydroxide into water and sodium oxide, Data obtained in recent measurements of the vapor pressure of water over fused NaOH give a value for AH® that agrees within 8% with a value Tests of mass transfer and corrosion of structural materials in fused NaOH were continued, The materials studied were Inconel, nickel, iron, Hastelloy B, and types 310 and 405 stainless steel. An apparatus is being constructed for studying the solubility and the rate of solution of the constituents of Inconel in sodium and in NaK as a function of temperature and of oxygen content of the liquid metal, in order to clarify the mecha- nism of the mass-transfer reaction observed when fiquid metals are circulated in inconel loops. Also, an apparatus was constructed for studying eutectic mixtures by zone melting. known from other measurements. 6. Metallurgy and Ceramics A third 500-kw NaK-to-qir high-conductivity-fin radiator, two 20-tube fuel-to-NaK heat exchangers, and two 100-tube bundles for intermediate heat exchanger No. 3 were fabricated. Job samples containing welded and brazed joints of the type used in the fabrication of intermediate heat ex- changer No. 3, submitted by outside vendors, were examined and evaluated. The two NaK-to-air radiators that failed in service were examined. One of the radiators was fabricated by ORNL and the other by the York Corp. to the same specifications. The ORNL radiater was found to have been so badly damaged by fire that the cause of failure could not be detected. However, it was found that the York Corp. radiator failed because of the initiation of a fracture in a braze fillet by shear forces and the propagation of this fracture through the tube wall by tensile forces or combinations of shear and tensile forces, The tensile loading was caused by differences in cooling rates between the support members and the finned tubes when cooling air was forced across the high-conductivity- fin surfaces. The shear forces were caused by a difference in cooling rate between the support members and the bottom flanged plate and the finned tubes. Modifications have been made to the radiator design to relieve the tensile and shear forces. The results obtained to date in the brazing-alloy development program were correlated with cor- rosion data, and the alloys that are satisfactory for use in the fabrication of heat exchangers and radiators were selected. The alloys selected for radiator fabrication on the basis of compatibility with both liquid sodium and air were Coast Metals alloy Nos. 50, 52, and 53; standard and low- melting Nicrobraz; General Electric alloy No. 81; and an 80% Ni—10% Cr-10% P alloy. The alloys selected for heat exchanger fabrication on the basis of compatibility with both fluoride fuel mixtures and sodium were an 80% Ni-10% Cr—10% P alloy; standard and low-melting Nicrobraz; Coast Metals alloys Nos. 50, 52, 53, and NP; a 70% Ni-13% Ge-11% Cr—6% Si alloy; and a 50% Ni—-25% Ge-25% Mo alloy. Alloys in the nickel-molybdenum system are being investigated in the search for structural materials with sufficient corrosion resistance and high-temperature strength for use in circulating- fuel reactors. The strength and corrosion re- sistance of Hastelloy B have proved to be superior to those of any previously tested material. How- ever, this alloy has a tendency to age-harden in the 1200 to 1500°F temperature range. The re- sulting decrease in ductility is so severe that its use as a structural material in a circulating-fuel reactor would be limited. Hastelloy W has been found to have strength properties similar to those of Hastelloy B, and the results of preliminary tests of the creep properties show that Hastelloy W has less tendency to age at the temperatures of interest. Several new vacuum-melted nickel-melybdenum alioys have also been creep-tested. Several of the alloys showed poor ductility that may be attributed to insufficient degassing in the vacuum-melting process. Despite the poor ductility of the alloys, their stress-rupture properties are equal, or su- perior, to those of Inconel, Preliminary work on the fabrication of seamless duplex tubing was completed, in that tube blanks of nickel-, Inconel-, and Monel-clad type 316 stainless steel were extruded, without difficulty. The conditions for extrusion of Hastelloy B are still not definite. Three Hastelloy B—clad type 316 stainless steel duplex billets were extruded, but they were unsatisfactory because of cracking and roughness that resulted from poor lubrication. Attempts to improve lubrication of Hastelloy B by flame-spraying with a heavy layer of type 304 stainless steel were unsuccessful. The extrusion PERIOD ENDING DECEMBER 10, 1955 of Hastelloy W tube blanks from forged billets was unsuccessful because of hot-shortness of the material, Attempts are being made to produce suitable B,C-base tiles for the ART neutron shield and to evaluate the properties of such tiles, Com- patibility tests of Inconel and B,C with various coatings and diffusion barrier materials in the temperature range of 1500 to 2000°F are under way. An irradiation tesf is being run to determine whether helium and lithium will be released from B,C at high temperatures. If these gases are released, it may be necessary to vent the shielding layer. Several other methods of preparing neutron shielding are being studied, such as casting mixtures of borides and metal and fabricating cermet compositions that are high in B,C or boride content, The problems associated with the fabrication of Inconel-clad niobium are being studied. A number of specimens fabricated with copper—stainless steel foil or tantalum foil diffusion barriers have been prepared for mechanical-property tests. Several compositions are being tested for use as a gamma-ray shield of low thermal conductivity for the ART pump impeller shafts. Two compo- sitions which are of satisfactory density are tantalum-constantan and tungsten carbide—con- stantan, Thermal-conductivity data are being obtained for these materials. It was demonstrated that control rods containing 30 vol % rare-earth oxide could be fabricated by canning a cermet-type core in a capsule of suitable cladding material and hot swaging the composite. Physical-property data were obtained for the ex- truded control-red parts containing rare-earth oxides and for iron-zirconium alloys. Investigation of the various methods by which small-diameter tubing may be inspected has con- tinved. The Cyclograph (an eddy-current instru- ment) gives very effective, high-speed indications of flaws as small as 10% of the tube-wall thick- ness. For detection of minute flaws, the ultra- sonic method is effective and can be applied to large-scale inspection. The eddy-current-probe instrument is being developed for use with the ultrasonic method as a simultaneous inspection on the same mechanical scanning operation. This will provide close correlation of the defect indi- cations from both methods, Tubing that is ]41 in, in diameter and 7 ft 6 in. in length can now be ANP PROJECT PROGRESS REPORT inspected with this dual inspection method, and, upon completion of a new probe coil, 3/]‘S-in.-dic:; tubing 28 ft long can also be inspected. 7. Heat Transfer and Physical Properties A forced-convection fused-salt heat-transfer system that includes a mechanical pump has been fabricated., This system is to be used to study the heat-transfer characteristics of the fused salts at Reynolds numbers higher than those covered previously. The friction characteristics of small, drawn, Inconel tubes such as those to be used in the ART fuel-to-NaK heat exchanger were de- termined over the Reynolds-number range 50,000 to 200,000; the friction factors of the Inconel tubes fell only about 6% above the normal corre- lation for flow in smooth tubes. Additional hydrodynamic studies were made of annuli of the type being considered for the fuel channel of the ART reactor core. Calculations of the pressure drop through a core fabricated according to the present ART design were made for straight-through rotational-flow case. A study of flow distribution in the annulus in which sodium will be circulated to cool the outer core shell of the ART has been initiated. The conceptual design, as well as the detail flow, as well as for a designs of the test section, entrance sections, exit section, and mixing chambers, of the volume- heat-source experiment for determining the temper- ature structure to be expected in the ART core has been completed, and fabrication of the test section is in progress. Several analytical heat- transfer solutions for volume-heat-source systems were developed; for example, turbulent flow in curved channels was investigated, and the thermo- couple error that results when a thermocouple is used to measure the temperature of a flowing fluid having a volume heat source was studied. A general study of the problem of heat removal from a fuel dump tank of an ART-type aircraft reactor after shutdown was initiated. The wall-temperature asymmetry predicted to occur in a helical pipe that has fluids with volume heat sources flowing through it was demonstrated experimentally in o glass helix that contained a circulating fluid with an electrically generated volume heat source. The predicted results correlated well with data ob- tained from operation of the MTR in-pile loop. The enthalpies and heat capacities of two rubidium-bearing fluoride mixtures were determined: namely, RbF-ZrF,.UF, (48-48-4 mole %) and LiF-RbF (43-57 mole %). The heats of fusion of nearly all fluoride mixtures that have been studied to date are presented. Viscosity measurements were made for nine different fluoride mixtures. A transient-cell method that has been used to measure the thermal conductivity of liquids was studied and used successfully for measurements on water and molten sodium hydroxide., Pre- liminary measurements made by the constant-gap method gave a value of about 1.2 Btu/hr-#12(°F /1) for the thermal conductivity of RbF-ZrF ,-UF (48-48-4 mole %). 8. Radiation Damage The in LA N 2 s 23 g /{// R /7///, - ) i B R T RN el L ST , N 20D SN \\ SN 5 3 RN s i o 45'/5/%41 N s, - | ,// ////'/ = LIQUID-LEVEL PROBES ~<—~—____ 3 iy ke , -~ o Yl 3 i 3 { A N N e 33 3 LT v : S L X \ 5 Ao 3 s 77 " N B L T T %3] 3 s T SN - SHIELD PUMP SUPPORT AND /9 : PLUG TANK COVER- , l ErT s /)/f/ ‘/.// s ot RE PRESSURE- BREAXDOWN PLUG— y ’ SLINGER /. . | ] MOUNTING RING— 77 - I - L AN 7K N . : .,\ | Y ! " TR ’ [ LW 11 © . o X : = = | VOLUTE TOP PLATE m: / E\ | ; \\\ /2 2 /;/ __C'i - \{ { /' ! 3 S ARV A R % TANK SHELL = e v iy . 5 3 %0 ¥ o o i o > A 5 E (oS \ | : - VOLUTE BOTTOM PLATE L% R LA o /j,///f/’/ 5/:, R “ X Ll AT PO, R ~ T —_— *_L ALL DIMENSIONS ARE IN INCHES. D ——FILL AND DRAIN LINE Fig. 2.7. High-Temperature Short-Circuit Pump-Test Loop. 39 ANP PROJECT PROGRESS REPORT UNCLASSIFIED UNCLASSIFIED ORNL—-LR—DWG 11266 5 ORNL—LR—DWG 11265 16 S " T i n B ‘ ! 1 o SN o | , |- HEAT INPUT TO SHIELD PLUG: 375w | | T o T ‘ 1 | | OlL COOLED | ! o NO SHIELD PLUG HEATING ] 14 ‘ l OIL QUTLET TEMPERATURE: 110°F | _ | 14 - / I T _{ " OIL NOT COOLED T+ | 11 | 0 a MEASUREMENTS MADE wiTH C_p b [ OIL OUTLET TEMPERATURE: 160°F 1 \T " THERMOCOUPLES LOCATED 'L’“i [ B ! { | 0,4 - MEASUREMENTS MADE WITH | : ON OPPOSITE SIDES OF ‘ o DL F,l ' % THERMOCOUPLES LOCATED ON .| 12 ~—4-—r—h— PUMP BARREL T - f | A } | | OPPOSITE SIDES OF PUMP | E ooy o ] < fill_{v b BA&R_EFTJF*# = ; fi“fi% — w j e o z b b | ] : | L NI P A \T Lol o e 210 ‘ T»fi—iTjfi-—l— — —] = | e ! . ' ! | T T W l \ ! | ‘ % fi,l,,!ifi P L L d —‘I —L‘ 4‘— L R ! - . U - 2 S —T—krlfig TJ w 8 j | ESIRED TEMPERATURE GRADIENT | ! > | - : : : ! . L g s a 1 o L = ' N ] < r ‘ r» Lt i K_,‘_ e _ . _ o b L gs| - | T S DESIR%%JSI\QZETRATURE < . i i \ e .. - . - p— L 1100 ’—+ - : b 200 — {000 - — - +——t — 25 w = _ N 180 " 500 i — 160 |- £ 800 [— 2.0 & o 140 |~ & 700 b—t 44 ] § = B - < 20}~ L 600 1.5 € w S o g Lz = — — —_ —_ 5 100 500 B CUMUL ATIVE z Y BOTTCM SEA ol 2 o S 60 |— 300 — + | —f — — 1 40 |— 200 — INLET OIL TEMPERATURE — | | L los ) — — e e 20 |— (00 | _CUMULATIVE TOP SEAL LEAKAGE o T - 1] ol 0 _ 1 [ J o o o F. » 8 10 12 14 16 8 20 22 24 26 28 30 32 34 36 38 40 42 44 46 48 50 52 DAYS OF OPERATION Fig. 2.10. Performance of MF-2 Pump with NaK as the Pumped Fluid, Run 3. 40 UNCLASSIFIED ORNL-LR-DWG {1268 * { 7 T T 1T 1 T T AI \‘Yfi NO SHIELD PLUG HEATING QIL NCT COOLED OlL OUTLET TEMPERATURE: {90°F — THERMOCOUPLES LCCATED ON QPPOSITE SIDES OF PUMP BARREL T | | J \ . O,A MEASUREMENTS MADE WITH - \ | ot \\ ‘ ! - E ¥ X =2 < w O 5 + 8 — ' S \ { ’ g - — | < N DESIRED TEMPERATURE GRADIENT W & | L1 s ‘\ ‘, ‘ ] < ! I | b ™~ q 1 L | ~ \ M, J < 2 — 1 T < EXPERIMENTAL DATA >\ 0 ~ ! | | ey . L L DS 0 200 400 600 800 1000 1200 TEMPERATURE (°F) Fig. 2,11, Temperaoture Gradient Along Pump Barrel in Run 3, The original test program called for simulation of gamma-ray and neutron heating in the shield plug by using a Calrod heater to produce about 2.5 kw of heat. However, as mentioned before, this part of the test program has been delayed be- cause of troubles with the plug heater. Apparently, several problems must be solved. A satisfactory means must be found for insulating the elecirical leads (close mechanical clearances complicate this) for operation up to 1500°F in a helium atmos- phere. Also, a heat source must be constructed that can supply approximately 7 w/cm® without reaching a temperature that will destroy the heater. As a further complication, two reactor operating conditions must be simulated by the shield plug. First, to simulate reactor operation at zero power, the lower surface of the plug must be maintained at a minimum temperature of 1000°F in order to prevent fuel from freezing on its lower face and seizing the slinger impeller of the pump. Second, to simulate reactor operation at full power, the plug must be able to conduct away enough heat to maintain the lower face at a temperature of 1500 to PERIOD ENDING DECEMBER 10, 1955 1600°F so that the mechanical strength of the Inconel can on the shield plug will not be impaired. A second test stand, similar to that shown in Fig. 2.7, will be in operation soon, This test [oop will circulate fuel, and a hydraulic motor rather than an electric motor will be used to drive the pump, Data will be taken on operation of the hydraulic drive, as well as on pump operation, High-Temperature Pump-Performance Test Stands R. Curry H. Young Pratt & Whitney Aircraft Two high-temperature loops, described previ- ously,? for testing MF-2 ART-type fuel pumps are being fabricated, Cavitation, performance, shake- down, endurance, and acceptance tests on MF-2 pump rotary assemblies will be made with NaK and with the fuel mixture I\Ic1|:-ZrF4-UF4 (50-46-4 mole %) as the circulated fluids and with fluid temperatures of up to 1400°F, The design layout has been completed for a highetemperature loop for testing MN-2 ART-type sodium pumps, Cavitation, performance, shake- down, and acceptance tests on MN-2 pump rotary assemblies will be made with sodium at tempera- tures of up to 1400°F. Most features of this loop are similar fo those of the loop for testing MF-2 ART-type fuel-pump performance. High—temperature performc:nce-tesfing loops are being designed for testing PK-2 and PK-A ART- type NaK pumps. Cavitation, performance, shake- down, endurance, and acceptance tests on NaK pump rotary assemblies are to be made at NaK temperatures of up to 1400°F, HEAT EXCHANGER DEVELOPMENT E. R. Dytko Pratt & Whitney Aircraft R. E. MacPherson Aircraft Reactor Engineering Division intermediate Heat Exchanger Tests R. D. Pedk M. H. Cooper J. M, Cooke L. R. Enstice Pratt & Whitney Aircraft Designs of the two fuel-to-NaK heat exchangers that are to be tested in stands B and C were com- pleted. Type IHE-3, shown in Fig. 2.12, is an 3R. Curry and H. Young, ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 44. 41 ANP PROJECT PROGRESS REPORT all-lnconel 100-tube design, for which 0,25-in,-0D, 5.6-ft-long, 0.025.in,~wall tubing is to be used. Type IHE-8, shown in Fig, 2.13, which is also all Inconel, is a 144-tube design, for which 0,1875-in.- OD, 5.6-ft-long, 0.025-in..wall tubing is to be used. These heat exchangers will be tested in a e 02510 0D MATERIAL : INCONEL A SPAGCER WIRE, 0.031 x 0.055 in. ({1} o — ——— — &737in- EXCHANGER TUBE ‘TOP FILLER BAR FUEL HEADER NoK HEADER NaK INLET Fig. 2.12. Fuel-to-NaK 144 TUBES 01875-in. 0D 0.025-in WALLS -._ SIDE FILLER BAR -~ \n\\\\t“ SECTION A-A T UBE SPAGER~- - SPAGER WIRE, 0.021 x 0.045 in. ({1} A ;§§w§m@§5“m§fi§'“mg§fififiw ; // -1 T A P g e s FUEL INLET 7 T % |§§; ]»:i‘ it _ e =7 ,jf f.,. e = = Y | J : \l 1 | e —¢ — e EXCHANGER ST SRR H e H SR o 4 ~-X_ FUEL HEADER "™ NaK HEADER GOVER NoK OUTLET ll—- S Fig. 2.13. Fuel-to-NaK 42 f \ ! —TUBE SPACER \ —BOTTOM FILLER BAR -TUBE SPACER TUBE SHEET, g-in. THIGK, 4-in. R, HOLES APITCH 0.300 x 0.270 in. regenerative system of the type shown in Fig. 2.14, with two tube bundles operating in series. Figures 2,15 and 2.16 show the range of operating conditions which can be expected from the two heat exchangers, and Figs. 2.17 and 2.18 show stresses which will be encountered in the test ORNEC= g— DWG 9622 ~100 TUBES 0025-in WALLS SECTION A-A SIDE FILLER BAR A= \EXCHANGER SHELL 2 o 2 4 & e 10 12 - : . S SCALE N INCHES Heat Exchanger Type IHE-3. L ORNL-LR-DWG 9623 NaK INLET l 7= FILLER BAR -~ {TOP AND BOTTOM = i/ SHELL - P~ TUBE SHEET, %-in. THICK, 4-in. R, HOLES A PITCH 0.334 x 0.304 in. ! MATERIAL : INGONEL J 2 0 2 4 6 A per ) SCALE IN INCHES | — — 87375 — —— «l Heat Exchanger Type IHE-8. PERIOD ENDING DECEMBER 10, 1955 . ORNL- LR-DWG 1269 AIR N AIR QUT TWO 500-kw RADIATCR UNITS [ NaK PUMP | Y NoK FLOW ™~ [ FUEL FLOW ] 1 FU PUMP NaK-TO-FUEL EL FUEL-TO- NaK HEAT EXCHANGER HEAT EXCHANGER TUBE BUNDLE NO. 4 TUBE BUNDLE NO.2 NG L i L S FUEL FLOW NaK FLOW 1-Mw GAS-FIRED FURNACE W ’ Fig. 2.14. Diagram of Intermediate Heat Exchanger Test Stand, units throughout the operating range. The thermal stresses (calculated by using Castigliano's theo- rem) are imposed on the heat exchanger tubing by the difference in temperature between the tubing and the sheil of the heat exchanger. These stresses arerelieved by creep and are the stresses of primary interest in a thermal-cycling program. The pressure stresses (calculated by using simple cantilever-beam theory) are imposed on the tubing by the drag forces of the flowing fluid and are not relieved by creep; rather, the tubing is steadily deformed, Test stand A, the first experimental assembly operated in the present series of intermediate heat exchanger tests, was described previously.? It was operated for 434 hr during this quarter in a 4R. D. Peak, M. H. Cooper, and L. R. Enstice, ANP Quar. Prog. Rep, Sept. 10, 1955, ORNL-1947, p 45. series of radiator and circulating-cold-trap tests, A chronological description of the test operations is presented in Table 2,4, The stand was op- erated with York radiator units No. 1 and No. 2 until unit No. 1 failed. When unit No. 1 had been removed, operation was resumed and continued until unit No. 2 failed. York unit No. 3 is cur- rently being installed, Examination of unit No. 1, which failed after 140 hr of operation, revealed severe buckling of the side plates of the radiator core because of differential thermal expansion be- tween the tube matrix and the side plates. With no air flow across the radiator core, the tempera- tures of these plates could be 125°F below the tube temperature, and, dwing periods of heat re- moval from the radiator, a femperature difference of 300 to 1200°F could exist. Temperature dif- ferentials of such magnitudes would cause severe compressive stress in the tubes at the start of air 43 ANP PROJECT PROGRESS REPORT ORNL -LR—-DWG 11270 1700 | | HE-3 ]f— NakK INTO BUNDLE NQ. 1& | ! 1600 - ! : — T | 1500 1400 1300 TEMPERATURE (°F) 1200 I"FyUEL QUT OF BUNDLE NO.2* - INTO BUNDLE NO.B— - i e AU ‘ 1100 |——- ; | NaK INTO BUNDLE NO.2 1000 180 — i i NoK FLOW RATE 160 ‘ e — ] G: J o =~ 440 b - e —_ — I i | o~ I WS 120 | - - —_ ] xZ | L g Lo ! 8~ 400 | — - __LOG MEAN TEMPERATURE __| w'a : DIFFERENTIAL IN HEAT - ‘ EXCHANGER | — a i . | R S w0 i Ly | o : Bl = | FUEL FLOW RATE S ; | ~ o i - ! o = 40 —mmm— - e - o 2 TOTAL FUEL - ar PRESSURE DROP - 20 [— =~ TOTAL NaK —— ,./ PRESSURE DROP — | —_— e ——ri 0 — ! 0 1000 2000 3000 4000 FUEL REYNOLDS NUMBER Fig. 2.15. Predicted Operating Performance of Fuel-to-NaK Heat Exchanger Type IHE-3 in Test Stand B. flow, and, if it is assumed that the compressive stress is relieved by creep, equadlly severe tensile stress would be present when the air flow was stopped. The results of metallurgical examination of the unit are reported in Sec. 5, ‘“Metallurgy and Ceramics.”’ Prior to initiation of operation with York unit No. 2, the side plates were split so that the tubes could move freely with core temperature changes. This unit subsequently failed during themal- cycling tests after 435 hr of operation. Figure 44 L ORNL—-LR--DWG 11271 1700 ‘ HE-8 | NoK INTO BUNDLE NO.ty | / 1600 |- p— ! — i | ! FUEL OUT OF BUNDLE NO.1 INTQ BUNDLE NO. 2 500 |- ———-TF"= — ‘ N : Tty | s a—— o w 1400 o 2 & 1300 L —I — & NaK OUT OF BUNDLE NO.{ | = w i 1200 I"LUEL OUT OF BUNDLE NO.2 INTO BUNDLE NO.t%y_.— ] T--'-—-_-—__ — ey —— 1100 {-NoK INTO BUNDLE NO. 2. | - ‘ J 1000 180 —1 —‘— . 160 '— —_— = Nok FLOW RATE e - 440 - ot . -l i T 5 5 W 3 120 —— — - - - w © ! TS i Lo LOG MEAN TEMPERATURE B —~ 100 —— - DIFFERENTIAL IN HEAT — g & EXCHANGER | x S £ ol o L E ‘ g g £ wn g W Ww = a O O -t 20 b——— . TOTAL FUEL PRESSURE DROP J ! 0 0 1000 2000 3000 4000 FUEL REYNOLDS NUMBER Fig. 2,16, Predicted Operating Performance of Fuel-to:NaK Heat Exchanger Type IHE-8 in Test Stand C, 2.19 shows the split side cover plates and the re- sults of the NaK leak and fire. Since the location of the leak was almost the same as the location of the leak in York unit No. 1, that is, near the side plate and near a horizontal stiffener plate, the same conditions which caused York unit No. 1 to fail undoubtedly weakened York unit No. 2. In the case of York unit No. 2, as with ORNL units Nos. 1 and 2, a marked increase in the radiator friction factor (~12%) occurred prior to failure, The exact cause of this increase, which was not e ORNL—LR-DWG 11272 7 x108 — . . ' IHE-3 6x10°% |——— — - — _ TUBE BUNDLE HEAT LOAD -~ 5X10° f—r flm_lfi £ \ ~ 2 | & axiw® % S O — a | Q J 3x108 kfi —_———— b i | o FURNACE HEAT LOAD T 6 ! | exwo® L ] Tfi st ] ‘ 0 18,000 _ [ 16,000 l __\‘ THERMAL STRESS 14000 f— ——o 3 12,000 F — - 10,000 | ‘ - ' . PRESSURE STRESS DUE TO FUEL SYSTEM QPERATION ! 1 | PRESSURE STRESS DUE TO BIFLUID OPERATION | PREs eos DUE TV TATION STRESS ON HEAT EXCHANGER TUBING (psi) 8,000 6,000 — 4,000 |- —J 2000 o PRESSURE STRESS DUE TO “ NaK SYSTEM OPERATION ! t 0 e | ] 0 1000 2000 3000 4000 FUEL REYNOLDS NUMBER Fig. 2.17. Predicted Operating Heat Loads and Stresses of Fuel-to-NaK Heat Exchanger Type IHE-3 in Test Stand B. encountered prior to the failure of York unit No, 1, has not yet been established. _ During the shutdown for installation of York radiator units Nos, | and 2, the diffusion cold trap initially used on this stand was changed to a cir- culating cold trap. The trap, as modified, con- sists of a 6-in, pipe, 36 in, long, that contains York Demister stainless steel packing in the top 30 in. The NaK flowed into the trap through an economizer section (]/Z-in. pipe inside a 2-in. pipe, 12 in. long} mounted on top of the cold-trap body, PERIOD ENDING DECEMBER 10, 1955 6 ORNL—LR—DWG 11273 7 X10 1 1 IHE—8 exio® - — I R [ TUBE BUNDLE HEAT LOAD 5 x10° ax10® b 3x10° HEAT LOAD (Btu/hr) 2x10° 18,000 ] #R\J: M p— \ 18,000 —— - — THERMAL STRES S 14,000 - S / 10,000 ;—h — PRESSURE STRESS DUE TO FUEL SYSTEM OPERATION - PRESSURE STRESS DUE TO BIFLUID OPERATION S5TRESS DUE I , 12,000 8000 S 6000 | ——— 4000 | PRESSURE STRESS ON HEAT EXCHANGER TUBING {psi) 2000 — - 0 1000 2000 3000 4000 FUEL REYNQOLDS NUMBER Fig. 2.18. Predicted Operating Heat Loads and Stresses of Fuel-to-NaK Heat Exchanger Type IHE-8 in Test Stand C. flowed through the central ]/2-in. pipe to the bottom of the trap, flowed back up over the packing to the economizer, and then flowed back to the system. The trap was cooled with forced air, and heaters were provided for local temperature control. It was installed in series with, and downstream from, the plug indicator. The series arrangement was not suitable, however, because cooling of the plug indicator caused plugging to start in the cold trap. A parallel installation was therefore devised, and with this installation the trap could be operated at 45 ANP PROJECT PROGRESS REPORT wd ABLEF 2.4, SUMMARY OF OPERATION OF INTERMEDIALE HEAT EXCHANGER TEST STAND A Operation continued from previous quarter. NaK system repaired by the installation of new York radiator units Nos. 1 and 2, new circulating cold trap, and plug indicator; trap and indicator piped in series. Blower operated to measure radiator air flow at 90°F, Filled system with NaK NaK system operated isothermally at 1500°F with flow rate of 80 gpm to check operation of cold Plug indicator gave erroneous measurements of sodium oxide because of interaction of Radiator tests with the NaK system operated at various flow rates and temperatures between Radiator tests with the NaK system operated at various flow rates and temperatures; blower at Tests terminated because of a {eak in York This radiator was cut out for examination, and the air duct was changed to The circulating cold trap was modified with a throttling valve so that the plug indicator and the cold trap operated in parallel NaK streams. NaK system operated isothermally at 1300°F with flow rate of 40 gpm to check operation of cold Radiator tests with the NaK system operated at various flow rates and temperatures; blower at Cold trap operated intermittently to Hours of Operation Remarks 690 (56% Na—44% K) at room temperature, 690 to 720 trap. cold teap in series with it. 720 to 748 1000 and 1650°F; blower run at various air flow rates, Eleven measurements taken. 748 to 823 NaK system operated isothermally at 1300°F with flow rate of 80 gpm. 823 to 830 various air flow rates. Three measurements taken. radiator unit No, 1. fit the remaining York radiator unit No. 2. Blower operated to measure radiator air flow with NaK system at 90°F. 830 to 909 trap. Sodium oxide plugging temperature reduced from 1000 to 750°F. 909 to 999 various air flow rates. Twenty-five measurements taken, control the axide content, 999 to 1017 1017 to 1712 1112 to 1124 NaK system operated isothermally at 1200°F with a flow rate of 40 gpm. Cold trap operated to reduce the oxide plugging temperature from 920 to 800°F. NaK system shut down to calibrate the electromagnetic flowmeter on the main NaK system, NaoK system operated isothermally at 1400°F with a flow rate of 40 gpm. Cold trap operated in an attempt to reduce oxide content; plugging temperatures varied erratically between 1010 and 720°F. Radiator thermally cycled according to the following schedule: high power for 4 hr; NaK into radiator at 1500°F, out at 1140°F; flow, 42 gpm; air in at 90°F, out at 1100°F; low power for 2 hr; NaK circulated isothermally at 1425°F; flow, 42 gpm; no air flow. York radiator unit No. 2 failed on the start of the third cycle to high power. The radiator was cut out for examination by the Metallurgy Division, and the cold trap was cut out for examination by the Materials Chemis- try Division. A new circulating cold trap, a new screen filter, a duplicate plug indicator, and York radiator unit No. 3 are being placed in the NaK system. 0.3-gpm NaK flow with a minimum temperature of 350°F. At the end of the test when the cold trap was cut open for examination by the Materials Chemistry Division, it was found that the trap had not effectively removed oxides. The equivalent of 75 g of O, was found, which corresponds to the Sibid., p 49. 46 removal of 260 ppm of O, from the 80 gal of NaK used in the system during operation of the cold trap. The NaK samples taken during this quarter indi- cate that the cleaning done to remove the fuel constituents which leaked into the NaK system at the time of failure of IHE No. 2, described pre- viously,®> was not complete. Fuel constituents LUNCLASSIFIED PHOTO 24962 Fig. 2.19. York NaK-to-Air Unit No. 2 from Inter- mediate Heat Exchanger Test Stand A after NaK Leak and Fire. continue to showup in miscellaneous NaK samples, as illustrated below: Zirconiom Uranium Content Coatent (wt %) Drain from circulating cold trap 0.007 g Leak from York radiator unit No. 1 1.4 wt % 1.9 L eak from York radiator unit No. 2 1.9 wt % Since the sump, plug indicator, and cold trap were opened and cleaned manually before reinstalliation on the chemically cleaned system, it is certain that the fuel constituents were held up in portions of the pump and furnace which were not adequately reached by the cleaning solutions. The analyses of the NaK drained from the circu- lating cold trap and of the water used to wash the trap are also of interest: PERIOD ENDING DECEMBER 10, 1955 Amount in NaK Amount in Water {ppm) (9) Nickel 10 10 Chromium 30 27 Iron 89 1.5 These results substantiate other evidence that chromium is preferentially leached from the Inconel in the systems and is carried to the cold regions of the loop in the NaK stream. Test stand B, described previously,® has com- pleted a total of 676 hr of operation, including 247 hr of bifluid operation in a series of radiator, heat exchanger, and circulating-cold-trap tests, A chronological description of the tests conducted and the difficulties encountered is given in Table 2.5, The system is now in steady«state operation with a temperature differential imposed. As stated in Table 2.5, considerable trouble was experienced with operation of the cold trap and the plug indicator, On stand B, these components are operated in parallel, with the radiator pressure drop as the driving force. The NaK that enters the cold trap is cooled by a forced-air precooler. It is convection-cooled as it passes through the trap, and then it is reheated at the exit end before being discharged through a throttling valve and an electromagnetic flowmeter and being returned to the system. The plug indicator is identical to that on stand A, Most of the trouble associated with operating the plug indicator and the cold trap has been due to plugging of the small-diameter, ?Ié-in. tubing on the discharge sides. This tubing was utilized in order to magnify the electromagnetic flowmeter signal. A cold trap of the type being installed in stand A will be installed in this loop, and the small-diameter tubing in both the cold«trap and plug-indicator circuits will be removed and re- placed with larger tubing. As stated in Table 2.5, the 2-in, IPS sched-40 pipe adjacent to the electromagnetic flowmeter on the main NaK circuit developed a leak after 374 hr of operation. The resulting fire destroyed the aluminum pieces and fused the pipe surface so that the exact cause could not be determined. Examination of the pipe after it was cut out showed no defects detectable by x ray or Dy-Chek. 47 ANP PROJECT PROGRESS REPORT TABLE 2.5. SUMMARY OF OPERATION OF INTERMEDIATE HEAT EXCHANGER TEST STAND B Hours of Operation Remarks a Stand assembled with Pratt & Whitney Aircraft radiator units Nos. 1 and 2 end with ORNL heat exchangers (type IHE-3) Nos. | and 2. Filled system with NaK (56% Na—44% K) at rcom temper- ature. Operated blower to measure radiator air flow at 90°F. Cto9 Heated NaK system to 1200°F with flow of 80 gpm. All NaK drained into sump when remotely operated dump valve failed to close when draining some NaK to adjust level in pump. Added manually operated handle to dump valve, 9 to 88 NaK system operated isothermally at 1400°F with flow of 80 gpm to check operation of circu- lating cold trap. Sodium oxide plugging temperature reduced from 940 to 660°F. 88 to 139 Radiator tests with the NaK system operated at various flow rates and at temperatures between 1000 and 1550°F; btower at various air flow rates. Cold trap valved out. Plug indicator plugged up and then unplugged when outside resistance heat was codded. The NaK-system friction factor® gradually increased 25% during tests. 139 to 173 NaK system operated isothermeally at 1400°F with flow of 80 gpm with cold trap on to reduce oxide content. Faulty operation of the trop washed oxide back into the system initially; how- ever, the trap rteduced the plugging temperature from 850 to 730°F. The NaK-system friction factor was reduced by 13%. 173 to 175 Furnace test with the NaK system operated at flow rate of 140 gpm, a temperature of 1350°F into the furnace and a temperature of 1500°F out. The Struthers Wells furnace burned 8100 sefh of natural gas with a thermal efficiency of 48%. Heat transfer to the NaK was 3,69 x 10% Bu/hr (1.68 Mw). Cold trap valved out. 175 to 350 The NaK system operated isothermally at 1400°F with a flow rate of 80 gpm. Cold trap in operation; plug indicator plugged, unplugged, and finally plugged seolid. 204 to 206 Fuel system filled for cleaning with N«:F-ZrF"-UF4 (50-46-4 mole %) and operated isothermally at 1400°F for 2 hr at a fuel flow rate of 30 gpm; fuel then dumped. System filled with new fuel, NaF-ZrF4-UF4 (50-46-4 mole %). Bifluid isothermal operation ot 1400°F with NaK flow of 80 gpm and fuel flow of 30 gpm. Test stopped because of leak in 2-in. NaK pipe adjacent to the electromagnetic flowmeter. Replaced NaK pipe and the plugged tubing downstream from the plug indicator. 374 to 388 Heated NaK system to 1200°F with flow of 80 gpm. NaK drained into sump while repairing NaK dump valve. Found helium line to NaK pump plugged with sodium oxide, 388 to 453 NaK system operated isothermally at 1400°F with a flow rate of B0 gpm to check cold-trap operation. Electromagnetic flowmeter on plug indicator failed., No change in NaK-system friction factor. 453 to 604 Heat exchanger tests in bifluid operation at various temperatures between 1000 and 1500°F; NaK, fuel, and blower air flow rates varied. Radiator data also token. The circulating cold trap plugged, NaK-system friction foctor increased 35%. Thirteen heat transfer measurements taken. 604 to 628 Bifluid isothermal operation at 1300°F; NaK flow rate, 145 gpm; fuel flow rate, 30 gpm. Attempted to unplug cold trap by using resistance heating on tubing. NaK-system friction factor did not change. 628 to 676 Bifluid operation with a temperature differential across system; NaK flow rate, 145 gpm; high NoK temperature, 1600°F; low NaK temperature, 1100°F; fuel flow rate, 42 gpm; high fuel temperature, 1435°F; low fuel temperature, 1250°F., NaK-system friction factor increased an additional 8%. Operation continuing. *The friction factor was taken to be the ratio of the NaK-system pressure drop to the square of the NaK flow rate. 48 Test stand C is being fabricated and is 25% complete, Radiator units supplied by the Cam- bridge Comp. and heat exchangers being built by Black, Sivalls & Bryson, Inc., will be tested first, Small Heat Exchanger Tests J. C. Amos Aircraft Reactor Engineering Division L. H. Devlin J. G, Tumer Pratt & Whitney Aircraft Small heat exchanger test stand B, shown in Fig. 2.20, was placed in operation on November 3, 1955, The test units installed in this stand are a 20-tube fuel-to-NaK heat exchanger and a 500-kw NaK-to-air radiator (ORNL No, 3), both built by the Metallurgy Division of ORNL. The heat ex- changer tubes, 0.25 in., in outside diameter and 0.025 in, in wall thickness, are amanged in a 4 by 5 matrix with 0,282-in. center-to-center s pacing, The test stand includes a circulating cold trap, described previously® and shown in Fig., 2.21, for removing oxides from the NaK. The NaK circu- lating in the cold trap is cooled by an economizer located in the inlet line to the cold trap and by air circulated in the cold-trap cooling coil. With this arrangement, it has been possible to operate at cold-trap temperatures below 150°F, with sys- tem temperatures as high as 1500°F. Relative oxide concentration in the system is determined by an air-cooled plugging indicator, also shown in Fig. 2.21. Much of the initial operation of this test system has been devoted to obtaining data on cold-trap operation, A preliminary analysis of the data indicates that cold-trap performance is re- producible and that the oxide concentration of the NaK can be repeatedly reduced, Small heat exchanger test stand C is approxi- mately 65% complete, This stand is identical with stand B, with minor exceptions. The heat exchanger is identical with the one in stand B and was also built by the Metallurgy Division of ORNL, but the 500kw NaKto-air radiator is York unit No. 5. : The design of a 25-tube heat exchanger, with 0.1875-in.-0D, 0.025-in.wall tubing and 0.2175-in, cenfer-to-center spacing, that is to operate at ART temperature and flow conditions was com- 6F. A. Anderson and J. J. Milich, ANP Quar, Prog, Rep. Sepr. 10, 1955, ORNL-1947, p 54. PERIOD ENDING DECEMBER 10, 1955 pleted. This unit is shown in Fig. 2,22, Negotia- tions are in progress to obtain four of these units from outside vendors. Present plans call for testing these four heat exchangers, two additignal 20-tube heat exchangers, and six additional 500k w radiators, in small heat exchanger test stand$§ 3 and C. 5 Heat-Transter and Pressure-Drop Correlations R. D. Peak Pratt & Whitney Aircraft J. C. Amos Aircraft Reactor Engineering Division The test data on radiator and heat exchanger heat transfer and pressure drop obtained with the intermediate and small heat exchanger test stands have been correlated. The parameters of the vari- ous radiators tested are presented in Table 2.6, and those of the several heat exchangers in Table 2.7. The air pressure-drop correlation for the radia- tors is shown in Fig. 2.23, where the pressure drop per tube row (in inches of water), comrected from the air film temperature to 60°F, is plotted against the ratio of air flow rate to the free-flow area. The points for ORNL-1,-2, PWA-1,-2, and York-1,-2 are averages for the two radiator units, since the two radiators were usually operated simultaneously. The air heat-transfer correlation for these radiators is shown in Fig, 2,24, where the air Nusselt number at the film temperature is plotted against the air Reynolds number at the film temperature. The air-side heat-transfer coef- ficient was calculated by taking into account the NaK-side and the tube and fin collar metal heat- transfer resistances. The NaK-side resistance was calculated by using the Lubarsky? equation: Nu = 0.625(Re x Pr)%-4. It was found that the fin efficiencies ranged from 88 to 94%. It is apparent that the seven radiators built by the three sources from the same specifications resulted in a wide range of performance. The NaK pressure-drop correlation is shown in Fig. 2.25, where the friction factor is plotted against Reynolds number. Also included in the correlation are two water tests: one with new, 0.1875-in.-0OD, 0,025 in.-wall, Inconel tubing, the ’B. Lubarsky and S. J. Kaufman, Review of Experi- mental Investigations of Liquid-Metal Heat Transfer, NACA-TN-3336 (March 1955). 49 0¢ CONTROL PANEL Fig. 2.20. Small Heat Exchanger Test Stand B. 1¥0d3Y S§53IY50¥d LI3Iroyd dNV s Fig. 2.21. Small Heat Exchanger Test Stand B Showing Cold Trap and Plug Indicator. UNCLASSIF!EDB PHOTO 24915 §S61 ‘0l ¥3gW3D34 ONIAN3 oIy 3d s ORNL-LR-DWG 11274 %~ TWENTY-FIVE, 0.1875~in.-0D, \ 0.025-in.- WALL TUBES NaK OUTLET NaK INLET SECTION A-A mm\‘m“‘mln\m ~—TUBE SHEET EXCHANGER SHELL | 2, FUEL HEADER LTI AT L I i FUEL’ OUTLET ————- \\\\\\\\\\\ TUBE SPACERS (12 SETS TOTAL) E; T I T % T | ) i . MATERIAL: INCONEL I % 2 i L e Fig. 2,.22. Twenty-Five Tube Fuel-to-NaK Heat Exchanger. ' LIS TSI N ARSI TI LA o -~ FUEL INLET =——-1 L30d3Y SSFAD0Ud 1D03Ir0dd dNV PERIOD ENDING DECEMBER 10, 1955 TABLE 2.6. PARAMETERS OF 500-kw NaK-TO-AIR RADIATORS FROM VARIOUS SOURCES ORNL-1,-24 Parameter PWA.1,.2 ORNL-37 York-1,-2¢ Tube material Inconel Inconel Inconel Number of tubes 72 72 72 Inside diameter of tubes, ft 0.0115 0.0112 0.0115 Tube wall plus fin collar thickness, ft 0.003 0.003 0.003 NaK free-flow area, ft2 0.00742 0.00714 0.00742 NaK-side heat-transfer area, f2 6.92 6.72 6.92 Mean tube plus collar heat-transfer area, ft2 8.68 8.48 8.68 Fin material Number of fins in vertical stack {_ength of air flow passage, ft Air-side fin heat-transfer areq, H2 Air-side tube plus collar heat-transfer areq, ft2 Air free-flow areq, ftz Air equivalent diameter, ft Fin average thermal cenductivity, Btu/hr§t-°F Type 310 stainless-steel-clad copper, 0.010 in. thick, sched 25.50-25 243 236 254 0.646 0.646 0.646 201 196 210 9 8 8 0.519 0.538 0.509 0.00884 0.00911 0.00832 109 109 109 “Manufactured by Oak Ridge National Laboratory Metallurgy Division. bMflnufactured by Pratt & Whitney Aircraft. Manufactured by York Corp. other with the tubing from the 20-tube fuel-to-NaK heat exchanger bundle, which showed indentations where the wire spacers had pressed into the tubing during 1500 hr of operation. All the data shown were taken during isothermal operation, the range of temperatures being 1200 to 1500°F, except for the IHE-3 data, which were taken during the radia- tor tests on intermediate heat exchanger test stand B, previously described in this report, During these tests the NaK-system friction factor was observed to rise 25%. An intensive investiga- tion is now being conducted on all heat exchanger test stands to detemine the reason for the ap- parent rise in pressure drop observed during op- eration with a temperature differential imposed on the system. The fuel pressure-drop correlation is shown in Fig. 2.26, where the sum of the skin friction and 8R. D. Peak and J. W, Kingsley, ANP Quar. Prog. Rep. June 10, 1955, ORNL-1896, p 35. the spacer friction is plotted against Reynolds number. The spacerfriction data, which were ob- tained by using water-test data (described previ- ously®) corrected for the various spacer thick- nesses and lengths by using the correlation of Cohen, Fraas, and LaVerne,? are also plotted separately for each heat exchanger. The fuel-side heat-transfer correlation is shown in Fig, 2.27 for two heat exchangers, The NaK- side heat-transfer resistance was calculated by using the Lubarsky equation.” The correlation for the 20-tube fuel-to-NaK heat exchanger differs from that presented previously'® because the %G. H. Cohen, A. P. Fraas, and M. E. LaVerne, Heat Transfer and Pressure Loss in Tube Bundles for High Performance Heat Exchangers and Fuel Elements, ORNL-1215 (Aug. 12, 1952), p 47. 10, c Amos, L. H. Devlin, and J. S, Turner, ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 51. 53 ANP PROJECT PROGRESS REPORT UNCLASSIFIED CRNL—LR-DWG 11275 2.0 1 | ! ‘ | ‘ ] | NoK TEMPERATURE ORNL —1,—2 {000~-16800°F YORK —1,—2 1000 —1600° F | | | 05 [ — 0.2 [______‘L___ - AIR PRESSURE DROP PER TUBE ROW (in. OF H,0) (CORRECTED FROM AR FILM TEMPERATURE TO 60°F ) | | l | | | : | ; 3.0 4.0 6.0 8.0 100 1 RATIO OF AIR FLOW RATE TO FREE FLOW AREA (Ib/SeC‘sz) | 0.05 L 4 06 o8 1.0 1.4 2 Fig. 2.23. Air Pressure-Drop Correlation for 500-kw NaK-to-Air Radiators, equivalent diameter used in the Nusselt number STRUCTURAL TESTS was Outer-Core-Shell Thermal-Stability Test . Cross-section area G. D. Whitman A. M. Smith wetted perimeter Aircraft Reactor Engineering Division rather than Work has continued on the fabrication of the one- fourth-scale outercore-shell model and testing 4 cross-section areq assembly, The model is to be subjected to the < tube perimeter cyclic thermal and pressure tests that were de- scribed in the previous report. All the component as used before. parts have been received, and the model assembly 24 PERIOD ENDING DECEMBER 10, 1955 cxowmws TABLE 2.7. PARAMETERS OF VARIQUS FUEL-TO-NoK HEAT EXCHANGERS 20-Tube Intermediate Heat Exchanger Parameter Fuel-to-NaK Heat Exchanger Type Heat Exchanger No. 2 IHE-3 Tube material lnconel Inconel {nconel Number of tubes 20 100 100 Tube outside diameter, in. 0.1875 0.1875 0.25 Tube wall thickness, in. 0.017 0.017 0.025 NoK flow length, ff 5.75 6,03 6.38 NaK free-flow area, ft2 0.00258 0.0129 0.0218 NoK-side heat-transfer area, ft2 3.90 22.6 29.4 Shell inside dimensions, in. 0.910 x 1.130 2.10 x 2.10 272 x 2.72 Spacer dimensions, in. 0.032 x 0.032 0.022 x 0.040 0.020 x 0.040 Number of spacers 1 13 1 Fuel flow length, ft 4.85 5,62 5.62 Fuel free-flow areq, ft2 0.00328 0.0115 0.0173 Fuel-side heat-transfer area, 12 4.75 27.6 36.7 Fuel equivalent diameter, ft 0.0102 AIR NUSSELT NUMBER AT FILM TEMPERATURE 10.0 UNCLASSIFIED ORNL—LR~OWG M2T76 5.0 20 s 4__ At; — o -+ 100 200 500 1000 AIR REYNOLDS NUMBER AT FILM TEMPERATURE 2000 ' 5000 Fig. 2.24. Air Heat-Transfer Correlation for 500-kw NaK-to-Air Radiators. 10,000 55 ANP PROJECT PROGRESS REPORT QA0 1 T T ‘ SO ORNL—LR-DWG 11277 | i 71 | ] i e e e e e S Pt ol 1Ly ® 20— TUBE FUEL-TO~NoK HEAT EXCHANGER AL l | {1 | 0 INTERMEDIATE HEAT €XCHANGER NO.?2 R SO S - /-T- e _L_~a__.l__ — ! ! | [ 1 ' | A INTERMEDIATE HEAT EXCHANGER TYPE IHE~—3, ORNL—1,~2 | Af*f“‘“*‘**”'""F***fl“"7“““%“%“‘w*“*f*" '“"**“'"“"fi'”'"'"“;'"”%’W‘j"fl%’*fi“f'f” WATER TEST OF NEW 0.4875-in.-00, ' | | | o 0.05 t————-——-—/0025-in WALL TUBING "—}’—‘i—‘T ~-—i- ~-—-~-~-~~-*—'--Mf—frmfj‘wflhfij*j‘ ] « | i ; | WATER TEST OF 20-TUBE FUEL-TO-NaK | o g ~~~~~~~~~ _ /’*‘""’T“"*T 4-7,42*1*/‘HEAJT EI>/min) 60 80 250 10 0 20 0 40 5 60 15 80 25 60 80 300 10 0 20 0 40 3 60 15 80 20 94 25 50 B0 150 10 5 20 15 40 35 60 90 80 150 In future tests the valve will be evaluated ot high temperatures in contact with fluoride fuels. The leakage rate, the opening and closing pres- sures, and the general reliability will be deter- mined in order to evaluate the valve for use in the ART. Cold Trap and Plugging Indicator J. J. Milich Pratt & Whitney Aircraft A test assembly for use in the development and testing of cold traps and plugging indicators is being fabricated. This test loop will include a circulating cold trap, an economizer, and a plug- ging indicator, and it will draw from a NaK reser- voir with a capacity of 80 gal. York Demister packing will be used in the circu- lating cold trap to provide the surface to hold precipitated oxides. Precautions will be taken to ensure that the cold-trap temperature is the lowest in the system so that the soturation temperature of the liquid metal with respect to Na,O will not be reached at any point in the system other than the cold trap. The plugging indicator will consist of a disk with 19 holes 0.040 in. in diameter. The rate of flow through the holes will be 1.0 gpm. The fluid 60 will be cooled until the oxide precipitates and partially plugs the holes. An electromagnetic flowmeter will give an indication of the change in flow when the plugging indicator partially plugs. When the change in flow rate occurs, the tempera- ture at the plugging indicator will be recorded and will be assumed to be the saturation temperature of the oxide. This temperature can be translated to oxide concentration by means of a solubility curve. ! Zirconium Fluoride Yapor Trap J. J. Milich Pratt & Whitney Aircraft J. W. Kingsley Aircraft Reactor Engineering Division A trap is being developed for collecting the zir- conium fluoride vapor that will be formed during operation of the ART, since if this vapor were allowed to enter the off-gas system the lines would become plugged with ZrF, and the off-gas system would become inoperative. Various designs are being studied, and a prototype trap is being tested ”C. B. Jackson (ed.), Ligquid Metals Handbook. Sodium-NaK Supplement, TID-5277 (July 1, 1955), p 8. PERIOD ENDING DECEMBER 10, 1955 TABLE 2.10. RESULTS OF HIGH-TEMPERATURE TEST ON ART PROTOTYPE DUMP VALVE Test Qil Pressure Oil Pressure Valve Seating Helium lL-edkage Temperature to Close to Open Pressure Pressure Rate (°F) (psi) (psi) (psi) (psi) (em® /min) 400 80 100 250 10 0 20 2 40 5 60 8 80 10 90 15 600 60 80 250 10 1 20 2,5 40 7 60 10 80 13 800 40 75 250 10 0 to 0.5 20 2 40 1.5 60 2.5 80 2.5 90 3 45 90 300 10 0 to 0.5 20 1 40 2 60 3 80 4.5t0 5 90 5 1000 145 170 250 10 0 t0 0.5 20 1.5 40 1.5 60 2 80 2 90 2 120 190 300 10 0 20 0 40 0 60 0 80 1 90 1 1100 120 245 250 10 0 to 0.5 20 0.5 to 1 40 2 60 2.5 80 2.5 90 3 61 ANP PROJECT PROGRESS REPORT TABLE 2.10 (continued) Test Qil Pressure Oil Pressure Valve Seating Helium Leakage Temperature to Close to Open Pressure Pressure Rate CF) (psi) (psi) (psi) (psi) (em? /min) 1100 190 160 (started to open) 300 10 0 310 (fully open) 20 0 40 0 60 0 80 1 20 1to 1.5 1200 190 210 (start) 250 10 0 320 (fully open} 20 0 40 0.5 60 0.5 80 0.5 90 0.5 240 240 (start) 250 10 0 350 (fully open) 20 0 40 0.5 60 0.5 80 0.5 to 0.75 90 0.75 to 1 290 240 (start) 300 10 0 350 (fully open) 20 0 40 0 60 0 80 0 to 0.25 90 0 to 0.25 1300 250 265 (start) 250 10 0 400 (fully open) 20 0 40 0 60 0 80 0 90 0 300 250 (start) 300 10 0 400 (fully open) 20 0 40 0 60 0 80 0 90 0 60 80 250 10 0 20 0 40 5 60 15 80 25 62 1) PERIOD ENDING DECEMBER 10, 1955 TABLE 2.10 (continued) Test Oil Pressure Oil Pressure Valve Seating Helium Leakage Temperature to Close to Open Pressure Pressure Rate CF) (psi) (psi) (psi) (psi) (em®/min) 1300 60 80 300 10 0 20 0 40 5 60 15 80 20 94 25 50 80 150 10 5 20 15 40 35 60 %0 80 150 TABLE 2.11. RESULTS OF TESTS OF VALVE AFTER IT HAD BEEN CLOSED FOR 168 hr Time Valve Temperature Valve Seating Pressure Helium Pressure Closed (°F) (psi) (psi) Leakage Rate 168 hr 1200 250 100 1.15 em>/he * 168 hr 1400 250 100 0.875 cm>/hr Valve Leakage After the Above Tests 30 min 1400 250 10 12 em3/min 20 21 cm3/min 40 42 cm3/min 60 65 cms/min 30 min 1400 250 10 N em3/min 20 18 cm3/min 40 36 cm>/min 60 56 cm3/min 85 hr 1400 250 90 No detectable leakage with the fuel mixture NoF—ZrF4-UF4 (50-46-4 mole %) as the ZrF, vapor source. The trap consists of a bundle of 30 tubes ar- ranged in a 6-in. sched-40 pipe. The tubes, through which the gas flows, are packed with 1-in. segments of Inconel mesh separated by 1-in. void spaces. 1he tubes are cooled by water circulating through the 6-in. pipe. The trap is connected to a sump tank that is filled with the fuel. To simulate the reactor conditions, helium is bubbled through the sump from a dip line. The helium, which will be present in the reactor to sweep the xenon from the fuel, agitates the fuel mixture and increases the surface area and thus increases the ZrF, vaporization rate, The ZrF , vapor then passes from the sump tank through the trap. The flow rate of the vapor is measured by wet test meters. The fuel is maintained at a temperature of 1300°F, and the exit line from the sump to the trap is held at 1400 * 25°F throughout its length, The results of the first attempts to run tests with equipment indicate that there must be closer temperature control of the inlet line to the trap. [f there is a temperature differential of 25°F or 63 ANP PROJECT PROGRESS REPORT more on the 0.5-in.-OD, 0.025-in.-wall inlet line, the ZrF , vapor precipitates and plugs the line. Water Test of Aluminum Mockup of Top of ART D. R. Ward Aircraft Reactor Engineering Division Several problems connected with the design and operation of the top portion of the ART are to be investigated by using a full-scale aluminum model. Fabrication of the model is under way. This test unit, as shown in Fig. 2.29, will be equipped with lnconel fuel and sodium pumps and the external piping needed to permit the pumping 64 of water under simulated reactor flow conditions. Operation of this system will provide means for studying the following problems: fabricating, as- sembling, and welding of the component parts; controtling pump speed; maintaining flow under both normal and unbalanced conditions; ingassing and degassing; filling, draining, and venting; con- trolling liquid levels; and pressurizing gases. Drawings of the fuel-circuit components of the test assembly have been completed, and parts are being fabricated. The sodium-circuit components will be fabricated later. A test room is being pre- pared, and the major auxiliary equipment has been ordered. 59 ORNL-LR—DH;!! 11280 FUEL. PUMP SODIUM PUMP FUEL PUMP (> (= / GATE VALVE é((l([’l{% SR 'O + ORIFICE 4 5 ){ .+ GATE VALVE FUEL CIRCUIT | / — @ l[h I — T Ty FiN | a0 [ -~ MAIN SODIUM CIRCUIT VISUAL v SODIUM ISLAND CIRCUIT r FLOW ' '@ WINDOW ¢ ! INDICATOR 2 / HEAT EXCHANGER | L 7 -~ D =, sopium t = E,T_. BYPASS f Neaa@n)Pe) e - AN ~ @47 e i “VISUAL W e FILL INDICATOR - || HEAT EXCHANGER N e (== o () — FILL uln[il'!-mi DRAIN DRAIN NN U SO SRR Fig. 2.29. Assembly for Water Tests of Full-Scale Aluminum Mockup of Top of ART. SS61 ‘0L ¥3gW3ID3A INIAONI 4Oy 3d ANP PROJECT PROGRESS REPORT 3. CRITICAL EXPERIMENTS A. D, Callihan . J. J. Lynn E. R. Rohrer Applied Nuclear Physics Division D. Scott, Aircraft Reactor Engineering Division R. M. Spencer, United States Air Force J. S. Crudele W, J. Fader E. V. Sandin S. Snyder Pratt & Whitney Aircraft The present series of critical experiments on the circulating-fuel reflector-moderated reactor, including one assembly operated at about 1200°F, has been completed. A program of critical ex- periments on reflector-moderated assemblies of a somewhat different design proposed by Nuclear Development Corporation of America initiated, has been ROOM-TEMPERATURE REFLECTOR-MODERATED-REACTOR CRITICAL EXPERIMENTS Critical experiments were recently completed on room-temperature reflector-moderated-reactor as- semblies that mocked up the reactor fuel annulus, the beryllium island, and the beryllium reflector and included extensions that corresponded to the entrance and exit fuel flow channels. The ex- tensions are referred to as ‘‘end ducts.”” The basic configuration was described earlier! as assemblies CA-21-1 and CA.21-2, which differed only in the uranium density in the fuel region and therefore in the available excess reactivity, Two modifications of the basic structure were examined, both having been described previously.! In one, assembly CA-22, the outer diameter of the fuel section of one end duct was increased from 5.28 to 6.79 in.; in the other modification, assembly CA-23, the radius of the central beryllium island was increased from 5,18 to 7.19 in, additional results of these experiments are re- ported here, Some Power and Neutron-Flux Distributions The relative power distribution in assembly CA.22 was measured along the radius at the 1A. D. Callihan et al., ANP Quar, Prog. Rep. Sept. 10, 1955, ORNL-1947, p 58. 66 midplane and at three locations in the large end duct. An additional radial traverse was made in the small end duct for comparison. The conven- tional method of measuring the fission-fragment activity caught on 5-mil-thick aluminum disks in contact with uranium foils was used. The results of these measurements, as shown in Fig. 3.1, are similar to those reported previously? for the assembly of basic dimensions. Two radial flux traverses were made in assembly CA-22, both with bare and with cadmium-covered gold foils. One traverse (Fig. 3.2) was at the midplane and extended from the axis a distance of 17.84 in. The other traverse (Fig. 3.3) was in the large end duct in a plane 11.5 in. from the midplane and extended between points 0.31] and 13.72 in. from the axis. It is to be remembered in analyzing these data that the low intensity of slow neutrons in the center of the fuel region introduces a large uncertainty in the values of the cadmium fractions derived from both power and flux measurements. Neutron Production in the Fuel-to-NaK Heat Exchanger The fissioning in the uranium that will be con- tained in the fuel-to-NaK heat exchanger of the circulating-fuel reflector-moderated reactor will be a source of high-energy neutrons which must be attenuated in the shield surrounding the re- actor. A measure of the rate of this fissioning was obtained from catcher foils placed in a mockup of a section of the heat exchanger that was in- corporated in assembly CA-21-2, the assembly having basic dimensions and the lower uranium density. The heat-exchanger-region mockup con- 25, D. Callihan et al., ANP Quar. Prog. Rep. June 10, 1955, ORNL-1896, p 47. CRITICAL ASSEMBLY U PER!IOD ENDING DECEMBER 10, 1955 ORNL—-LR-DWG 11025 \\\\\\\\\\\\\\\\\\\\\\\\\ o ; J':f i (\Ej i BERYLLIUM REFLECTOR x T ):E - E ; g—— :I FUEL \\\i{_-fll_-_j-jfi-_ ’LE;;,T___': ___'_1 \ 7 I T, BERYLLIUM ISLAND END OF REACTOR— 80 T [ 70 ] = 60 — ! 50 i 4 i r l 1 r \ . | © BARE FOIL . ‘ { A CADMIUM COVERED - } DMIUM FRACTION LT J | B | | RELATIVE ACTIVITY {garbitrary units ) *fiq‘t* 410 40 o TL - : —fir | ¥ L VLT 08 L w L LE 30 m&\;l l—fi-# «F v# Jr{——‘» ‘4 06 g | B ] - 20 S fl -- LWL L %Mfi {043 ‘Kc._{L [ ; ] fi | fl é 10 - - ‘L;FL : (—fi‘fi >! i lLL 1~d 02 © 0 L f J l | 1 ‘ }t ‘ | \\ || 0 048 0 & 27 o ¢ 47 o 7 a7 o £ 17 LAYERS OF FUEL FROM ISLAND (! LAYER = 0.004 in. OF U AND 0.142 in. OF TEFLON ) Fig. 3.1. Power Distributions Across the Fuel Annulus and in the Large End Duct of a Room-Tempera- ture Reflector-Moderated-Reactor Critical Assembly, sisted of a layer of boral (300 mg of boron per square centimeter) adjacent to the reflector be- followed by a 1.,3-in.-thick fuel zone containing uranivm metal and Teflon, and a layer of stainless steel 1.437 in. thick. It was located approximately at the midplane, and its dimensions in the plane perpendicular to the flux gradient were 8.625 by 8 in. The arrangement and the results are shown in Fig. 3.4, together with the fission-rate distribution across the core midplane ryllivm, for comparison, Other measurements of neutron leakage from the beryllium reflector have shown that a layer of boral having o density of about 300 mg of boron per square centimeter reduces the relative value of the flux from 1000 to 3; an increase of the boron density by 50% only reduced the flux to a relative value of 2, Radial Importance of Uranium in the Fuel Annulus The effect of adding uranium at different po- sitions across the fuel annulus was measured in assembly CA-23, the assembly with the enlarged central island. Two 7.1875- by 2.875-in. sheets containing a total of 45.58 g of U235 were added to the normal 18-sheet loading and to a reduced loading of 16 sheets. The 16-sheet base was established by the removal of two uranium sheets from their respective positions at one-third and at two-thirds of the distance across the annulus. The test section was at the top of the core region in the assembly and was centered ¥ in. from the midplane. The results are given in Table 3.1 and Fig. 3.5. Importance of Beryllium at End of Reactor The locations of the sodium heat exchangers 67 ANP PROJECT PROGRESS REPORT ACTIVITY (arbitrary units) " AL v ORNL—LR-DWG 11026 1.0 T T T T i ’ ’ 1.0 | | : | - | i i | I l : | f ‘ | | | oo b — — — — A4 --F + — — T -+ ——— o9 | | i < BERYLLIUM ISLAND - —»f=— —— FUEL — — »f= :—-m— BERYLLIUM REFLECTOR ! o : | ‘ i i ‘ | //_..q | | ‘ e N — - — - — % — - — - - - — ———1 08 | B @ i | J/ < - INCONEL CORE | l / | : SHELLS — —— = ' *L // [ J 07— — —\ — + - v’— -t o7 | . ‘ ' | ( | | | }A\CADMIUM FRACTION | \ ; /| | | b )LJ—‘)T—‘ o dos | | # s | | | z ' | S ' l 5 I ‘ g l /S | | : e s crimicaL assemaLy I | ’ | z CA-22 T | | | 2 J | - -, —'l—— L {104 | | CO~p. _ BARE GOLD ACTIVATION | | o \ | , ] NS N, b | | | | | | | ' i S | I | | i | | gr _ ~LA’ - 7$7L+7+7 | Y | |, CADMIUM~ COVERED \ | | il e /GOLD ACTIVATION ! - ' \ | 144 .( '.‘k - 0 — e .N v 1 ‘ | \-—T/ { ’ ] | | 1 6 8 10 12 RADIAL DISTANCE FROM AXIS (in.) Fig. 3.2, Neutron-Flux Distribution Along Midplane of a Room-Temperature Reflector-Moderated-Reactor Critical Assembly with ¢ Large End Duct. and the swirl chambers of the pumps of the ART have necessitated the removal beryllium from one end of the reactor. An estimate of the loss in reactivity to be expected by this removal has been made in some measurements on assembly CA-22. 68 of some of the the resulting degreases in from the critical rods. positions In a second series In one series of three measurements the be- ryllium was removed from reflector and the island in the end of both the 1-in. increments, and reactivity were noted of calibrated control the island beryllium PERIOD ENDING DECEMBER 10, 1955 * AN g v CRNL—LR—DWG 11027 1.0 [ | 1.0 i - r ‘ | | CRITICAL ASSEMBLY ? | CA-22 ! J‘ , ‘ 0.9 ——_*% — —wjfiflf—L— rrrrr fi .9 J e BERYLLIUM ISLAND BERYLLIUM REFLECTOR — —— —= | | | | | . i = j 08 f———— R —_—— L 7T o 1 ! ? } 7 ! INCONEL CORE | ~ / ‘ | J SHELLS ~ / : | | a o7t — — 07 i [ ! } fifADMlUM FRACTION ‘ | /. | | 0.6 — e 1 08 ! » | \ | 5 P s | | / | | 5 ; o / i & § E w a JH N IR 0 ! 2 ‘ z > ~ ‘ o i 5 O ~N I | ) ‘ ; % o4 N R S — .._____,_Lfi_fi_i 0.4 \ { | | | ‘ | _BARE GOLD ACTIVATION 0.3 i~—4lf RT I o =g & a—i —— 03 | ! I ‘ o 5 oel Se_ Y ] | . | l | | i | o CADMIUM ~ COVERED ® GOLD ACTIVATION oo ~e | | - @ | e ; . | | 0 % . 0 2 8 10 12 14 | RADIAL DISTANCE FROM AXIS (in.} ! | Fig. 3.3. Neutron-Flux Distribution Along a Plane 11.5 in, from Midplane of a Room-Temperature Reflector-Moderated-Reactor Critical Assembly with a Laorge End Duct. 69 ANP PROJECT PROGRESS REPORT s R I 77 = ORNL—LR-DWG 11028 N 90 = BERYLLIUM //A=——— FUEL ANNULUS ——>}«""BERYLLIUM | HEaT | /7 I1SLAND ————m] V'///REFLECTOR / EXCHANGER ’ . 1 / F ¢ 1.5 in. —~ Z=4\Y MOckup FUEL N N REFLECTOR — 30 20 _ . — % MULTIPLY ORDINATE 7] L SCALE BY 10~2 10 L . R / | / / 0 22 9 LAYERS OF FUEL FROM ISLAND LAYERS OF FUEL FROM BORAL (1 LAYER = 0.004 in. OF U AND (1 LAYER = 0.004 in. OF U AND 0.142 in. OF TEFLON) Q473 in. OF TEFLON) N N\ o _ RsmanLess\ [ N\ STEEL 0NN **\125 in. a-q BORAL, Y in.=~-3 2222 72 | | s SURFACE OFj§ // E yd RELATIVE ACTIVITY (arbitrary units) N 7 | oz CRITICAL ASSEMBLY CA—-21-2 Fig. 3.4. Power Distribution Along Midplane of Fuel Annulus and Heat Exchanger Mockup of a Room- Temperature Reflector-Moderated-Reactor Critical Assembly, was replaced and an additional increment was removed from the reflector. The results are re- ported in Table 3.2 and Fig. 3.6. Axial Importance of a Neuvtron Source The effect of the axial position of a neutron source (Po-Be) on the neutron level in a sub- critical reactor {assembly CA-23) with a constant neutron multiplication of about 600 has been observed. The source was moved along a channel in the beryllium island, and the change in the neutron intensity at each of several external detectors was observed. A fission chamber and 70 two BF3 ionization chambers were located sym- metrically about the end of the reactor away from the direction of source removal, and a third BF3 ionization chamber, channel A, was placed near the axis at the other end of the reactor. The results from the first three detectors agreed, and their average is given in Table 3.3 and Fig. 3.7, together with the data from channel A. Ail count- ing rates were normalized to those observed with the source at midplane. These results also illustrate the importance of source-counter geome- try on the interpretation of multiplication curves in the approach to critical. TABLE 3,1. ¥ g, PERIOD ENDING DECEMBER 10, 1955 RADIAL IMPORTANCE OF URANIUM IN FUEL ANNULUS Number of Sheets Between Island Distance Between Island and and Added Uranium Change in Reactivity from Added Uranium Normal 18 Z —40 -] 2 e E-so b — —— — P —_— z ISLAND AND REFLECTOR BERYLLIUM REMOVED B ‘LBI -60 S - — ii T ‘_;_T T Z % ; 5 -70 — — - Nd—_— ] | -8 L CRITICAL ASSEMBLY —| — — . B T L ca—z2 1 ~90 ' ‘ 2 20 19 18 17 15 DISTANCE FROM MIDPLANE TO END OF BERYLLIUM REFLECTOR (in.) Fig. 3.6. Importance of Beryllium at End of a Room-Temperature Reflector-Moderated-Reactor Critical Assembly. TABLE 3.3. VARIATION OF COUNTING RATE WITH NEUTRON-SOURCE POSITION fonrat Al Source Position Along Axis; Counting Rate (Normalized to Midplane Rate) Distance from Midplane (in.) Average for Three Chambers Channel A 0 1.000 1.000 5 0.974 0.982 10 0.791 0.837 15 0.458 0.618 20 0.116 0.382 between somewhat thicker CaF, disks in Inconel tubes. During the assembly® of the core shells these tubes were mounted at three axial positions inside the annular fuel region, and they remained there during the entire experiment. measures Relative of the power-production distribution across the fuel region were subsequently deter- mined from the fission-product activity which accumulated in the disks. A schematic drawing of a cross section of the reactor in the longitudinal midplane is shown in Fig. 3.8 with the positions of the foil capsules 3P. Patriarca, ANP (uar. Prog. Rep. Sept. 10, 19553, ORNL-1947, p 133. 72 and the beryllium island and reflector noted. An effective-poison-rod location is also noted. The relative activations of the uranium foils, normalized for decay, are also shown in the figure. The curves represent the power production across the fuel annulus at three longitudinal positions. The proximity of the lowest traverse to the borori-copper plate at the bottom of the assembly accounts for the depression of the fission rate there. At the other two elevations the fission rate on the island side of the fuel region is lower than that on the reflector side. It is not known why the relation between these two values is the inverse of observations made with the room- temperature critical experiments, as shown, for example, in Fig. 3.1. i U PERIOD ENDING DECEMBER 10, 1955 ORNL—LR—DWG {1039 l 0.8 -F — J 0a ____%*F - | 4 AVERAGE OF THREE CHAMBERS 0.6 w—_Jr* - ] COUNTING RATE { NORMALIZED TO MIDPLANE RATE) CRITICAL ASSEMBLY CA—-23 . | C 5 10 15 20 SOURCE POSITION ALONG AXIS (DISTANCE FROM MIDPLANE) (in.} Fig. 3.7. Axial Importance of a Neutron Source in a Room-Temperature Reflector-Moderated-Reactor Critical Assembly, COMPACT-CORE REFLECTOR-MODERATED-REACTOR CRITICAL EXPERIMENTS A reactor embodying the reflector-moderator con- cept and consisting of solid fuel elements located in the channels between the beryllium reflector and the beryllium island has been proposed by NDA.* The fuel is cooled by a stream of sodium, A program of critical experiments has been initi- ated, in collaboration with NDA, which is basically similar to the recently completed ORNL program, 4CCR-2: A Compact Core Reactor for Aircraft Pro- pulsion, NY0-3080 (July 30, 1954). The fuel is formed by alternating lamince of uranivm foil and sheets of stainless steel and aluminum, the latter to represent the sodium of the reactor., The core region is essentially a 20.125.in.-long cylindrical annulus 10 in. in inside diameter and 4.312 in. in width. The central 17.25-in, section of the annulus contains the vranium, (The bounds of the annulus are actually octagonal, rather than circular.) A preliminary loading indicates that the predicted critical mass has been significantly underestimated, A 35% increase in the uranium loading and a 23% de- crease in the steel content of the core were necessary in order to make the assembly critical, The core now contains 31 kg of U235, 73 mfi'v"'-"fl.*u v ORNL—LR—DWG 11032 EFFECTIVE 1] [ CONTROL ROD POSITION, || f} / J O BERYLLIUM ; REFLECTOR FUEL ANNULUS SAMPLE A URANIUM FOILS IN INCONEL TUBE o BERYLLIUM 9 ISLAND ; SAMPLE A 8 CONTROL ROD THIMBLE . I SAMPLE B 1 6 ' RELATIVE ACTIVITY (arbitrary units) 2 3 4 5 6 0 J DISTANCE FROM INNER EDGE BORON - COPPER OF FUEL ANNULUS (in.) <<(/E\\T %/// —ows 10115“”]“ . Fig. 3.8. Relative Fission-Rate Distributions Across Fuel Annulus of High-Temperature Reflector- Moderated-Reactor Critical Assembly. 74 Part || MATERIALS RESEARCH e 4. CHEMISTRY OF REACTOR MATERIALS W. R. Grimes Materials Chemistry Division Phase equilibrium studies were made of the sys. tems ZrF UF ,, LiF-UF , NaF-LiF-UF , KF-UF NaF-KF-ZrF,, NaF-LiF-BeF, NaF.LiF-BeF,- UF ,, KF-BeF,, and NaF-KF-BeF , and of systems containing alkaline-earth fluorides, Additional work was done in investigating the equilibrium reduction of FeF, by hydrogen in NaZ:F , the reduction of UF , by structural metals, the stability and solubility of chromium fluerides in various molten fluorides, the reaction of UF with alkali fluorides, and the reduction of alkali fluorides by uranium metal. A method for preparing pure CrF ; was devised, and x-ray and petrographic data were obtained for compounds of ZrF , with CrF,, NiF,, and FeF ,. Fuel purification and production research in- cluded further study of methods for the recovery of contaminated fuel for re-use and the preparation of special materials. Methods for evaluating raw ma- terials were developed and used for evaluating material from an outside vendor., Adequate supplies of ZrF, are now being obtained from a commercial source, Difficulties with vacuum pumps and reactor cans used in production processes were resolved. Methods have been devised for obtaining relative measurements of viscosity and density of molten salts without removing the pure material from the conventional equipment used for fuel purification, Measurements of electromotive forces of cells in which molten NaF.ZrF, is used as the solvent were continued, and optical and x-ray data were obtained for various compounds in fluoride sys- tems. PHASE EQUILIBRIUM STUDIES C. J. Barton R. E. Moore F. F. Blankenship R. E. Thoma Materials Chemistry Division H. Insley, Consultant The System Z¢F .UF, R. P. Metcalf Materials Chemistry Division The simple phase diagram for the system Z¢F - UF, is presented in Fig. 4.1. Solid solutions are the only crystalline phases which have been ob- served in this system.! Examination of quenched samples by x-ray diffraction and with the petro- graphic microscope shows that the compositions of the solid solutions vary continuously over the entire composition range. There is no appreciable miscibility gap in this system. The System LiF-UF, R. E. Moore Materials Chemistry Division A phase diagram based almost entirely on ther- mal analysis was presented previously? for the LiF-UF, system. A detailed investigation of this binary system has now been completed, and a re- vised diagram, based on petrographic and x-ray diffraction examination of quenched samples, as well as thermal data, is shown in Fig. 4.2. A summary of the data obtained from the quench. ing experiments is shown in Table 4,1. The samples from which these data were taken were equilibrated for 16 to 54 hr at temperature before being quenched. Data obtained by thermal analy- sis were used to determine the liquidus curves from 0 to 23 and 60 to 100 mole % UFA. A compound shown by petrographic examination to be the result of quench growth was found to have an x-ray diffraction pattern which did not cor- respond to the patterns of any of the established compounds in the system. This compound was found above the solidus temperature in mixtures containing 20 to 25 mole % UF,. A composition and a temperature range for its stable existence could not be found. It appears to be either a metastable compound, a compound stable only at a relatively low temperature, or a high-temperature modification of Li , UF, with a very namow tem- perature range of stability. 'R. P. Metcalf and R. E. Thoma, ANP Quarn Prog. Rep. Sept. 10, 1955, ORNL-1947, p 67. 2R, E. Moore et al., ANP Quar. Prog. Rep. Dec. 10, 1950, ORNL-919, Fig. 10.4, p 245. 77 ANP PROJECT PROGRESS REPORT 1050 ORNL—LR-DWG 11524 t0C0 950 300 850 TEMPERATURE (°C) 800 f——— 750 | 700 ZrF4 {mole %) Fig. 4.1. Simple Phase Diagram for the System IrF ,-UF . The System NaF.LiF-UF H. A. Friedman R. E. Meadows B. A. Soderberg Materials Chemistry Division R. E. Cleary H. Davis Pratt & Whitney Aircraft There does not appear to be a mixture in the NaF-LiF-UF , system that would be a suitable fuel for a circulating-fuel reactor. No mixtures that contain less than 25 mole % UF , have been found that have liquidus temperatures in the region of 500°C. A ternary compound with more than 50 mole % UF, has appeared sporadically in slowly cooled melts, but in a preliminary survey made with the use of quenching methods the compound has not been found. It probably melts incongruently and has a limited stability range. Compatibility triangles in the NaF-LiF-UF ; sys- 78 tem are defined by the joins LiF-3NaF-UF , LiF- 2NaF-UF ,, LiF-7NaF.6UF ,, 4LiF.UF ;~7NaF-6UF,, 7LiF-6UF ,-7NaF-6UF ,, and 7NaF-6UF ~LiF.4UF ,. Of these joins, only the LiF-7NaF-6UF, is a quasi binary. The boundary curves and three- phase triangles are shown in Fig. 4.3. The dotted lines indicate tentative paths for the boundary curves. In order for the compatibility triangle 7NaF-6UF ,-7LiF-6UF 4LiF.UF , to be valid, the invariant point common to the primary phase field 7NaF-6UF ,, 7LiF.6UF ,, and 4LiF-UF , must be a eutectic or its composition must be displaced be- low the join 7NaF.6UF ,-4LiF.UF . The System KF.UF H. A. Friedman Materials Chemistry Division A preliminary phase diagram based entirely on data obtained from thermal analysis was presented e e PERIOD ENDING DECEMBER 10, 1955 UNCLASSIFIED ORNL-LR-DWG 10410 HOO T 1000 l - 4 900 Jfi—J* _ UF, + LIQUID 2N | o 800 — ] 2%\, / t._. . 2 Li,UFg + LIQUID | W 700 \ —t , J — a LiU,F,, + LIQUID wi = _ L 600 LiF +LIQUID '\ P T . LiU,F,> + UF %716%1 W 447 4 500 jL/ LIQUID _ Li'?UG F31 + LiU4Fi7 ] ) ) . © Li, UF L < LIF + LiUf | 4°'8 = U 400 3 Li7zUgF3 3| 3 LiF 10 20 30 40 50 60 70 80 90 UF, UF4 {mole %) Fig. 4.2. Phase Equilibrium Diagram for the System LiF-UF . previously? for the KFUF, system. This system was recently resexamined because of its possible similarity to the RbF-UF, system, and the results are presented in Fig. 4.4, In an early x-ray diffraction analysis of this system, Zachariasen? reported the presence of two compounds that contained more than 70% UF . No evidence has been found in this study for com- pounds containing more UF , than that indicated by the formula KF:2UF ,. In addition, the compound 7KF-6UF, is stable, whereas the compound KF-UF, apparently is not. The optical charac- teristics of 7KF-6UF, are very similar to those of the corresponding NaF and LiF compounds. 3. p. Blakely et al.,, ANP Quar. Prog. Rep. March 10, 1951, ANP-60, p 131. 4w, H. Zachariasen, Crystal Structure Studies in the Systems KF-UF4, KF-ThF4 and KF-LaF3, MDDC-1283 (Feb. 9, 1946; dec!. Sept. 5, 1947). The System NaF-KF.ZrF , H. A. Friedman B. A. Soderberg Materials Chemistry Division H. Davis Pratt & Whitney Aircraft Information on the complex ternary system Nak- KF-ZrF, has been obtained by thermal analyses and the careful examination of slowly cooled melts. The ternary compounds whose compositions are be- lieved to be certain are 2NaF:3KF.5ZrF, (con- gruent; mp, 432°C), 3NaF.3KF.2ZrF, (incongru- ent; mp, ~735°C), NaF-KF-ZrF, (incongruent; mp, ~493°C), and 3NaF-K F-27_rl:4 (congruent; mp, 593°C). In the compound 3NaF-3KF-2ZrF , the primary phase is 3KF.ZrF , and in the compound NaF.KF.ZrF, the primary phase is 3NaF.3KF. 2ZiF ,. A fifth ternary compound exists, but its composition has not yet been defined. |t appears 79 ANP PROJECT PROGRESS REPORT TABLE 4.1, SUMMARY OF QUENCHING DATA ON THE LiF-UF4 SYSTEM Phases Composition Liquidus Primary P eritectic Secondary Solidus Existing {mole % Temperature Phase Temperature Phase Temperature Below UF,) (°C) (<) o) Solidus 20 581 LiF 498 Li4U FB 498 Li4UF8 25 525 LiF 500 Li UFg 490 Li Uy LizUgF 3 28 503 Li7U6F3] Li4U F8 489 Lf4U FS 33.3 565 LijUFa, Li UFq 495 LiUFg LizUgF3y 37.5 603 Li7U6F3-| 40 612 Li7U6F3] 46.2 683 LiU F,, 605 LisU Fa 605 LiyUgF g 50 718 LiU,F 588 LigUgF s,y LIU4F~|7 60 804 UF4 775 Li4UFB 66.7 UF4 767 Li4UF8 75 >716 |..iU4F]7 >601 LTU4F]7 LizUgF 3 80 >716 LiU4F]7 >716 LiU4Fl7 to consist of approximately 24 mole % NaF, 33 at 395°C. The lowest melting eutectic below mole % KF, and 43 mole % ZrF . Petrographic and x-ray diffraction data show that the compatibility triangles in the system exist as shown in Fig. 4.5. The heavy lines indicate the well-established triangles. The remainder of the triangles (dotted lines) are much less certain be- cause of the inherent difficulty in obtaining valid data from slowly cooled melts when incongruently melting compounds are present. The true quasi binaries in the system are NcF-3KF-ZrF4 (mini- mum melting temperature, ~765°C at 40 mole % KF), 7NaF.6ZrF 4-2NaF 3KF-5ZrF , (minimum melt- ing temperature, ~430°C ot 28 mole % KF), 2NaF3KF.5ZrF -ZrF , (minimum melting tempera- ture, ~423°C at 27.5 mole % KF), and 2NaF.3KF. SZrF (KF.ZrF , (minimum melting temperature, ~425°C at 33 mole % KF). The entire region bee. tween 35 and 50 mole % ZrF , has liquidus tem- peratures between 395 and 500°C. The lowest melting region of the diagram is at 10 mole % NaF, 48 mole % KF, and 42 mole % ZrF ,, with a liquidus 80 35 mole % ZrF , is at 30 mole % NaF, 63 mole % KF, and 7 mole % ZrF, and has a liquidus at 684°C. A program of quenching experiments will be initiated in the near future to provide the basis for a final determination of the melting points and phase relations in this system. The System NaF-LiF-BeF, L.. M. Braicher R. J. Sheil B. H. Clampitt Materials Chemistry Division G. D. White Metallurgy Division Viscosity data reported in Sec. 7, ““Heat Trans- ter and Physical Properties,"” for the mixture NqF-LiF-BeF2 (63.5-7.5-29.0 mole %) and previ- ously reported data on the mixture NaF-LiF-BeF, (64-5-31 mole %)> when compared to the viscosity 55. 1. Cohen, ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, Table 7.3, p 157. t 900 750 PERIOD ENDING DECEMBER 10, 1955 g ORNL-LR-DWG 14525 UF, Li Uy Fi7 (945) TEMPERATURES ARE IN °C 800 EUTECTIC (680) Na; UgF3, 770 e SR 700 ' NGZUFG flv-‘h.h EUTECTIC Li; Ug Fzy LN EUTECTIC &) Y ‘A‘ - 500 RS 950°C NoF 650 LiF Fig. 4.3. Boundary Curves and Three-Phase Triangles in the NuF-LiF-UF4 System, of the binary mixture NaF-BeF, (69.8-30.2 mole %)° show that ternary mixtures containing a small amount of LiF and approximately 30 mole % BeF, have a higher viscosity than that of the NaF-BeF, mixture with the same BeF, concentration. These data caused a shift in interest to the regions of lower BeF, concentrations along the join between the LiF-NaF eutectic (60-40 mole %) and the LiF-Na,BeF , eutectic (16 mole % LiF). Thermal- analysis data obtained by adding NaF or Li,BeF, to several compositions along this join showed that it is close to a drainage valley between the two eutectics. Viscosity measurements (Sec. 7, “‘Heat Transfer and Physical Properties’) were made on one mixture along this join, NaF-LiF- BeF, (53-24-23 mole %). The kinematic viscosity found at 600°C was lower than that of any previ- ously measured BeF ,-containing melt, except the NaF-BeF, (69.8-30.2 mole %) mixture. However, the small improvement in viscosity obtained by reducing the BeF, concentration in the ternary system to 23 mole % suggests that the point of diminishing returns has been reached in this ap- proach. This conclusion will be tested with one more composition on the same join, that is, with a mixture containing only 15 mole % BeF . The gradient quenching technique has been ap- plied to the study of two compounds reported to be 81 8 ORNL-LR-DWG 441526 1000 N 900 \ N I - ] 2KF-UF, + LIQUID UF, + LIQUID 3KF - UF, 3KF - UF, 7TKF-8UF, +LIQUID } 8OO +LIQUID [+LIQUID — — —— — — KF+ ~—1—— KF- 2UF, + LIQUID ¢ 700 m%r—mi‘ B - — T L % a—2KF.UF4+ Q—ZKF'UF4+ = 3KF- UFy - 7KF-8UF, ax [TE] o 2 600 -y - — —_t [ o KF:-2UF,+ 5 7KF-6UF, KF-2UFs+ UF, KF+ 3KF -UF, o ® 500 — | B-2KF-UF,+ J TKF-6UF, 7 L Ly 35 B-2KF-UF, + 5 2 & 3KF-UF, |- an_ ; L N M~ 400 - “JT$*‘“ - 2KF-UF, + Y~ 2KF-UF, + et 4 3KE-UF, 2 FKF-BUF, 300 + T | KF 20 30 40 50 60 70 80 90 UF, UF4(m0|eWO) Fig, 4.4. Phase Equilibrium Diagram of the KF-UF , System, ¥y e L¥0dIY SS3¥00¥d LIO3r0¥d dNV in this system. Quenches of the Na,LiBe,F, com- position showed only the crystaliine phase desig- nated by the formula. Further work will be re- quired to determine the liquidus temperature for this composition, which undercools strongly, and to demonstrate that it melts congruently, as is presently believed. A number of quenches have been performed with a mixture having a composi- tion corresponding to the compound Na,LiBe F,, reported by John,® Quenches covering the tem- UNCLASSIFIED ORNL—LR—-DWG 11527 - Rt “=-\2KF-Zrfy -~ Fig. 4.5. Compatibility Triangles in the NaF-KF- ZrF , System, PERIOD ENDING DECEMBER 10, 1955 perature range 296 to 620°C have shown only isotropic materials above 339 + 5°C and a mixture of phases, including LiF and an unidentified biaxial-positive crystalline phase, at lower tem- peratures, Further studies are being made in order to determine the composition and melting behavior of the unidentified compound. The System NaF-LiF-BeF -UF, L. M. Bratcher R. J. Sheil B. H. Clampitt Materials Chemistry Division G. D. White Metallurgy Division Investigations of mixtures in the NaF-LiF-Be F,- UF, system containing 3.0 mole % UF, and low quantities of BeF, were continued. These mix- tures are considered to be promising from the vis- cosity standpoint. Thermal-analysis and filtration techniques are being used in this investigation. The available data, which are summarized in Table 4.2, demonstrate that it is possible to ob- tain mixtures in this system containing 3.0 mole % UF, and 20 to 22 mole % BeF, that have liquidus temperatures of less than 505°C. Investigation of this system is continuing in an effort to determine the best fuel composition from the standpoint of both viscosity and liquidus temperature. W, Jahn, Z. anorg. w allgem. Chem. 277, 274 (1954). TABLE 4.2, LIQUIDUS TEMPERATURES IN THE SYSTEM NoF.LiF-BeF -UF , Composition of Mixture (mole %) Techniae Used T::p:ii:jm NaF LiF Ber UI:4 (°C) 61.6 7.3 28.1 3.0 Filtration 520< T < 525 54.3 15.5 27.2 3.0 Filtration 476 < T <520 53.3 22.3 21.4 3.0 Filtration <494 53.3 22,3 21.4 3.0 Filtration 440 < T < 474 56.9 17.8 22.3 3.0 Fittration 475 < T <500 55.6 21.0 20.0 34 Filtration 505< T <515 55.8 21.0 20.2 3.0 Thermal analysis 502 58.7 17.0 21.3 3.0 Thermal analysis 495 51.4 23.3 22.3 3.0 Thermal analysis 530 56.7 19.3 21.0 3.0 Thermal analysis 545 ANP PROJECT PROGRESS REPORT The System KF-BeF, L. M. Bratcher R. J. Sheil B. H. Clampitt Materials Chemistry Division G. D. White Metallurgy Division The existence of two compounds in the KF-BeF, system was reported ;::reviously:7 namely, KzBeF4 and KBeF3. melts Further studies of slowly cooled have demonstrated the existence of two additional compounds, which are believed to have the formulas K;BeF . and KBe,F .. It appears from the available thermal-analysis data that K BeF, melts congruently at about 738°C and that the KsBeFS-KZBeF4 eutectic, which melts at 730°C, contains approximately 27.5 mole % BeF,. Mix- tures containing 66%3 mole % BeF, showed only one thermal effect on cooling curves at about 325°C, but it is not clear whether this compound The basis is that mixtures melts congruently or incongruently. for the compound formulation of this composition were homogeneous, well- crystallized material having optical and x-ray diffraction properties that were different from those of other compounds in this system. The discovery of these compounds makes it desirable to continue the study of this system through use of the quenching technique, particularly in the region of the KBe,F. compound, where thermal analysis does not give reliable data. The System NoF-KF-BeF, L. M. Bratcher R. J. Sheil B. H. Clampitt Materials Chemistry Division G. D. White Metallurgy Division All four of the known KF-BeF, compounds have been observed in slowly cooled melts in the NaF-KF-BeF, system, in addition to the two NaF-BeF, compounds and one well-established 7¢. J. Barton et al., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 71. 84 ternary compound, NeKBeF,. It seems certain that one additional ternary compound exists in this system, and possibly there are others, but further work will be required to establish their identity. On the basis of the results obtained to date from petrographic and x-ray diffraction studies of slowly cooled melts, the following compatibility triangles are tentatively postulated: NaF-KF-K,BeF.; NaF-K,BeF.-K,BeF,; NaF- KzBeF4-NaKBeF4; KZBeF4-NoKBeF4-KBeF3; KBer-NoKBeF4-KBe2F5. On the join between KBe2F5 and NaBeF , pounds, the mixture containing 57.5 mole % BeF, which are crystalline com- solidified to a transparent glass with no notice- able thermal effect on the cooling curve. sitions on both sides of this mixture contained glass and the neighboring binary compound. Work Compo- on this system is continuing. Systems Containing Alkaline-Earth Fluorides L. M. Bratcher R. J. Sheil B. H. Clampitt Materials Chemistry Division G. D. White Metallurgy Division The alkaline-earth fluorides MgF, and CaF, have properties, such as low cross section, low vapor pressure, and high-temperature stability, that cause them to appear to be attractive for use as fuel carriers. However, consideration of these materials in the fused-salt research program was previously discouraged because the published data show only high-melting-point eutectics in alkali fluoride~MgF , and —CaF, systems. Long- range interest in materials for producing higher reactor temperatures than those presently being considered makes it desirable that pertinent data now lacking be obtained and that some of the published data be checked. Preliminary thermal- analysis data are presented in Table 4.3, together with the available published values. The data that have been obtained for the LiF-NaF-CaF, system are sufficient to locate the eutectic compo- sition at approximately 53-36-11 mole %, TABLE 4.3, EUTECTIC TEMPERATURES IN ALKALI FLUORIDE —MgF , AND ~CaF , SYSTEMS Eutectic Temperature (°C) System Literature ORNL Value Value LiF-CuF2 773 765 Nc:F-Cc:F2 810 812 N(:ll':-MgF2 830 825 LiF-NaF-Mg F2 630 620 LiF-MgFZ-Con 672 LiF-NoF-CuF2 616 Na F-Mng-COF:Z 750 LiF-NaF-MgF -CaF, 588 CHEMICAL REACTIONS IN MOLTEN SALTS L. G. Overholser F. F. Blankenship G. M. Watson Materials Chemistry Division R. F. Newton Research Director’s Division Equilibrium Reduction of FeF, by H, in NaZtF C. M, Blood Materials Chemistry Division Numerous experiments have been performed, as reported in previous reports in this series, to study the reaction FeF, (d) + H, (g) &=Fe (s) + 2HF (g) with a molten mixture of NaF-ZrF , (53-47 mole %) used as the solvent, The apparent equilibrium constants (mole fractions of dissolved species assumed as activities) previously determined had, unfortunately, only qualitative significance be- cause of the lack of an adequate method of ap- proaching equilibrium, With the present experi- mental assembly and technique, however, equilibrium can be approached effectively from either the forward or the reverse direction of the reaction. Furthermore, equilibration can be continued under constant conditions for periods as long as several weeks, that is, until it is apparent that the compo- sition of the system is constant within the limita- PERIOD ENDING DECEMBER 10, 1955 tion of analytical procedure and inherent experi- mental variations, The present experimental assembly is, in general, similar to the one described previously for use with the equilibration method.® However, the limited-capacity HF generator used previously has been replaced by a practically unlimited source of a gaseous mixture of HF and H.,. The method being used consists in bubbling a gaseous mixture of H, and HF of constant composition through the melt containing FeF,, which is contained in a mild-steel-lined nickel reactor, The equilibration of the melt is continued until the concentration of FeF, is constant over a period of several days, the remaining variables being held constant. Periodically, liquid samples are withdrawn through sintered-nicke| filters for chemical analysis. The compositions of the reactor influent and effluent gas streams are determined by continuously bubbling accurately metered portions of these gases through standard KOH solution, In the earlier groups of equilibration experiments, the attainment of equilibrium was judged by the approximate equilization of composition of influent and effluent gas streams of the reactor, This criterion has been found to lack sufficient sensi- tivity for better-than-qualitative estimates. This is particularly true if the equilibrium is approached from the reverse direction of the reaction. The modified source of HF consists of a 10-1b cylinder of liquid HF placed in a thermostat in which the temperature is controlled to approxi- mately £0.03°C. The tank is connected to a mixing chamber through a copper tube held at a tempera- ture (80°C) considerably higher than the tempera- ture of the liquid HF eylinder. The HF vapor on its way to the mixing chamber is made to diffuse through a sufficient number of sintered-nickel barriers (0.0004-in. pore size) to reduce the volu- metric flow rate to an experimentally suitable value and to stabilize the flow rate at this value. The streams of H, and HF are fed to the mixing chamber, which is also held at constant tempera- ture (340°C). The volumetric flow rate of the hydrogen is precisely controlled by a combination of commercially available gas-flow regulators and is further stabilized by ditfusion through porous barriers. The composition of the resulting gas mixture can be varied at will by changing the H2 BC. M. Blood and G. M. Watson, ANP Quar, Prog. Rep. Sept. 10, 1954, ORNL-1771, p 66. 85 ANP PROJECT PROGRESS REPORT flow rate or the temperature of the thermostat en- closing the HF cylinder, These compositions can be held constant for indefinitely long periods with an arithmetic mean deviation of approximately 14%, The current experiment was started during the week ending August 19, 1955, and the apparatus has been operating continuously and without inci- dent to the present time. Measurements have been made at 800 and at 700°C, The iron concentration range studied at 800°C was approximately 300 to 1300 ppm. At 700°C the range studied was approxi- mately 3300 to 2000 ppm. Additional measurements at 700°C are now being made in the range from 1000 to 600 ppm. A summary of the experimental results obtained to date is given in Table 4.4, As mentioned previously, it was found experi- mentally that several days of equilibration time were required before constancy of the FeF, con- centration was attained. This was particularly true when sizable changes in the partial pressure of HF were imposed on the system, During the transition periods, the changes in composition of the effluent gas stream were rather minute, This may be explained satisfactorily by comparing the extremely slow rate of the reaction (in the order of several days) with the swift passage of gas bubbles through the melt (in the order of seconds). For the calculations of the mele fraction of FeF in solution, the total number of moles in the quuié mixture was assumed to be the same as the number of moles of constituents added (10.0 moles per kilogram of mixture for this composition). The partial pressures of HF were calculated from con- centration values, with the assumption that HF was monomeri¢ at 700 and at 800°C. The experimentally determined equilibrium con- stants obtained to date are not far different from those that may be calculated from available tabu- lations? of free energies of formation, that is, AF® for FeF, (s) and HF (g). By comparison of the calculated and the experimental constants, the activity coefficients of the dissolved FeF may also be easily obtained. These calculations will be made when the experiment has been com- pleted. L. Brewer et al.,, pp 65, 110 in The Chemistry and Metallurgy of Miscellaneous Materials, Thermodynamics, ed. by L. L. Quill, McGraw-Hill, New York, 1950. 86 Reduction of UF, by Structural Metals J. D. Redman Materials Chemistry Division The apparatus and methods used to measure the reduction of UF, by metallic chromium or iron in molten fluorides have been described in previous reports in this series, The data obtained for these reactions with NaF-ZrF , (50-50 mole %), NaF-LiF- KF (11.5-46.5-42.0 mole %), NaF-ZrF , (53-47 mole %), or Nc:F-LiF-ZrF4 (22-55-23 mole %) as reaction media have also been reported. During the past quarter, studies have been made on the reduction of UF, by various structural metals, other than chromium or iron, in molten NaF-LiF-KF (11.5- 46.5-42.0 mole %). The earlier work showed that chromium-containing alloys are not stable at 800°C when in contact with UF4 dissolved in the NaF- LiF-KF eutectic, and the possibility of replacing the chromium with some less-active metal was therefore considered, The other structural metals being studied are niobium, tantalum, vanadium, and tungsten. Molybdenum, which was studied previously, was found to be more stable than chromium, The results of the studies on the reduction of UF4 by vanadium at 600 and 800°C with NaF-LiF- KF (11.5-46.5-42.0 mole %) as the reaction medium are given in Table 4.5, In all the experiments 2 g of vanadium was reacted with UF:4 (15 wt %, 2.3 mole %) dissolved in approximately 20 g of the NaF-LiF-KF mixture contained in nickel. As may be seen from the data presented in Table 4.5, the vanadium concentrations found in the filtrates were extremely low. In fact, these values are the lowest found for any metal in equilibrium with UF4 in the NaF-LLiF-KF mixture at these temperatures, with the possible exception of nickel. These results indicate that vanadium metal is stable under these conditions if it is assumed that the vanadium fluoride formed is not appreciably vola- tile. This assumption appears to be valid in view of the extremely low concentrations of vanadium fluoride present and the large excess of alkali fluorides available to complex any metal fluoride formed. Data for the reduction of UF, by niobium at 600 and 800°C with NaF-LiF-KF (11.5-46.5-42.0 mole %) as the reaction medium are given in Table 4.6. In these experiments, as in the experiments with vanadium, 2 g of niobium was reacted with 15 wt % PERIOD ENDING DECEMBER 10, 1955 TABLE 4.4, APPARENT EQUILIBRIUM CONSTANTS FOR REDUCTION OF FeF2 BY H, IN NaF-ZrF, (53-47 mole %) Reaction: FeF2 (dy + H2 (g) = Fe (s) + 2HF (g) Average total pressure: 740 mm Hg Initial charge: 6.0 kg of NaF-ZrF4 and 30,0 g of FeF2 Container wall: mild steel (98% Fe) Partial Pressure Sampling Time Fe in Melt of HF in Y. Sy B - (days at constant K_* (ppm) Effluent Gas * temperature) 2 (atm x 104) At 800°C 470 7 4.28 2.3 370 8 3.92 2.5 335 10 3,53 2,2 895 18 5.70 2,2 830 19 5.70 2.4 1295 21 7.74 2.9 960 25 6.00 2,3 985 26 6.27 2.5 965 27 6.17 2.4 1015 28 6.35 2.4 955 34 6.65 2.9 1005 35 6.50 2.6 1025 36 6.55 2,6 Av 2,510.2 At 700°C 2630 4 5.43 0.69 2810 5 5.55 0.66 2880 6 5.63 0.67 3070 10 5.89 0.69 3160 11 5.82 0.66 3205 12 5.87 0.66 3285 13 5.83 0.63 3232 14 5.63 0.60 2220 19 4.88 0.64 2525 20 4,99 0.60 2430 21 4.87 0.59 2190 24 4.84 0.64 2285 25 5.02 0.67 2335 26 4.81 0.61 2390 27 4.83 0.59 Av 0.64 1 0.04 PO *Kx = PaF/XFerpHZ’ where X is mole fraction and P is pressure in atmosphere. 87 ANP PROJECT PROGRESS REPORT TABLE 4,5, EQUILIBRIUM DATA FOR THE REACTION OF UF, WITH VANADIUM IN MOLTEN NoF-LiF-KF (11.5-46.5-42.0 mole %) AT 600 AND 800°C Conditions of Found in Filtrat Equilibration ound n Tiitrate TABLE 4.6. EQUILIBRIUM DATA FOR THE REACTION OF UF , WITH NIOBIUM IN MOLTEN NaF-LiF-KF (11,5-46.5-42.0 mole %) AT 600 AND 800°C Conditions of . Found in Filtrate Equilibration Temperatore Time Total Total Total Temperature Time Total Total Total °C) (hr) Uranium Vanadivm* Nickel ©C) (hr) Uranium Niobium* Nickel (wt %) (ppm) {ppm} (wt %) (ppm) (ppm) 600 3 10.7 30 1 4600 3 11.2 675 20 3 10.8 15 1 3 11.3 725 65 5 10.6 25 35 5 11.1 655 1 5 10.7 15 1 5 11.2 640 6 800 3 10.4 25 i 800 3 10.8 1090 35 3 10.5 35 i 3 10.5 1200 15 5 10.3 25 20 5 1.1 1540 6 5 10.6 35 20 5 11.2 1580 40 *Blank of 15 ppm of vanadium at 800°C. 12 1.0 1830 12 10.9 1470 of UF, (2.3 mole %) dissolved in approximately 12 11.0 1930 20 g of the NaF-LiF-KF mixture contained in nickel. 12 11 1310 As may be seen from the data presented in ' Table 4.6, niobium is not stable under these con- 12 11.4 1230 ditions. Equilibrium conditions are reached within 12 12.0 1360 3 hr at 600°C and after a somewhat longer period at 800°C. The values given for the 12-hr equili- bration periods are not so precise as desired, but it appears to be safe to conclude that no significant increase in niobium concentration occurs in going from a 5- to a 12-hr reaction period, The reaction involved is quite temperature sensi- tive, as evidenced by the doubling of the niobium concentration when the temperature was raised from 600 to 800°C. This suggests that mass transfer of niobium would be very likely to occur in a circulating fuel based on the NaF-KF-LiF eutectic. No further work on this metal-fuel system is contemplated, since all the data collected indi- cate very strongly that niobium and UF, are not compatible in this solvent, Studies have been made on the UF ,-tantalum system, but no analytical data are available at present. Preliminary results for the UF -tungsten system suggest that tungsten probably behaves in a manner comparable to that reported for niobium. 10), D, Redman and C. F. Weaver, ANP Quar. Prog. Rep. Sept, 10, 1955, ORNL-1947, p 75. 88 *Blank of 315 ppm of niobium at 800°C. Stability and Solubility of Chromium Fluorides in Yarious Molten Fluorides J. D. Redman Materials Chemistry Division The results of studies on the stability and solu- bility of CrF,, in NaF-ZrF , (53-47 mole %) at 600 and 800°C were reported previously, 1% and a more exhaustive study of this system has now been made in which the CrF, was purer than it was in the earlier studies. The effect of varying the ZrF -to-CrF, ratio was investigated at 600 and 700°C, since it was suspected that the solid phase present was not CrF, but, rather, CrFyZrF . I this was the case, some variations in the zirconium- to-sodium ratio should be observable in the fil- trates. The resvits of the recent experiments, which were carried out in nickel equipment with the proper quantity of CrF, added to approximately —— - 40 g of the NaF-ZrF , mixture and then equilibrated for 5 hr, are given in Table 4.7, An examination of the data presented in Table 4,7 shows the marked effect that the change in the zirconium-to-chromium ratio has on the solu- bility of CrF,, particularly at 600°C. The increase in the solubility of the CrF, as the zirconium-to- chromium ratio decreases must be associated with the separation of the solid phase GrF,-ZrF . If excess CrF, is added to this system, ZrF, is removed, and the solvent is enriched in NaF, which, in turn, increases the solubility of CrF ., Evidence for this postulated process is found in a comparison of the zirconium-to-sodiumratios present in the filtrates with the ratio of 0.88 of the charge material, All values for the zirconium-to-sodium ratios in the filtrates are considerably lower than the ratio in the starting material. Unfortunately, the chromium values found for the tests at 600°C in which small amounts of CrF2 were added are not precise, and therefore those tests will be rerun, PERIOD ENDING DECEMBER 10, 1955 However, the results indicate that the zirconium- to-chromium ratio is not critical at the higher CrF, values. A comparison of the Cr** values with the total chromium values shows that at least 90% of the chromium is present as Cr**, and there- fore the disproportionation of CrF, cannot be an important factor in this solvent, Data relating to the stability and solubility of CrF, in NaF-LiF-KF (11.5-46.5-42.0 mole %) at 400 and 800°C were also presented previously, 1011 Since the values were not in agreement, additional experiments have been performed in a further attempt to find reliable values. A purer batch of CrF, and a different batch of solvent were used in the tests reported in Table 4.8. The values given in Table 4.8 for the tests at 600°C are in agreement with the values reported for one of the eariier experiments. However, the values found at 800°C are approximately one-third ”J. D. Redman and C, F. Weaver, ANP Quar. Prog. Rep. Dec. 10, 1954, ORNL-1816, p 64. TABLE 4,7. SOLUBILITY AND STABILITY OF CrF2 IN MOLTEN NuF-ZrF4 (53-47 mole %) Equilibration time: 5 hr Conditions of Equilibration Found in Filtrate +4 Temperature CrF2 Added Zr-to-Cr Cr Cr Zr-to-Na (°C) (wt % Cr) Ratio* (wt %) {wt %) Ratio* 600 1.4 17 0.79 0.67 0.80 1.4 17 0.65 1.37 0.74 4.0 5.8 0.80 1.00 0.74 4,0 5.8 0.70 0.67 0.75 6.4 3.4 3.5 3.5 0.71 6.4 3.4 3.4 3.6 0.70 9.6 2.1 5.0 5.2 0.69 9.6 2.1 5.6 5.9 0.48 700 4.0 5.8 3.6 3.5 0.76 4,0 5.8 3.4 3.6 0.77 6.4 3.4 5.4 5.8 0.76 6.4 3.4 5.5 5.7 0.77 6.4** 3.4 5.7 5.8 0.76 B.4** 3.4 5.6 5.7 0.80 9.6 2.1 8.1 8.8 0.82 9.6 2,1 8.0 8.7 0.80 800 9.6 2.1 8.1 8.3 0.79 9.6 2.1 8.2 8.2 0.78 *Ratio calculated from mole fractions. **Equilibration time, 12 hr. 89 ANP PROJECT PROGRESS REPORT TABLE 4.8, SOLUBILITY AND STABILITY OF CtF 5 IN MOLTEN NoF-LiF-KF (11.5-46.5+42,0 mole %) Conditions of Equilibration Found in Filtrate Temperature Time CrF; Added Cr++ Cr Ni (°C) (hr) (wt % Cr) (ppm) (wt %) (ppm) 600 5 2.5 1 0.19 70 5 2.5 1 0.19 65 12 2.5 145 0.24 145 12 2.5 1 0.23 145 700 5 5.0 50 0.90 50 5 5.0 55 1.1 60 12 5.0 45 1.1 35 12 5.0 50 0.8 50 800 5 5.0 475 1.1 280 5 5.0 1 1.2 110 12 5.0 ] 1.1 240 12 5.0 1 1.3 145 as large as those found in the two earlier experi- ments. No values had been determined previously at 700°C, and hence no comparison at this tem- perature can be made. The lack of agreement be- tween values obtained at 800°C cannot be satis- factorily explained at this time, Variations in temperature can hardly account for differences of this magnitude, and it appears to be unlikely that variations in the different batches of CrF, could be responsible, Reaction of UF3 with Alkali Fluorides B. H. Clampitt C. J. Barton Materials Chemistry Division Attempts to measure solubilities of UF in molten fluorides have been complicated by the presence of variable and unexplained quantities of U4* in the filtered specimens. While the UF3 available con- tains some UF,, the amount of U4* found after the tests often greatly exceeds the amount present from this source. This tetravalent uranium ap- parently arises from two reactions, which appear to be independent. The reactions are the reduction of one of the melt constituents by UF_ and the disproportionation of UF, to UF, and uranium. The disproportionation reaction occurs both during equilibration and during filtiration of the melts. Experiments made previously showed that solu- tions of UF, in molten alkali fluorides were more 90 stable in copper than in nickel containers, and therefore some studies of UF3 mixtures with indi- vidual alkali fluorides were carried out in copper apparatus, The results obtained (Table 4.9) indi- cate that UF, is completely stable at 900°C in molten LiF contained in copper. However, both NaF and KF yield the corresponding alkali metal when heated with UF . under these conditions. The amount of UF , in the melt after the heating period is greater than that to be expected from the equation UF, + MF = UF, + M° in either NaF or KF solution. However, both the amount of alkali metal produced and the discrepancy between the UF, and the free alkali-metal con- tents are much larger when KF is the solvent fluo- ride than when it is NaF, The effect of pressure on the reaction of UF3 with KF has been examined by heating mixtures of the NaF-LiF-KF eutectic (15 g) with UF, (6 g) at 650°C in copper crucibles for 2 hr under helium atmospheres at total pressures ranging from 760 to 0.04 mm Hg and analyzing the melt to determine the U3* concentration. The data obtained are shown in Fig. 4.6. At total pressures above 50 mm Hg the U3% content remains essentially constant at 12.4 wt %, about 55% of the original value. However, when the total pressure is below 50 mm Hg, the U3* concentration decreases rapidly e - A e - PERIOD ENDING DECEMBER 10, 1955 TABLE 4.9. REACTION OF UF3 WITH ALKALI FLUGRIDES AT 900°C IN COPPER Originloll Filtrate Analysis (wt %) Alkali Metal U4% in Filtrate Composition 3T Produced (meq) (mole %) u Total U (meq) LiF-UF3 (73-27) 55.5 64.1 None None NaF-UF3 (71-29) 41.6 58.2 4,65 5.9 KF-UF3 (85-15) 3.24 34.96 15.4 30.5 *Calculated by difference from filtrate analysis and corrected for U4t in original - ORNL-LR-DWG 11528 S M T fl,. s 7,7!’7 ;"777 Y %IL _L) — L — d | - T T 3 [ 4o e ol {00 200 300 400 200 600 700 800 900 HELIUM PRESSURE (mm Hg) Q RESIDUAL ud+ {wt %)} AFTER EQUILIBRATION FOR 2 hr » C | | Fig. 4.6. Effect of Pressure on Reaction Between UF3 and KF in NaF.LiF-KF Eutectic at 650°C, with decreasing pressure. This behavior suggests that at low pressures the potassium metal escapes from the liquid system quite rapidly, and the reaction UF, + KF = K° + UF, proceeds to completion. Under these circum- stances the melts show colors ranging from red to olive, depending on the UF3 and UF, concen- trations. However, when potassium metal is added to the NaF-LiF-KF eutectic, the liquid becomes an opaque, turquoise green and shows a metallic luster. The green liquid is probably potassium metal saturated with the salt mixture, U3+, Reduction of Alkali Fluorides by Uranium Metal C. J. Barton B. H. Clampitt Materials Chemistry Division An investigation of uranium metal as a reducing agent for alkali fluorides was initiated because the disproportionation equilibrium for UF, involves uranium metal, In the previous report experiments were described in which KF and NaF in the NaoF-LiF-KF eutectic were reduced by uranium metal at 650°C and NaF in the NaF-LiF eutectic was similarly reduced at 570°C. The results of additional experiments with the individual fluo- rides are presented in Table 4.10. In each ex- periment, 3 g of uranium was placed in 20 g of the fluoride, and filtered samples were withdrawn for analysis after equilibration with uranium metal, The reaction of liquid NaF with uranium metal was not studied because of the high melting point of NaF. However, since there is no reaction between LiF and uranium metal at 920°C and, yet, the LiF-NaF eutectic does react at 750°C, NaF must be reduced by uranium metal. The reaction of KF and uranium metal at 900°C pro- ceeded so rapidly that the gas exit lines were plugged after 1 hr, and the experiment had to be discontinued. The potassium collected in the exit line accounted for 60 to 70% of the uranium found in the filtrate. Probably not all the po- tassium was trapped in the exit line, Experimental Preparation of Pure Fluorides B. J. Sturm Materials Chemistry Division Preparation of Pure CrF3. —~ In the past, CrF, has been prepared by using CrF3-3‘/3H20 as the N ANP PROJECT PROGRESS REPORT TABLE 4.10, REACTION OF URANIUM METAL WITH LiF AND WITH KF " . Analysis of Filtrate Alkali Metal , Equilibration Alkali _ Temperature (wt %) Recovered as Fluoride Time Q) " Total OH™ (hr) U Tofd' U (meq) KF 0.75 900 0.14 14.01 26.2 KF 1.00 900 0.14 13.33 31.4 LiF 2.0 920 0.0 0.0* 0.0 *Not detectable colorimetrically. b ™ starting material, converting this to (NH )3C1'F3 NiF,‘,.ZrF4 by treatment with excess NH,HF,, and then decomposing this intermediate at 450°C to yield CrF,. The principal objection to this material is the high iron content of the hydrated chromic fluoride used as the starting material, Recently, the preparation of pure CrF_, has been ac- complished by hydrofluorination of sublimed an- hydrous CrCl, at 600°C. This CrF, will be used as the starting material for the preparation of CrF,; hydrogen reduction of CrF, at 750°C has been shown to be rapid and complete. Compounds of ZrF, with CrF,, NiF,, or FeF . - Earlier studies showed that complex fluorides of the type MFZ-ZrF4 were formed when ZrF, was fused at 950°C with equimolar quantities of CrF,, NiF,, or FeF,. The compound FeF,-ZrF, slowly decomposes when exposed to the atmosphere, but CrF,-ZrF, decomposes so rapidly that it is im- possible to obtain reliable x-ray or petrographic data if this material is handled in laboratory air. The compound NiF,.ZrF, is relatively stable under ordinary ofmospfiweric conditions, Mixtures corresponding to Cer-ZrF , Fer-ZrFu and NiF,_.ZrF, were fused in nickel capsules at 950°C, and the capsules were opened in a helium- filled dry box in an attempt to eliminate atmos- pheric contamination, Samples for x-ray exami- nation were sealed in a polystyrene holder by using dry mounting tissue to seal a strip of polystyrene film to the holder. Samples for petrographic examination were kept in sealed containers until they could be studied. The following optical and x-ray data were reported for the compounds protected in the manner de- scribed: 92 Optical data: Cubic; rather poorly crystallized; probably a single phase; refractive index = 1.442; pale yellow or green-yellow color X-ray data: Probably cubic with a few ex- traneous lines FeF..‘,-ZrF4 Optical data: Cubic; single phase; refractive index = 1.432; white or colorless X-ray data: Probably cubic; isomorphous with NiF,.ZrF, but slight shift of lines CtF,.ZrF, Optical data: Single phase; probably ortho- rhombic; average refractive index = 1.432; biaxial positive; 2V = 20 deg; birefringence = 0.008; pale gray or blue-gray color X-ray data: Pattern similar to that of Fer-ZrF4 with a slight shift of lines; some double lines present that suggest an orthorhombic structure that approaches a cubic form PRODUCTION OF PURIFIED FLUORIDES G. J. Nessle G. M, Watson Materials Chemistry Division Removal of CrF, from NaF-ZrF ,-UF, Mixtures F. L. Daley J. Truitt Materials Chemistry Division It was demonstrated previously that a purified 5-1b batch of Nt:F-ZrI=4-UF4 mixture contaminated with added CrF, could be processed to an ac- ceptable chromium content for use as fuel-testing material. The procedure for the chromium removal was found to require a treatment step that included reduction of the CrF, to chromium metal with the use of zirconium and then removal of the chromium metal by filtration, Additional experiments have now been carried out with 5-Ib batches of used fuel materials (NoF-ZrF4-UF4) having various amounts of CrF,, as well as oxidizing contaminants. In these experiments the fuel was equilibrated at 800°C for 3 hr with zirconium metal chips, filtered, and subjected to the standard hydroflucrination-hydro- genation treatment. In addition, two 50-lb batches of used material were treated under similar con- ditions. The data from these experiments, presented in Table 4.11, show that the history of the material is very important. Any handling of the fuel mixture that increases its oxide content may greatly reduce the effectiveness of the zirconium-reduction step. Reduction of U4" to U3* in the melt is controlled primarily by the amount of zirconium metal added in excess of the amount necessary to reduce the CrF2 to chromium. On this basis, a theoretical amount of reduction of U4* to U3t was calculated and compared with the experimental value so that the percentage of theoretical reduction obtained could be computed. It is significant that the percentage of theoretical reduction was found to decrease sharply with increased previous handling PERIOD ENDING DECEMBER 10, 1955 and exposure of the fuel to oxidizing conditions. When fuel material that contained essentially no oxidizing impurities (no previous handling after purification) was used, over 90% of the theoretical reduction was obtained. When used fuels that had been protected from exposure were processed, the observed reduction of UF ;| decreased over fivefold; a twentyfold decrease was obtained for a used fuel that had not been protected from air and water. Some rather interesting trends were noted. At a given temperature the resulting concentrations of CrF2 are inversely proportional to the square of the U3t/U4t ratio, within the experimental limitations. ~ The calculated values for K _ for the reaction CrI:2 + 2UF3F———'\ 2UF4 + Cr were 2 x 10=2 and 1 x 10~¢ at 800 and 650°C, respectively., The number of experiments was quite limited, however, and it is felt that no special significance should be attached to these values, Efforts to protect the filter from plugging by oxides and oxyfluorides included cooling the melt to 650°C before filtration. It was hoped that these oxycompounds would precipitate in the reactor at the lower temperature and thus be left there when the filtration was made. Comparisons of the TABLE 4,11, SUMMARY OF RESULTS OF REPROCESSING EXPERIMENTS CrF2 Filtration Concentration ae, 4t Percentage of Source of Material Temperature (ppm) u° ' /u Theoretical Kx °C) Reduction Before After 5.1b Batches Previously purified fuel + Cr F2 800 11,500 90 0.43 100 2x107° 11,500 1400 0.13 91 3 x 10> Used fuel stored in closed 650 2,500 440 0.045 17 1x10° 8 containers -6 500 345 0.057 14 1x10 Used fuel completely exposed 800 2,000 1540 5 to atmospheric contamination 50-1% Batches Used fuel stored in closed 800 270 190 containers 305 110 93 ANP PROJECT PROGRESS REPORT quantity of filter cake obtained at 800 and at 650°C did not indicate that any advantage was gained by lowering the temperature. From these data it can be concluded that CrF, removal is feasible on a small-scale basis if proper care is exercised during handling of the fuel mixture to prevent excessive oxide contamination. Results of the 50-lb-batch reprocessing experiments indi- cate that the treatment is, in general, feasible, but the materials studied thus far have been too low in chromium content for really definitive data to be obtained. Laboratory-Scale Purification Operations F. L. Daley F. W. Miles Materials Chemistry Division The standard hydrofluorination-hydrogenation purification procedure was used in the preparation of a 1.5-kg batch of LiF-NaF (60-40 mole %). No difficulties were encountered, and a satisfactory product was obtained. The transfer was nearly complete; only 24 g of the mixture remained in the reactor, Four batches of RbF(KF)-ZrF ,-UF, of approxi- mately 2 kg each were prepcreé 'Ighe first two RbF(KF)-ZrF ,-UF, mixtures had a nominal compo- sition of 48-48-4 mole %, and the last two, 50-46-4 mole %. Since the available RbF contained about 20 mole % KF, the compositions were calculated by using the average molecular weight of the RbF-KF mixture. Unusually large residues remained in the reactor in every case. The hydrofluorination times em- ployed were 2, 2 , 6, and 18 hr, respectively; in each case the composmon was cooled to 450°C before transfer. Slight thermal breaks were de- tected at ~ 540 to 565°C in the first three experi- ments. No thermal break was detected in this range on the fourth experiment. All four products froze at ~405 to 410°C. Oxyfluorides were present in the reactor residues from the first three experiments, despite the rather drastic hydro- fluorination treatments, Petrographic, x-ray, and chemical analyses showed the compositions of the first three products to be rather different from those of the residues. Consequently, these products were not issued for viscosity studies. The composition of the last product was found to be quite close to the theo- retical composition, and the product was therefore issued for viscosity and phase studies. 94 Special Preparation of NaF-ZrF ,-UF ,-UF F. L. Daley Materials Chemistry Division At the request of Wright Air Development Center, a 50-lb batch of chF-Zer-UF:;-UF4 was prepared by a procedure previously described.’? The nominal, final composition was NaF-ZrF -UF (44-53-3 mole %). A chemical analysis of the product gave the following: In weight percentages Found Theoretical Na 10.9 8.7 Zr 41.8 41.7 Total U 6.24 6.2 3+ U 5.15 F 41.1 43.4 In parts per million Fe 260 Cr 100 Ni 90 Evaluation of Raw Materials for Fuel Preparation F. L. Daley Materials Chemistry Division It has been shown that rather small amounts (~2 wt %) of oxides or oxyfluorides in the raw materials greatly increase the time required for purification. The presence of these impurities at low concentration is not, to date, readily detectable by either petrographic, x-ray, or chemical The more reliable index of purity appears to be the actual preparation of small quantities (3 kg) of fuels by using the raw ma- terials in question. Thus samples from a batch of ZrF and from a batch of NCIZI'FS submitted by oufs:de suppliers were evaluated by determining their processing characteristics. The ZrF, was found by petrographic examination to contain a very small amount of finely divided oxide or hydroxide, which was not found by x-ray diffraction and chemical analysis. The 3-kg charge of NaF-ZrF,-UF, (50-46-4 mole %) prepared with this material by the standard procedure was normal in freezing-point, chemical-composition, and optical properties, analyses. I‘2(:. M. Blood et al., ANP Quar. Prog. Rep. June 10, 1955, ORNL-1896, p 72. The sample of NaZrF showed a large quantity (20 to 25%) of a brownish material that was identified by petrographic examination as hydrated Na,Zr,F. .. No processing difficulties were en- countered with this but the product contained brownish aggregates that were tenta- tively identified as oxyfluorides. On the basis of these the commercially obtained ZrF material, results, appears to be satisfactory, while the Nt:erF5 seems to be marginal at best. Pilot-Scale Purification Operations C. R. Croft J. Truitt J. P. Blakely W. T. Ward Materials Chemistry Division Fluoride mixtures to be used in small-scale corrosion testing, phase equilibrium studies, and physical-property studies are prepared in the pilot- scale facility. During the past quarter, 68 batches totaling approximately 800 Ib were prepared. Production-Scale Operations J. E. Eorgan J. P. Blakely Materials Chemistry Division The shortage of ZrF,, which has prevailed for several months because of the low-capacity pro- duction of the Y-12 Area facility, apparently no longer exists. The first 2500-1b shipment of a 10,000-Ib order from General Chemical Company was received early in November, and chemical analysis showed this material to be satisfactory. Therefore the ZrF, supply is assured for the next six months, This material contains hafnium and will be used only in the production of fuels for component tests which do not involve nuclear activity. The vacuum pumps originally installed in the processing facility are beginning to wear out and have caused three major shutdowns or delays of approximately 48 hr each. The vacuum-pump repair shop has recommended that these pumps be re- placed because the parts are so worn that periodic trouble will occur. Spare parts are not available for these Beech-Russ pumps. Therefore they are to be replaced with Kinney pumps, for which parts are kept in stock, An increasing incidence of failures of reactor cans, which are fabricated from A-nickel, has led to a closer investigation of the cause of such failures. These failures consist of cracks that PERIOD ENDING DECEMBER 10, 1955 form in the walls of the reactor vessel during processing. Some reactors withstand the proc- essing of 12 batches before failing; others have failed while processing a second batch. Recently, the failures have been occurring most frequently after four or five batches have been processed. Originally, it was believed that sulfur attack of the nickel, which caused embrittlement, was the source of the trouble. The raw materials NaF, ZrF4, (NGF)4-ZrF4, and UF4 were therefore ex- amined, and new specifications were set for the sulfur content. The zirconium compounds were the worst offenders, with sulfur contents, some- times, of over 500 ppm. The new specifications called for a maximum sulfur content of 100 ppm. The incidence of failures tended to increase, however, rather than to decrease as was expected. Recently an ASTM specification was reviewed that recommends low-carbon nickel for high-temper- ature usage rather than A-nickel, which normally has a high carbon content. At high temperatures, carbonization occurs that causes fractures in the parent metal, With this mind, another examination of past failures was under- taken. Efforts to find evidence of sulfur attack failed, but o grayish-black crystal growth was found in the grain boundaries of the parent nickel where the fractures occurred; this tended to sub- stantiate the carbonization theory. Low-carbon nickel reactor cans have therefore been ordered. The delivery dates being quoted vary from three to six months. The long-expected increase in demand for proc- essed fluorides has failed to materialize. As a consequence the fuel stockpile is at an all-time high, and the supply of storage containers is nearly exhausted. Since no actual cancellation of orders has occurred, there is only a small surplus., However, if usage does not increase in the very near future, this facility will be shut down temporarily. A total of 5500 [b of fluoride compositions in 22 batches was processed during the quarter. information in Batching and Dispensing Operations F. A, Doss J. P. Blakely R. G. Wiley Materials Chemistry Division During the quarter, 85 batches totaling approxi- mately 3950 Ib of fluoride mixtures were dispensed 95 ANP PROJECT PROGRESS REPORT in batch sizes ranging from 1 to 250 lb. A material balance for the past quarter follows: u il Amount ateria (ib) On hand at beginning of quarter 6,650 Produced during quarter 5,479 Total 12,129 Dispensed during the quarter 3,939 On hand at end of quarter 8,190 Special Services F. A. Doss J. P. Blakely N. V. Smith W. T. Ward Materials Chemistry Division Filling, Draining, and Sampling Operations. Filling, draining, and sampling operations at ex- perimental facilities were at normal levels during the quarter. The fuel mixture was removed from the ART high-temperature critical experiment, for salvage, without difficulty. In-Pile Loops. — Two in-pile loops were filled with an enriched-uranium-bearing fluoride mixture during the quarter. Both loops were shipped to NRTS for testing in the MTR. A fourth in-pile loop is to be filled in January 1956. Testing of ART Enrichment Apparatus. — In order to obtain more information on the behavior of the proposed ART enricher system, additional tests are being made of the prototype apparatus used for the high-temperature critical experiment. Approximately 75% of the material originally charged into the enricher is still available in the equipment. Graphite cups will be used to catch specified increments from the enricher for weighing so that a thorough calibration of the enricher over a larger range of plunger levels can be made. A box has been fabricated to fit over the injection nozzle so that a protective atmosphere can be provided to prevent excessive oxidation of the nozzle and the fuel during these tests. FUNDAMENTAL CHEMISTRY OF FUSED SALTS Relative Yiscosity-Composition Studies of NdF-LiF-Zl’F4 Mixtures at 600°C G. M. Watson F.W. Miles Materials Chemistry Division Relative measurements of viscosity and density of molten salts can be obtained with the con- 96 ventional equipment used for fuel purification. The experimental assembly is such that changes in chemical composition of the molten mixtures can be effected with ease, and the viscosity and density measurements can be made without re- moval of the pure material from the container, Thus it is possible to determine whether signifi- cant changes in these physical properties occur with changes in chemical composition. When such changes are found, efforts can be made to correlate the observed changes with the for- mation or disappearance of probable ionic and molecular species in the liquid mixtures as the compositions are changed. Viscosities are being measured with a Brookfield viscometer. The only modifications required to the standard fuel purification assembly were the incorporation of a support for the viscometer and a slight change in the reactor top to permit easier introduction of the viscometer spindle, The viscometer assembly is calibrated before each measurement by using an identical reactor con- small taining a glycerine-water solution of known compo- sition. Since the measurements are to be relative, no correction is introduced to account for the effect of the thermal expansion of the spindle between the calibration temperature, 25°C, and the melt temperature. Measurements of viscosity on a number of mixtures of the same composition have indicated the precision to be better than +10%. The first study undertaken was an investigation of the relative viscosities of mixtures whose compositions lie along the join, on the phase diagram, between NaF-LiF-ZrF , (22.55-23 mole %) and Nc:F-ZrF4 (53-47 mole %). The experimental results, which are summarized in Table 4.12, do not indicate marked changes in viscosity due to composition changes. Some simple additions have been made to the existing experimental assembly so that density measurements can be made. The density measure- ments will be attempted by noting differences in the maximum hydrostatic heads required to blow bubbles of inert gas into the melt at two different depths. A milling-machine cross-feed, recovered from salvage, was permanently incorporated into the experimental assembly for introducing nickel tubing to the desired depth for bubbling the gas into the melt. The screw and micrometer vernier scale of the cross-feed were calibrated with a cathe- tometer. The hydrostatic heads are measured with TABLE 4.12. RELATIVE VISCOSITY vs COMPOSITION OF NaF-LiF-ZrFA MIXTURES AT 600°C Composition Relative Viscosity,* (mole %) / NaF LiF ZiF 7 22 55 23 1.0 25 50 25 1.0 31 39 30 1.0 37 28 35 0.94 44 16 40 0.94 50 5 45 0.97 53 0 47 1.0 M = wscoslty measurement for NaF-LiF-ZrF 4 (22-55- 23 mole %) = 12.0 centipoises at 600°C, as repor'red by S. 1. Cohen and T. N. Jones, Measurement of Viscosity o{ Compositions 81 and 82,0RNL CF-55-5-58 (May 16, 1955 PS5 a U-tube manometer that uses water as the mano- metric fluid. A cold trap was installed between the manometer and the reactor to protect the melt from water-vapor contamination. Preliminary density determinations have been made with apparently satisfactory precision. The apparatus will be calibrated in the near future with molten KCI and KCI-NaCl mixtures as standards. 3 The measurement of surface tension will be attempted by the maximum-bubble-pressure method, It is hoped that the same apparatus used to de- termine densities will be applicable with no substantial changes. EMF Measurements in Fused Salts L. E. Topol Materials Chemistry Division Measurements of electromotive forces of cells in which molten NoF-ZrF, (53-47 mole %) is used as the solvent were conhnued in order to obtain activities and activity coefficients of constituents of the melts. The measurements were made in the temperature range 550 to 700°C in a helium atmosphere. The melts were contained in re- crystallized alumina (Morganite) crucibles, and the half cells were connected by a bridge of porous ZrQ, impregnated with the NaF-ZrF , mixture. The cells measured were of the general type M|MF ,(a,), NaF(a,), ZrF (a,) || NaF(a]), ZrF (a3), PERIOD ENDING DECEMBER 10, 1955 where M and M“may be Cr, Fe, or Ni and M and M”* may be identical. Previously, saturated cells with identical saturating phases, MF2 ZrF and M'F ZrF ,, were measured, If it is cssumed that N0+ ond F = are the sole carriers of current, the net result of passage of two faradays of electricity is M+ ZrF (a,) + M'F - ZrF ,(s) + 2tNaF (a,) = MF,ZrF ,(s) + M"+ ZrF ,(a;) + 2tNaF(a7) where ¢ is the transference number of the Na* ion, If the amounts of added metal fluorides are ap- proximately equal and the solubilities of the complexes (MF ,-Z:F, and M°F ,-ZrF ) are low, the above equation reguces to M+ M'Fz-ZrF4(s) = MF,-ZrF (s) + M” , and the emf’s of these cells, together with the free energies of formation of the pure fluorides, indicate the relative stabilities of the solid complexes, In the cells reported in Tabie 4.13 the addition of metal fluorides was low enough to eliminate the presence of solid salt phases at high tempera- tures (600 to 700°C), and the net cell reaction was M + M'F2(aé) + 2tNaF(a,) = M’ + MF ,(a;) + 2:NaF(a;) With small additions of MF, and M‘F,, the activities of NaF on the two sndes are neuriy equal, and the Nernst equation for the reaction becomes RT = “3 RT = Y3*3 E=E°-—-—F n —=E° -~ — In ——, n aq nF V3% where a,a’, x,x°, and y,y’ are the activities, mole fractions, and activity coefficients, respectively. From Brewer's!4 estimates of free energies of formation, E° values of 0.33, 0.20, and 0.53 v are obtained for the Cr-Fe, Fe-Ni, and Cr-Ni cells; these values should be nearly independent of the temperature over this range. From these E° values 13E. R. Van Artsdalen and I. S. Yaffe, J. Chem. Phys. 59, 118 (1955). 14l... Brewer et al., p 107 in The Chemistry and Metallurgy of Miscellaneous Materials, Thermodynamics, ed, by L. L. Quill, McGraw-Hill, New York, 1950. M7F jag) M, 97 ANP PROJECT PROGRESS REPORT TABLE 4.13. POTENTIALS OF CELLS OF THE TYPE M MF ag), NaF(ay), ZrF (a,)|INaF(a}), ZiF (a3), M F (ad) [M" Measured emf< (v) Summation of Temperature Cr-Fe and Fe-Ni (°C) Cr-Fe® Fe-Ni¢ Cr-Ni? Valves (v) 550 0.325 0.420 0.751 0.745 600 0.340 0.394 0.745 0.734 650 0.341 0.399 0.749 0.740 700 0.337 0.409 0.753 0.746 %Mean of values from two similar cells. These emf's have been corrected for thermoelectric potentials between dif- ferent metal electrodes and leads. The corrections are 8 to 10 mv for the Fe-Ni and the Cr-Fe cells. The Cr-Ni cells needed no correction because the Cr electrodes were attached to Ni lead wires, with the junction at the cell temperature. bCer concentration, 1.55 mole %; F"eF2 concentration, 0.77 mole %. cFer concentration, 0.74 mole %; NiF2 concentration, 0.21 mole %. dCrF2 concentration, 1.40 mole %; NiF2 concentration, 0.31 mole %. and the values in Table 4.13, it is found that at 650°C YerF, = 035 , YFeF, YFeF 2 ~ = 0.002 |, )’NiF2 yCrF2 [a V] = 000006 ’ VNin if the convention that a = 1 for pure, solid MF2 is used, These activity coefficients, together with results of hydrogen equilibrium measurements which show that the activity coefficient of FeF in very dilute solution is nearly unity, if solié FeF . is used as a base, indicate that the activity coef?icien’r of NiF, is nearly 500. Since the saturated NiF, solution at 650°C contains about 0.25 mole % NiF ,, the saturated solution apparently has an activity of approximately unity (500 x 0.0025), if solid NiF2 is used as a base. This result is consistent either with having the satu- rating solid phase NiF, instead of NiF ,-ZrF ,, at the temperature of measurement, or with a phase diagram in which NiF,-ZrF, has an incongruent 98 melting point that is not much above the tempera- tures used, in which case the solution saturated with NiF_.ZrF, is nearly the same as the meta- stable solution saturated with NiF . Concentration cells were measured to find the change of the activity coefficient with concen- tration, ldentical chromium electrodes but differ- ent CrF_ concentrations (Table 4.14) and identical iron electrodes but different FeF2 concentrations (Table 4.15) were used., [f, as before, it is assumed that the Na* and F= ions carry all the current, the cell reaction is MF (x) + 2(NaF (x;) = MF ,(x,) + 2tNaF(x]) . With low concentrations of MF., the activities of NaF are nearly equal on both sides, and the potential is given by RT @3 RT Y3*3 E=—lpn—= ~-—n T . nF a, nF Y3%3 Duplicate sets of cells were measured at inter- vals over a period of two days., All behaved quite similarly, and the data were, in general, quite reproducible, with the data from the cells con- taining iron being more closely reproducible than those from the cells containing chromium., The most reproducible readings were obtained at 550°C, but the very low values indicate that the solubilities of CrF, and FeF, were exceeded PERIOD ENDING DECEMBER 10, 1955 TABLE 4.14, ACTIVITY RATIOS FOR CrF2 IN NuF-ZrF“ {53-47 mole %) FROM THE CELL Cr|CrF (x ), NaF(x,), ZrF (xp) || ZrF (x9), NaF(x), CrF(x))[Cr Temperature i E (v) al/“; Y'|/Y; (°C) Measured ldeal For Crf:2 Mole Fraction Ratio xl/x; = 0.0099/0.0307 = 0,322 700 6.0377 0.0475 0.406 1.26 650 0.0357 0.0450 0,407 1.27 600 0.0247 0.0425 0.518 1.61 550 0.0005 0.04M 1.0* For CerMole Fraction Ratio x]/x; =0,0098/0,0310 = 0,316 700 0.038 0.0475 0.403 1.28 650 0.0363 0.0450 0.401 1.27 600 0.0335 0.0425 0.410 1.30 550 0.0018 0.0401 * For CrF, Mole Fraction Ratio x /x| =0.0027/0,031 = 0.871 700 0.0827 0.1023 0.139 1.60 650 0.0791 0.0970 0.134 1.56 600 0.0682 0.0917 0.163 1.87 550 0.0307 0.0865? 0.4207?* 4.827 For CrF, Mole Fraction Ratio x /%] = 0,0027/0.0276 = 0.0963 700 0.0743 0.0982 0.170 1.77 650 0.0700 0.0931 0.172 1.79 600 0.0660 0.0880 0.173 1.80 550 0.0330 0.0830 * *Solubility of CrF, is less than added concentrations at this temperature. Y 2 in one or both half cells at this temperature. Since the saturating phase in these systems is believed to be MF ,-ZrF,, for which no thermochemical information is available, the data at this tempera- ture cannot be interpreted in terms of activity coefficients, However, the results at 550°C indicate that at least one half cell was saturated, and bounds can be set on the solubilities, From the emf values in Table 4.14 it can be determined that the solubility of CrF, is equal to or greater than 2.8 mole % at 600°C and is between 0.27 and 0.98 moie % at 550°C. From the emf values in Table 4.15 it is found that the solubility of FeF ,at 550°C is less than or equal to 0.26 mole % and at 600°C is equal to or greater than (.74 mole %. From previous work it is known that the solubility of FeF, at 550°C is equal to or greater than 0.23 mole % and at 600°C is less than or equal to 0.80 mole %. At all other temperatures the observed potentials are less than would be expected if the solutions were ideal, and, ac- cordingly, the activity coefficients decrease slightly with increased concentrations of added salt, Optical Properties and X-Ray Patterns for Yarious Compounds in Fluoride Systems R. E. Thoma Materials Chemistry Division G. D. White Metallurgy Division H. Insley T. N. McVay Consultants The compounds identifying characteristics of some new encountered in phase studies are 99 ANP PROJECT PROGRESS REPORT TABLE 4.15. ACTIVITY RATIOS FOR FeF, IN NaF-ZrF, (53-47 mole %) FROM THE CELL FeIFer(x.l ) NaF(xy), ZrF (x3) || ZrF (x5 ), NaF(x, ), FeF,(x;)|Fe Temperature E (v) a/a’ y /)/’ (°C) Measured Ideal 1 17 For CrF, Mole Fraction Ratio x,/x, = 0.0027/0.0074 = 0.370 700 0.0366 0.0417 0.418 1.13 650 0.0340 0.0396 0.425 1.15 600 0.0330 0.0374 0.416 1.12 550 0.0008 (0.0353) * For CtF,, Mole Fraction Ratio x,/x; = 0.0026/0,0074 = 0.347 700 0.0363 0.0444 0.421 1.21 650 0.0350 0.0422 0.415 1.20 600 0.0335 0.0398 0.410 1.18 550 0.0019 (0.0376) . *Solubility of FEF2 is less than added concentraticns at this temperature, listed below. The symbol d(/?\) means the distance 7.70 100 between reflecting planes measured in angstroms; 6.20 3 I/I] refers to the relative intensity as compared 5.57 with an arbitrary value of 100 for the strongest 5.15 8 line, Under optical properties, N, and N_ refer 5.01 19 to the lowest and highest indices of refraction, 4.77 8 respectively, and 2V refers to the acute angie 4.65 36 between the optic axes of biaxial crystals; 0 and 4,29 42 E refer to the ordinary and extraordinary indices of 3.86 100 refraction of unaxial crystals, 3.75 13 3.52 15 2N¢:|F°3KF'SZrF4 3.40 48 3.23 6 Optical data: 3.17 60 Probably meneoclinic 3,09 33 N, = 1.478 2.87 5 N,y = 1.488 2.80 9 Biaxial negative 2,58 6 2V, very large (™ 80 deg) 2.47 5 Marked cleavage 2.39 3 Xeray data: 2.33 8 Monoclinic 2.19 3 aO = 5.6 é 2.16 8 by = 5-7 A 2.14 48 0 o o = 913 A 2.09 8 a = 93 deg 2.06 8 1.95 27 d(A) i1 1.94 90 1.91 10 8.51 100 1.89 10 100 PERIOD ENDING DECEMBER 10, 1955 1.88 16 1.90 15 1.84 7 1.87 20 1.81 7 1.83 20 1.76 8 1.77 30 1.70 13 1.76 30 1.68 12 RbF°ZrF4 KF-2BeF Optical data; 4 Lath-shaped crystals Optical data: Probably monoclinic or biclinic Uniaxial negative Na = 1.490 % 0.002 0O = 1.319 N,}, = 1.502 * 0.002 E = 1.312 Biaxial negative 2V = 75 1 10 deg X-ray data: Xeray data: d(z) I/I] d(A) VI, 5.99 10 3.66 15 5.87 10 3.31 60 5461 10 3.00 100 5.04 9 2.65 4 4.77 16 2.55 4 4.50 12 0.44 4 4.29 65 2.39 13 3.98 25 2.37 4 3.82 30 2.29 15 3.78 16 2.21 15 3.72 16 2.14 15 3.63 32 2.08 4 3.59 95 2.01 5 3.52 8 2.00 6 3.44 100 .83 4 3.42 100 1.78 10 3.36 100 1.60 6 3.29 100 3.05 12 3KF-BeF, 2.60 25 2.51 30 Optical data: 2.39 12 Uniaxial positive 2.35 75 O = 1.357 2.28 1 E = 1.366 2.27 12 2.26 12 X-ray data: 2.25 12 ° 2.24 8 d(A) I/I] 2.23 12 2.15 45 3.42 4 2.07 12 3.30 7 2.03 12 3.07 22 1.96 30 2.98 75 1.93 30 2.81 ' 10 101 ANP PROJECT PROGRESS REPORT 102 o d(A) 2.64 2.54 2.51 2.39 2.27 2.04 1.96 1.85 1.78 1.73 1.72 1.69 1.60 1.56 1.54 (Nl"[4)25n|=4 Optical data: Uniaxial negative O = 1.570 E = 1.540 Colorless CSZZnF4 Optical data; Biaxial positive 2V, small N, = 1.446 N 1.458 Colorless it /1 30 37 15 100 70 12 12 SOy FleF6 Optical data: Isotopic Average refractive index = 1.432 Red brown NinF6 Optical data: Isotopic Average refractive index = 1.442 Yellow green KzBeF4 Optical data: Biaxial positive 2V = 30 deg N 1.357 a N 1.366 X A ¢ i v Colorless Probably moncclinic PERIOD ENDING DECEMBER 10, 1955 5. CORROSION RESEARCH W. D. Manly Metallurgy Division W. R. Grimes F. Kertesz Materials Chemistry Division Several Inconel forced-circulation loops that were operated with fluoride mixtures and with sodium as the circulated fluids were examined. The effects of loop wall temperature and of vari- ations in temperature differential on corrosion and mass transfer were studied for loops that circu- fated @ zirconium-base fluoride fuel mixture. The effect on corrosion of adding UF_ to an alkali- metal-base fluoride fuel mixture circulated in an Inconel loop was also investigated, The effect of the oxide content of the sodium circulated in Inconel loops was studied through the use of cold traps and the addition of oxides or deoxidizers. Additional Inconel thermal-convection loops were examined in a study of the effects of the difference between the wall temperature and the fluid temper- ature and of ZrH, additions to the fuel mixture. Thermal-convection loops fabricated from special Inconel-type alloys and various nickel-molybdenum alloys were also examined, after having circulated chF-Zr[:“-UF4 (50-46-4 mole %). Two other loops were studied in which the helium cover gas normally used was replaced with nitrogen. Brazed Inconel T-joints were tested zirconium-base fuel mixture in a special thermal- convection loop and in sodium and fuel mixtures in seesaw apparatus. Buttons of various braze materials were tested in static sodium, in a zirconium-base fuel mixture, and in NaOH. In another series of static tests, Hastelloy B T-joints brazed with various alloys were exposed to sodium and to NaF-ZrF4-UF4 (53.5-40-6.5 mole %). Ther- mocouple welds with both high and low contents of Chromel-Alumel were exposed to sodium and to NaF-ZrF ,-UF, in seesaw apparatus. Additional data were obtained on the effect of ruthenium on the physical properties of Inconel. A second boiling-sodium=Inconel loop was operated, and a third is being fabricated. Static tests were made in order to study the decarburi- zation of mild steel by sodium at high temperatures and to study the corrosion of special Stellite heats in lithium. A differential-thermal-analysis method and a solubility-temperature apparatus were used in a in attempts to determine the solubility of lithium in NaK. Several titanium carbide cermets were subjected to screening tests in NaF-ZrF4-UF4 (53.5-40-6.5 mole %) in seesaw apparatus. Inconel valve parts flame-plated with a mixture of tungsten carbide and cobalt were subjected to thermal- cycling tests, and Kentanium cermet valve parts brazed to Inconel were exposed to a static zirconium-base fluoride fuel mixture. Data were obtained on the thermal decomposition of NaOH, and corrosion and mass transfer of various structural materials in NaOH were studied. Chemical studies were made of the reaction of Inconel with sodium and with NaK and of the reaction of NaOH with nickel. A zone-melting method was developed for establishing the eutectic point of binary and ternary mixtures, FORCED-CIRCULATION STUDIES J. H. DeVan Metallurgy Division Fluorides in Inconel Several Inconel were operated with fluoride mixtures were ex- amined. The operating conditions of these loops, which were operated by Aircraft Reactor Engi- neering Division personnel,! are listed in Table 5.1. Loops 4935-5 and 4935-7 were operated to provide information for evaluating the effects of loop wall temperatures on corrosion and mass transfer. The bulk-fluoride-mixture temperatures were the same for both loops, but the fluid velocities and the distribution of heat to the hot legs were varied in order to vary the maximum wall temperature in the heated sections of the loops. The operation of both loops was interrupted by power failures and the resultant freezing of the fluoride mixtures. [n attempts to restart the loops, leaks developed in the tubing, and further forced-circulation loops that operation was prevented, ]C. P. Coughlen, P. G, Smith, and R. A. Dreisbach, ANP Quaf. PTOg. Rep. Sept. 10: 1955! ORNL"]947| P 38- 103 vol TABLE 5.1, OPERATING CONDITIONS OF FUSED-SALT-INCONEL FORCED-CIRCULATION LOOPS Loop No. 4935-5 4935-7 7425-41 4950-6 7425-1A 7425-2 Fluoride mixture circulated NuF-ZrF4-UF4a NaF-ZrFA-UF“a NaF-ZrFA-UF“a NoF-Zer-UF4a NaF-ZrF ,-UF % NaF-KF-LiF- 4 4 b (UF4 + UF3) Operating time, hr 681 332.5 1000 1000 1001 550.5 Preliminary operating period at 22 24 162 24 24 15.25 isothermal temperature, hr Maximum fluoride-mixture 1500 1500 to 1510 1650 1520 1500 1500 temperature, °F Maximum tube-wall tempera- "~ 1582 ~1710 1700 1620 1600 1570 ture, °F Temperature gradient of 200 200 200 300 200 200 fluoride mixture, °F Reynolds Number 9,000 to 10,000 ~ 6000 2750 ~ 8000 10,000 10,000 Yelocity, fps 7.05 4.6 2.06 5.22 6.5 4.1 Length of heated tube, ft 23 23 First section 9.08 7.92 7.92 7.92 Second section 7.92 9.08 9.08 .08 Total length of loop, ft 53 53 55.5 51 51 51 Method of heating Gas Gas Electrical Electrical Electrical Electrical resistance resistance resistance resistance Shape of heated section Coiled Coiled Straight Straight Straight Straight Ratio of hot-leg surface to 2.86 2.86 7.34 7.34 7.34 7.34 loop volume, in.2/in.3 Cause of termination Power failure Power failure Scheduled Scheduled Scheduled Pump shaft seized Maximum depth of attack, mils 5 8 10 8 9.5 1.5 Composition: 50-46-4 mole %. bComposifion: 11.2-41-45.3-2.5 mole %, with 1.07% of the total of 11.9 wt % uranium reported to be present as u3t, LA0dTY SSIY00A¥d LDO3r0dd ANV Although these loops did not complete the scheduled operating period of 1000 hr, the results of metallographic examination qualitatively demon- strate that, for a constant bulk-fluoride-mixture temperature, increases in wall temperature sig- nificantly increase the amount of corrosion. As shown in Figs., 5.1 and 5.2, the attack in loop 4935-7 after 332 hr at the higher wall temperature (~1710°F) was to a depth of 8 mils, and the attack after 681 hr in loop 4935-5, which operated with a maximum wall temperature of about 1582°F, was to a depth of 5 mils. The examination of loop 7425-41, in which the fluid temperature reached 1650°F, with an ac- companying maximum wall temperature of about 1700°F, provided further evidence? that wall temperature is a more critical variable than bulk- fluoride-mixture temperature. The maximum attack in this loop, which was heated by electrical resistance, was only 10 mils after 1000 hr, that is, only 1 to 2 mils greater than the attack found in loops with similar temperature differentials but with a fluid temperature of 1500°F and a wall temperature of 1650°F. Two loops (4950-6 and 7425-1A) of a series being operated to evaluate variations in temper- 26, M. Adamson, R. 5. Crouse, and A. Taboada, ANP Quar, Prog. Rep. Sept. 10, 1955, ORNL-1947, p 97. UNCLASSIFIED ] T-8111 ES;?.ng] T INCH T i fl; 2 E'la R B Fig. 5.1. Moaximum Attack in Inconel Forced- Circulation Loop 4935-7 After Circulating NaF- ZrF -UF , (50-46-4 mole %) for 332 hr at a Maximum Fluid Temperature of 1510°F and a Maximum Wall Temperature of About 1710°F. (Smggt with caption.) PERIOD ENDING DECEMBER 10, 1955 UNCLASSIFIED ] | . T-8269 ¥ Loot ! ‘ L L x s z " juo08 | 007 100w ' ‘ L w Sl FE] T% Fig. 5.2, Maximum Attack in Inconel Forced- Circulation Loop 4935.-5 After Circulating NaF- ZrF ,-UF, (50-46-4 mole %) for 681 hr at a Maximum Fluid Temperature of 1500°F and a Maximum Well Temperature of About 1582°F. (&@® with caption.) ature differential were examined following the scheduled 1000 hr of operation. In the operation of these loops the maximum wall and fluid temper- atures were maintained constant, while the fluid velocity and the hot-leg temperature pattern were varied to achieve a specified cold-zone temper- ature and fluid-temperature drop. Since varying the Reynolds number, or velocity, has not had a measurable effect on comosion results, as evidenced by compamsons of results of thermal- convection loop tests with results of forced- loop tests, it was felt that any differences in the corrosion in these two loops would be attributable directly to the difference in temperature drop. The temperature differentials of the two loops were 200 and 300°F, and the maximum hot-leg attack was very nearly the same in each, that is, 8 to 9.5 mils. It appears, however, from data on other loops operated with a 200°F temperature drop, that the attack in loop 7425-1A was ex- cessive and would normally have been about 7 mils. A leak which developed in this loop following a pump failure and subsequent reheating for startup necessitated the replacement of a short section of tubing between the cooler and the pump after 22 hr of operation. This interruption may have caused the attack to be deeper than it had been expected to be. convection 105 ANP PROJECT PROGRESS REPORT Negligible attack was found in loop 7425-2, which circulated an alkali-metal-base fluoride mixture, NaF-KF-LiF (11.5-42-46.5 mole %), with sufficient UF3 and UF4 added to give 11.9 wt % uranium, The portion of the uranium that was present as U?* was reported to be 1.07 wt %. The loop operated only 550 hr because of a pump motor failure, and the maximum attack was to a depth of 1.5 mils in the hottest section, as shown in Fig. 5.3. A continuous metal deposit, probably of uranium, that was found in the cold leg of the UNCLASSIFIED T-§509 BRI £ o o T INC!:ES T A i E g EFEE ~ o w : [o Fig. 5.3. Maximum Attack in Inconel Forced- Circulation Loop 7425-2 After Circulating for 550 hr the Fluoride Mixture NaF-KF-LiF-(UF4 + UF,) (11.2-41-45.3-2.5 mole %), with 1% of the Total Uranium Present as U3*, (dweset with caption.) loop indicates that the amount of UF, present was larger than that found by chemical analysis, since no deposit was found in the thermal- convection loop operated with material that was reported by chemical analysis to be the same. The chemical analyses showed no significant increase in chromium content of the fluoride mixture during the test, and in this respect the analyses confirm the observed attack. Sodium in Inconel A series of 1000-hr tests in Inconel forced- circulation loops has been completed in which the oxide level in the sodium being circulated was varied through the use of cold traps, deoxidizers, or oxide additions. These tests concluded the first series of tests in a study of the effect of oxide contamination on mass transfer in sodium- Inconel forced-circulation systems. The results obtained for this series of tests are presented in Table 5.2. The total weight of the deposit, which is a convenient parameter for comparison of mass- transfer effects, was determined by brushing all the deposits from the loop sections after the sodium had been removed. The metal deposit found in loop 7426-2, which circulated sodium to which 0.05% O, was added in the form of Na,0O,, weighed slightly more than that found in control2 loop 4951-8, which included a cold trap to remove oxides. The deposit weighed much less, however, than that found in loop 4951-5, which circulated sodium to which 0.15% c, TABLE 5.2. MASS TRANSFER IN INCONEL FORCED-CIRCULATION LOOPS THAT CIRCULATED SODIUM CONTAINING YARIOUS AMOUNTS OF OXYGEN Maximum fluid temperature: 1500°F System temperature differential: 300°F Operating time: 1000 hr Deposit Oxide Content Loop No. Variable Maximum Thickness Weight Initial Final (mils) (9) (%) (%) 4951-8 Cold trap in system 14 13 0.074 0.035 (control) 7426-2 0.05% 02 added to system as Na,0, 20 15 0.035 0.025 4951-5 0.15% 02 added to system as Nc1202 30 26 0.036 0.027 7426-1 1% barium added; cold trap in system 20 - 13 0.035 106 A e ke e = - b e was added. The Na,O, additions were not de- tectable by oxide analyses of the sodium in either loop. The addition of barium to act as a deoxidant in the sodium circulated in loop 7426-1 effected no observable decrease in the weight of the mass- transferred deposit in comparison with the weight of the deposit found in control loop 4951-8. The only evidence of an effect attributable to the barium was seen in the hot leg, where metallic layers distinctly different from the base metal were found in conjunction with intergranular pene- frations, Metallographic examinations showed intergranular attack to a maximum depth of 2 mils in the hot legs of all loops. The attack found in loop 4951-8 is shown in Fig. 5.4. Additional corrosion in the form of uniform surface removal possibly occurred, but such corrosion cannot be evaluated easily. UNCLASSIFIED . T«8504 EE] o &> e |' 1 B 1 ¥ I§ iNCHES Lt \ u o o S & < = PEE e w £l a - o l Fig. 5.4. Intergranular Attack on Hot-Leg Surface of Inconel Forced-Circulation Loop 4951-8, Which Circulated Sodium at 1500°F for 1000 hr. The operation of a sodium-Inconel loop with a maximum fluid temperature of 1000°F has shown the effect of temperature on mass transfer to be quite significant., A visuval inspection of the loop, which was operated for 1000 hr with a temperature differential of 200°F, showed the original oxide film on the Inconel tubing to be undisturbed. The loop was also shown metallographically to be free of deposited material, and there was no hot-leg attack. PERIOD ENDING DECEMBER 10, 1955 THERMAL-CONVECTION STUDIES J. H. DeVan E. A. Kovacevich Metallurgy Division Effect of Difference Between Loop Wall Temperature and Fluid Temperature A special thermal-convection loop (832) was operated for the purpose of obtaining the wall temperature distribution in a standard loop having a 1500°F maximum fluoride-mixture temperature and a 250°F fluoride-mixture temperature drop. The fluoride mixture used was NaF-ZrF ,-UF, (50-46-4 mole %). Eight thermocouples were attached by four dif- ferent methods, as shown in Fig. 5.5, to the Inconel pipe under the first heater on the hot leg of the loop. The thermocouples were welded or brazed to the pipe, were placed in a single well with the tip of the thermocouple welded or peened into place, were placed in a double well with each wire of the thermocouple peened into place, or each wire was placed in a hole drilled into a weld bead. A maximum difference of 10°F was observed among the four types of thermocouples, and, as expected, the well, or sunken, type of attachment gave the lower readings. Thermocouples were also attached by the double- well method to other sections of the hot leg, as indicated by Fig. 5.5. An analysis of the temper- ature readings of all the thermocouples indicates that a difference of 160°F existed between the maximum fluid temperature and the hottest section of the loop. Hence, loops described as having operated with a fluid temperature of 1500°F had wall temperatures as high as 1660°F in the heated Zone, The maximum wall temperature is known to be an important parameter in fluoride-mixture corrosion, and it must be considered when thermal-convection and forced-circulation loops are compared. The forced-circulation loops have, in general, lower wall temperatures for a given maximum fluid temperature — a factor which, in part, accounts for the attack observed in forced-circulation loops being lower than that found in thermal-convection loops with similar bulk fluid temperatures. Effect of Zirconium Hydride Additions to Fluoride Mixture A series of six Inconel thermal-convection loops were operated with the fuel mixture Nch-ZrFA-UF4 107 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 11529 "Bg-in. IPS RISER LEG- 6-in. CLAMSHELL HEATERS — 26in. SCH-10 {C.085~in WALL) Ay Z-Z A, By Ay g A, Ag !?ERMOCOUPL.E DISTANCE FROM TEMPERATURE TYPE RISER LEG (in.) (°F) A 9 1610 A2 9 1659 A3 7 1664 A4 1 1649 Ag 14 1611 Ag 18 1644 Ap 23 1614 Ag 30 1544 B, 9 1661 Ba g 1693 B3 3 1500 Ba - 1300 Bs - 1290 A B C D Bg - 1265 { - / i// A/_/ B, 125 1625 SPECIAL THERMCCOUPLE RING Co g i653 POSITICN Z-Z TYPES OF THERMOCCUPLE ATTACHMENT D, 9 1653 Dy 9 1641 Fig. 5.5. Thermocouple Arrangement of Special The perature Distribution. (50-46-4 mole %) to which various amounts of ZrH had been added. The ZrH, reduces the UF present in the fluoride mixture to UF,, according to the reactions: ZrH2—~% Ir + H2 2UF, + H,—> 2UF, + 2HF 108 rmal-Convection Loop (832) for Studying Wall Tem- Since the corrosive attack of Inconel systems by fluoride mixtures is believed to be associated with the reaction UF, + C° = 3UF, + GF, it was hoped that an increase in the UF, content would shift the equilibrium reaction to the left and thereby considerably reduce the leaching of chromium from the Inconel and the resulting void formation. Chips of ZrH, were added to the fluoride mixtures to be used in fi'le six loops in amounts calculated to give weight percentages of 0.4, 0.5, 0.6, 0.7, 0.8, and 0.9, respectively. The resulting molten mixtures were then allowed to pass through o 30-4 nickel filter so that particles larger than 30 1 would be removed and so that an indication of the solubility of the added ZrH, would be obtained. As may be seen from the data presented in Table 5.3, the quantity of U3* found in the fuel mixture increased with the increases in the quantity of ZrH, added. However, when the U3* in the sample taken during filling was greater than 1.5%, the U3* in the sample taken after circulating was less than 1.5% because of the disproportion- ation of UF, to UF, and the alloying of the PER!IOD ENDING DECEMBER 10, 1955 uranium with the walls of the loop. The total uranium decreased during operation of the loop in each case, and the decrease was apparently associated with the appearance of a metallic layer on the loop walls. The corrosive attack of the mixtures containing ZrH, was lower in all cases than the attack found in loop 800. The maximum attack (0.5 mil) in loop 789 is shown in Fig. 5.6. The observed reductions in corrosive attack are verified by the low chromium contents of the fluoride mixtures after circulation, Thin metallic layers were found on both the hot and the cold legs of the loops with more than 0.4 wt % ZrH,, except loop 790. The thickness of the layer, identified as metallic uranium, increased with increased ZrH., content of the fluoride mixture. In loop 792 Ehe layer was 0.5 to 1.0 mil thick. The absence of the layer in loop 790 is thought to be due to the low percentage of U3*. Unfortunately the U3* content was not TABLE 5.3, ANALYSES OF N<:|F--1'!rF4-UF4 (50+46-4 mole %) MIXTURES TO WHICH ZrH, WAS ADDED BEFORE CIRCULATION IN STANDARD INCONEL THERMAL-CONVECTION LOOPS Loop operating time: 500 he Mo .~ Maximom fluid temperature: 1500°F . Uranium Content Impurities Loop ZrH, Added ~ D°Pth of Hot-Leg (wt %) (opm) Attack Fluid Sampled ° PP No, (W" %) 3+ (mils) Total U U Ni Cr Fe 787 0.4 5 During filling 8.77 0.90 15 85 95 After circulating 8.52 0.96 30 110 45 788 0.5 2 During filling 9.30 0.71 3 15 50 After circulating 7.47 0.81 270 65 85 789 0.6 0.5 During filling 8.31 1.03 15 35 75 After circulating 6.62 1.23 <1 60 65 790 0.7 1.5 During filling 6.61 * 40 25 30 After circulating 6.15 1.03 15 90 85 After circulating 6.31 1.23 <1 65 90 792 0.9 1 During filling 6.72 5.35 30 35 45 After circulating 6.30 1.56 <] 40 85 800 None 10 During filling 8.64 20 105 30 {control) After circulating 8.46 40 850 90 *Not reported, 109 ANP PROJECT PROGRESS REPORT ascertained before operation of the loop, but the relatively large amount of chromium present in the fuel after operation of the loop indicates that the ZrH, addition was not entirely effective in UNCLASSIFIED T-8794 EEE] 9 T Y INCHE ) d B S o ~ l' 3 s B © lg . * - . . N : " .t / » . . ' Ly -~ s « ¥ ’ ot . / n E > ol Fig. 5.6. Maximum Attack in Inconel Thermal- Convection Loop 789 After Circulating NaF-ZiF ,- 'JF4 (50-46-4 mole %), with 0.6 wt % Zer Added, for 500 hr ot 1500°F. (Jwef® vith caption.) producing U3*. As may be seen, additions of more than 0.6 wt % ZrH2 were not effective in further reducing the corrosive attack, and they had the deleterious effect of causing the formation of a metallic uranium deposit. Loops Fabricated from Special Inconel-Type Alloys A group of special Inconel-type alloys were vacuum-cast by the Fabrication Group of the Metallurgy Division and were drawn by the Superior Tube Co. into shapes satisfactory for thermal- convection-loop fabrication. The compositions of these alloys, as may be seen in Table 5.4, varied in chromium content from the nominal [nconel composition, and molybdenum was added to one heat. Standard-size loops were fabricated from the special alloys, but Y.in.-dia, 0.065-in.-wall tubing was used rather tfion the standard ¥ -in. IPS sched-10 pipe. The results obtained from the operation of these loops are presented in Table 5.4. The data show considerable reduction in depth of attack of those alloys with chromium contents of less than 10%. Furthermore, the addition of 10% molybdenum had little effect in TABLE 5.4. RESULTS OF OPERATION OF THERMAL-CONVECTION LOOPS FABRICATED FROM SPECIAL INCONEL-TYPE ALLOYS Circulated fluid: NuF-ZrF4-UF4 (50-46-4 mole %) Maximum fluid temperature: 1500°F Loop Ne. Alloy Composition Operating Time Maximum Depth of Attack Heat No. (wt %) (hr) (mils) 520 14 76 Ni-7 Fe~17 Cr 1000 13 (standard Inconel) 522 15 76 Ni-14 Fe-10 Cr 823* 8 523 15 76 Ni—-14 Fe=10 Cr 1000 12 524 16 76 Ni-19 Fe=5Cr 647* 3 525 16 76 Ni--19 Fe-5 Cr 460* 3 526 17 83 Ni—~7 Fe-10 Cr 1000 13 527 17 83 Ni~7 Fe~10 Cr 1000 15 521 18 74 Ni-10 Fe—6 Cr=10 Mo 1000 2 472 18 74 Ni-10 Fe—6 Cr=10 Mo 500 3.5 *|_eak developed in loop. 110 an alloy with 6% chromium in comparison with the effect of reducing the chromium content to 5%. The attack in standard Inconel loop 520 and that in loop 521, which was fabricated from the alloy containing 6% chromium and 10% molybdenum, are compared in Fig. 5.7. Nickel-Molybdenum Alloy Loops A series of nickel-molybdenum alloys were prepared by the Fabrication Group to provide material for a study of the corrosion properties of nickel-molybdenum alloys having compositions modified from the composition of commercial Hastelloy B. Thermal-convection loops fabricated from materials with the compositions shown in Table 5.5 were operated for 1000 hr at 1500°F with NaF-ZrF4-UF (50-46-4 mole %) as the circu- lated fluid. The deep attack found in loop 1008 was undoubtedly caused by the air leak which interrupted the test, since loop 1013, which was identical, showed attack to a depth of only 0.5 mil after operating for 831 hr. The results obtained with the special alloys and, for comparison purposes, with Hastelloy B and Inconel are shown in Table 5.5. Varying the molybdenum and chromium contents in a nickel-molybdenum alloy does not appear to have much effect on corrosion of the alloy in NaF-ZrF4-UF4 (50-46-4 mole %) at 1500°F. The attack found in loop 1009, which PERIOD ENDING DECEMBER 10, 1955 was fabricated from an 85% Ni-15% Mo alloy is shown in Fig. 5.8. Effect of Nitrogen Atmosphere Two Inconel thermal-convection loops, 801 and 802, were tested with NaF-ZrF -UF, (50-46-4 mole %) under a nitrogen atmosphere to determine the possibility of substituting this cover gas for the currently used helium. After operation for 500 hr at 1500°F, both loops showed light surface attack to a depth of 1 mil in all sections, with subsurface void formation to a depth of 10 to 13 mils in the hottest section. Slight evidence of surface-layer formation was found in the hot zone, but no deposit or layer was seen in the cold leg. The maximum attack in these loops slightly exceeded that for a standard Inconel loop with helium used as the cover gas; the attack in the standard loop is normally less than 11 mils. The fluoride mixture was removed from loop 801 by melting, and a gold-colored deposit, which was believed to be CrN or Cer, was observed on the walls of the trap. X-ray diffraction analysis of this deposit failed, however, to show an identi- fiable pattern other than that of the fuel mixture. The following metallic elements were determined to be present, spectrographically: 2 wt % Cr, 1 wt % Ni, 0.5 wt % Fe, 5.0 wt % Ti, 0.2 wt % Al The fluoride mixture circulated in loop 802 was Fig. 5.7. Comparison of Attack Found in (a) Standard Inconel Loop 520 with That Found in (b) Loop 521 Fabricated from a Special Inconel-Type Alloy Containing 6% Chromium and 10% Molybdenum After Circulating NaF-ZrF ,-UF , (50-46-4 mole %) at 1500°F for 1000 hr. (Swwwe®t with caption.) 111 ANP PROJECT PROGRESS REPORT TABLE 5,5. RESULTS OF OPERATION OF NICKEL-MOLYBDENUM ALLOY THERMAL-CONVECTION LOOPS COMPARED WITH RESULTS FROM HASTELLOY B AND INCONEL LOOPS Circulated fluid: NaF-ZrF4-UF4 {50-46-4 mole %) Maximum fluid temperature: 1500°F Operating Maximum Depth Alloy Composition ) L.oop No. Heat No. (wh %) Time of Attack (hr) {mils) 181 Hastelloy B 29 Mo—-66 Ni—5 Fe 1000 1 1004 DP1-30 24 Mo-76 Ni 791* 0 1006 -37 3 Cr-20 Mo—77 Ni 1000 1 1007 .33 15 Mo-85 Ni 1000 1 1008 -40 5 Cr-20 Mo-75 Ni 240* 10 1009 -33 15 Mo—-85 Ni 1000 1 1010 Inconel 76 Ni=~17 Cr-7 Fe 1000 12 1013 DP1-40 5 Cr-20 Mo=75 Ni 831~ 6.5 1014 -30 24 Mo—76 Ni 1000 2 1015 -30 24 Mo—76 Ni 1000 0.5 *Terminated because of a leak. e GENERAL CORROSION STUDIES {001 | E. E. Hoffman vos W. H. Cook D. H. Jansen 14 Metallurgy Division L= R. Carlander o] Pratt & Whitney Aircraft : :2: Thermal-Convection Loop Tests of Brazed Inconel T-Joints in NaF-ZrF ,-UF ST 4 4 Lo D. H. Jansen L e Metallurgy Division - A thermal-convection loop fabricated from ]/2-in. i sched-40 Inconel tubing has been operated with Fig. 5.8, Attack Found in Loop 1009, Which Was Fabricated from an 85% Niw15% Mo Alloy, After Circulating NuF-ZrF‘-UF‘4 (50-46-4 mole %) for 1000 hr at 1500°F. removed mechanically from the trap rather than being melted out. This trap also showed a deposit of the type found ih the trap of loop 801, but identification of this deposit has not yet been completed. .k 112 NaF-ZrF ,-UF, (50-46-4 mole %) as the circulated fluid, During fabrication of the loop, six Inconel segments were brazed into the hot-leg section with Coast Metals No. 52 (89% Ni—-5% Si—4% B-2% Fe) brazing alloy. An Inconel sleeve was welded around this section, as shown in Fig. 5.9. Two typical Inconel segments are shown in Fig. 5.10. The duration of the test was 1000 hr, and the hot and cold legs were maintained at 1500 and 1100°F, respectively. The inner walls of the segments and two samples from each brazed joint were examined metallo- UNCLASSIFIED ORNL-LR-DWG 9715 . INCONEL JACKET WELDED AROUND BRAZED INSERT = s BRAZED JOINTS e e e e e e e L= ol ?DIF\’ECTION OF FLOW Fig. 5.9. Diagram of Thermal-Convection Loop with Brazed Insert in Hot-Leg Section, HHCLASSIFIED Y-1690% Fig. 5.10. Inconel Segments of the Type Brazed into the Hot-Leg Section of an Inconel Thermal- Convection Loop. PERIOD ENDING DECEMBER 10, 1955 graphically, after the test, for evidence of attack. The results of the examinations are summarized in Table 5.6. No evidence of mass transfer was found in the loop. However, the brazed joints were porous and shrinkage voids were present, as shown in Fig. 5.11. The area where the braze material was applied to the outer walls of two Inconel segments is shown in Fig. 5.12, and two brazed joints on the inner wall, which was exposed to the fuel mixture, are shown in Fig. 3.13. TABLE 5.6. RESULTS OF OPERATION OF A THERMAL-CONVECTION LOOP WITH BRAZED INCONEL SEGMENTS IN HOT-LEG SECTION Circulated fluid: NuF-ZrFA-UF4 {50-46-4 mole %) Hot-leg temperature: 1500°F Cold-leg temperature: 1100°F Duration of test: 1000 hr Attack on Braze Brazed (mils) Attack on Wall Joint (mils) Sample 1 Sample 2 1 1.4 1 4 2 3 2 3 3 0.6 2 4 4 1.3 3.5 3 5 9 2 4 Seesaw Tests of Brazed Inconel T-Joints in Sodium and in Fuel Mixtures D. H. Jansen Metallurgy Division The seesaw test apparatus has been used for a series of tests in sodium and in fuel mixtures of Incone! T-joints brazed with the alloys listed in Table 5.7. The duration of each test was 100 hr, and the hot- and cold-zone temperatures were 1500 and 1100°F, respectively. Two T-joints brazed with the 80% Ni-10% Cr-10% P alloy were placed in the same tube. One of these, joint No. 1, was cut from a T-bar on hand, and the other, joint No. 2, was a new T-joint. The difference in attack on these two joints, as shown in Figs. 5.14 and 5.15, is be- lieved to have resulted from unknown variations in the phosphorus content of the brazing alloy. 113 ANP PROJECT PROGRESS REPORT Fig. 5.11. Brazed Joints from Inconel Insert in Hot-Leg Section of a Thermal-Convection Loop Showing Porosity of Braze Material, Etched with 10% oxaiic acid. 7X. Ni PLATE] INCHES .03 T 00X i Fig. 5.12. Outer Walls of Two Inconel Segments Showing Area Where Braze Material Was Applied. Etched with 10% oxalic acid. 100X. Reduced 13%. 114 ; PERIOD ENDING DECEMBER 10, 1955 e . A . Fig. 5.13. Inner Surface of Inconel Insert Showing Portions of Two Brazed Joints After Exposure to Flowing NaF.ZrF .UF , (50-46-4 mole %) ot 1500°F for 1000 hr. Etched with 10% oxalic acid, (et with caption.) 100X, Reduced 13%. UNCLASSIFIED Y-T6389 - ! Fig. 5.14. Inconel T-Joint No. 1 Brazed with an 80% Ni-10% Cr~10% P Alloy After Exposure to Sodium for 100 hr in Seesaw Apparatus at a Hot-Zone Temperature of 1500°F. Attack to a depth of 11 mils may be seen. Etched with 10% oxalic acid. 150X. Reduced 12%. 115 ANP PROJECT PROGRESS REPORT BLTTTEITTTEITTT BT e £ ox 1 15 4 ! Fig. 5.15. Inconel T-Joint No. 2 Brazed with an 80% Ni~10% Cr-10% P Alloy After Exposure to Sodium for-100 hr in Seesaw Apparatus at a Hot-Zone Temperature of 1500°F. No attack may be seen, Etched with 10% oxalic acid. 150X. Reduced 12%. UNCLASSIFIED UNCLASSIFIED Y-16820 N Y-14819 Fig, 5.16. Inconel T-Joints Brazed with a 75% Ni-25% Ge Alloy After Exposure to (a) Sodium and to (b) NaF-ZrF ,-UF , (50-46-4 mole %) for 100 hr in Seesaw Apparatus at a Hot-Zone Temperature of 1500°F. Note attack to a depth of 2 mils on both joints. The dark network in the grain boundaries is a fine, lamellar type of structure. (Secret with caption.) 200X. Reduced 19%. ' 116 A e - e e b mime oo am = This alloy will be tested again in sodium, with the phosphorus content closely controlled. The 75% Ni-25% Ge alloy has fair resistance to both the sodium and the fuel mixture, as shown in Fig. 5.16. The results of the tests are presented in Table 5.7, in which the brazing alloys are listed in order of decreasing corrosion resistance, as determined by the following arbitrary scale: Resistonce Depth of Attack (mils) Good Otol Fair 1to3 Poor 3toé Bad Over 6 PERIOD ENDING DECEMBER 10, 1955 Static Tests of Braze Materials D. H. Jansen Metallurgy Division Buttons of six different braze materials were corrosion tested in static sodium and in static NaF- ZI‘F“-UF4 (50-46-4 mole %), and two palladium- rich buttons were tested in NaOH and in the fluoride mixture. The buttons, which were polished on one side, were tested in the as-received con- dition, After the tests, the buttons were cut in a line perpendicular to the polished face in order to ascertain the depth of attack. Coast Metals alloy No. 52 (89% Ni-4% B-5% Si—-2% Fe) and General Electric alloy No, 81 TABLE 5.7. RESULTS OF SEESAW TESTS OF BRAZED INCONEL T-JOINTS IN LIQUID SODIUM AND IN FUSED SALTS Duration of test: 100 hr Hot-zone temperature: 1500°F Cold-zone temperature: 1100°F - II * . Corrosive Medium Brazing Alloy Weight Change Cor'rOsadn Metallographic Notes (wt %) (%) Resistance Sodium 80 Ni-10 Cr-10 P ~0.174 Good No attack (jeint No. 2) 75 Ni—25 Ge —0.052 Fair Attack to a depth of 2 mils 50 Ni-25 Ge-25 Mo —0.252 Poor Nonuniform attack to a depth of 3'2 mi|5 80 Ni—10 Cr-10 P -0.41 Poor Attack to a depth of 11 mils (joint No. 1) N"'F‘Zl’F“'UF"4 80 Ni-10 Cr-10 P -0.102 Good No attack; fillet cracked (50-46-4 mole %) NUF-ZrF4-UF4 50 Ni-25 Ge—-25 Mo 0.0 Good Erratic attack to a depth of (53.5-40-6.5 mole %) 1 mil l"~!c:F-ZrF4-UF4 75 Ni=25 Ge ~0.060 Fair Attack to a depth of 2 mils (50-46-4 mole %) Nt::F"ZrF"-UF4 82 Au-18 Ni +0.16 Poor Attack to a depth of 4 mils (53-5'40'605 mole %) Nt:IF--ZrF“-UF4 80 Au-20 Cu —-0.072 Poor Nonuniform attack to a maximum (50-46+4 mole %) depth of 4.5 mils 117 ANP PROJECT PROGRESS REPORT (66% Ni—-19% Cr—-10% Si—4% Fe—1% Mn) showed, respectively, good and fair corrosion resistance in sodium, as shown in Figs. 5.17 and 5.18. The two 93% Pd—7% Al buttons showed poor resistance to the fluoride mixture and very poor resistance to the NaOH. A Coast Metals alloy No. 52 button tested in the fluoride mixture showed a very deep attack in only one area; the remainder of the button area exhibited attack to a depth of 0.5 mil. This erratic attack was probably an isolated case, since Coast Metals alloy No. 52 has exhibited good corrosion resistance to fluorides when used on nickel and Inconel T-joints. Buttons of the 60% Pd—-40% Ni alloy showed good corrosion resistance to both the fluoride mixture and the NaOH. The brazing materials tested are described and listed in order of decreasing corrosion re- sistance in Table 5.8. TABLE 5.8. RESULTS OF TESTS OF BUTTONS OF VARIOUS BRAZE MATERIALS IN STATIC SODIUM, NaOH, AND NuF-ZrF“-UF4 (50-46+4 mole %) FOR 100 hr AT 1500°F Alloy Composition Button Weight Corrosion Corrosive Medium (wt %) Change Resistance Metallographic Notes (%) NQF'ZrF4'UF4 60 Pd—40 Ni —-0.043 Good No attack (50-46-4 mole %) g3 py_7 Al +0.16 Poor Uniform attack to a depth of 3.2 mils 80 Ni—-10 Cr-10 P ~0.26 P oor Uniform attack to a depth of 3.8 mils 66 Ni-19 Cr=10 Si—-4 Fe-1 Mn -0.37 Poor Uniform attack to a depth (General Electric No. 81) of 4.4 mils 55 Mn-35 Ni-10 Cr ~7.3 Poor Stringer type of attack to a depth of 9.4 mils 60 Mn—40 Ni ~8.4 Poor Stringer type of attack to a depth of 21 mils 89 Ni-5 5i—-4 B-2 Fe -0.104 Poor Very nonuniform attack to (Coast Metals No. 52) a maximum depth of 26 mils in one isclated area 65 Ni—=25 Ge-=10 Cr -0,22 Poor Very nonuniform attack to a moximum depth of 37.7 mils NaOH 6C Pd—40 Ni +1.18 Good No attack 93 Pd-7 Al Poor Button partially disselved Sodium 89 Ni-5 Si—4 B-2 Fe -0.22 Good No attack (Coast Metals No. 52) 66 Ni—19 Cr—10 Si—-4 Fe—1 Mn -0.047 Fair Attack to a depth of (General Electric No. 81) 1.4 mils 65 Ni—25 Ge-10 Cr -0.083 Fair Attack to a depth of 2 mils 80 Ni-=10 Cr-10 P -0.3 Fair Attack to a depth of 3'2 mils 55 Mn=35 Ni-10 Cr -1.6 Poor Stringer type of attack to a depth of 4 to 5 mils 118 —— - PERIOD ENDING DECEMBER 10, 1955 UNCLASSIFIED CUY16616 7 : S Wf - d P w5 T T u T i < . ‘ _} e SERNT e 1% 1 Foi i | Py ¥ it o' /) m?ffl“%: ) ‘ . ‘ # Pt f‘-; X ,,-x u . - 7 : : T INCHRES 1 B ;004 .008 Fig. 5.17. Coast Metals Alloy No. 52 (89% Ni=5% Si~4% B=2% Fe) After Exposure to Static Sodium for 100 hr at 1500°F. No attack can be seen, but the second phase has been leached out at the surface. Etched with 10% oxalic acid. 500X. Reduced 10.5%. UNCLASSIFIED Y:14536 Q EEE T INCHES A T i 1] o ~ EEEEEE Fig. 5.18. General Electric Alloy No. 81 (66% Ni=19% Cr10% Si=4% Fe=1% Mn) After Exposure to Static Sodium for 100 hr at 1500°F. Attack to a depth of 1.4 mils may be seen. Etched with 10% oxalic acid. 250X. Reduced 10.5%. 119 ANP PROJECT PROGRESS REPORT Static Tests of Brazed Hastelloy B T-Joints in Sodium and in NGF-Z!‘F4-UF4 D. H. Jansen Metallurgy Division A series of Hastelloy B T-joints brazed with various alloys were tested in sodium and in N<:|!=-Zrl:4-Ul:4 (53.5-40-6.5 mole %) at 1500°F for 100 hr. The results of these tests are summarized in Table 5.9. It may be noted that only the 80% Ni—5% Cr—5% Fe-5% Si—5% B alloy has fair, or better, resistance in both the sodium and the fluoride mixture. Seesaw Tests of Chromel-Alumel Thermocouple Joins to Inconel Thermocouple Wells D. H. Jansen Metallurgy Division A number of Chromel-Alumel thermocouple as- semblies fabricated from 0.125-in.-OD Inconel tubing and 0.020-in. Chromel-Alumel wires were exposed to sodium and to NCIF'ZI‘F4'UF4 (50-46-4 mole %) in seesaw apparatus. The Chromel-Alumel content of the weld nuggets of the wire-to-tubing joins was varied to determine whether the quantity of silicon, manganese, and aluminum in the nugget would affect the corrosion rate. Portions of the thermocouple wires were melted to form the weld nuggets with high Chromel-Alumel content, whereas the Inconel tubing around the wires was melted to form the nuggets with low Chromel-Alumel content. The results of the corrosion tests conducted on these thermocouple assemblies are summarized in Table 5.10 and illustrated in Figs. 5.19 and 5.20. The Inconel tubes on which the welds of high Chromel-Alumel content were made were more heavily attacked in the nonweld areas than were the tubes with the welds of low Chromel- Alumel content. All the attack measurements were made before the specimens were etched. TABLE 5.9. RESULTS OF TESTS OF BRAZED HASTELLOQY B T-JOINTS EXPOSED TO STATIC SODIUM AND TO STATIC Nt‘:F-ZrF“»UF4 (53.5-40-6.5 mole %) FOR 100 hr AT 1500°F Brazing Alloy (wt %) Corrosive Medium Weight Change Metallographic Notes (%) grap Resistance NaF-ZrF4-UF4 80 Ni=5 Cr--5 Fe=5S5i-5B (53.5-40-6.5 mole %) 69 Ni-15 Cr-5 B.5 Si—5 Fe-1 C 69 Ni-20 Cr—11 Si 90 Ni-4 B—4 Si-2 Fe 69 Ni—-20 Cr-11 Si Sodium 69 Ni-15 Cr-5 B=5 Si-5 Fe-1 C 80 Ni=5 Cr-5 Fe-55S5i-5B 90 Ni—4 B—4 S5i-2 Fe -0.035 Good No attack along surface of fillet +0.041 Fair No surface attack, but several subsurface voids to a depth of 4 mils -0.025 Poor Uniform surface attack to a depth of 4 mils -0.014 Poor Layer of small voids 5 mils in from surface -0.049 Good Attack along fillet sur- face to a depth of 0.5 mil 0 Fair No surface attack, but several subsurface voids to a depth of 6 mils ~0.041 Fair Layer of subsurface voids to a depth of 1 mil ~(.052 Poor Layer of small voids 6 mils in from surface 120 PERIOD ENDING DECEMBER 10, 1955 TABLE 5.10. RESULTS OF SEESAW TESTS OF THERMOCOUPLE ASSEMBLIES EXPOSED TO SODIUM AND TO NaF-ZrF ,.UF , (50.46-4 mole %) FOR 100 hr AT 1500°F Corrosive Medium Type of Thermocouple Weld Attack {mils) NaF-ZrFA-UF4 (50-46-4 mole %) Sodium High Chromel-Alumel content, ground flat High Chromel-Alumel content, as welded Low Chromel-Alumel content, as welded l.ow Chromel-Alumel content, as welded inconel Tube Weld 4.5 ' 0.5 4.5 <0.5 2.0 None None None e 0COUP UNCLASSIFIED Y+17029 INCHES UNCLASSIFIED o Ye17027 Fig. 5.19. Thermocouple Assemblies with Welds of Low Chromel-Alumel Content After Exposure for 100 hr at 1500°F in Seesaw Apparatus to (a) Sodium and (b) NuF-ZrF4-UF4 (50-46-4 mole %). Inconel tube unattacked in sodium; attacked to a depth of 2 mils in fluoride mixture; no attack apparent on weld. Etched with 10% oxalic acid. (@@= with caption.) 121 ANP PROJECT PROGRESS REPORT [Ni PLATE ] ~ THERMOCOUPLE Figo 5-200 UNCLASSIFIED Y-17024 o “m-—l I > Z L i INCONEL TUBE 0.02 0.03 » L O Q Thermocouple Assembly with Welds of High Chromel-Alumel Content After Exposure for 100 hr at 1500°F in Seesaw Apparatus to NaF-ZrF4-UF4 (50-46-4 mole %). Inconel tube attacked to a depth of 4.5 mils; weld attacked to a depth of 0.5 mils. caption.) Effect of Ruthenium on Physical Properties of Inconel D. H. Jansen Metallurgy Division Additional data were obtained on the creep- rupture properties of ruthenium-plated Inconel. An additional nonplated specimen was tested because it was found that the nonplated specimen tested previously3 had not received the same prior heat treatment as that given to the plated specimens. The results obtained are presented in Table 5.11. The plated specimens were tested as received and without a spectrographic check for the positive presence of a ruthenium plate. Therefore ad- ditional tests of spectrographically checked speci- mens are under way. 73C. F. Leitten, Jr., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 106. 122 Etched with 10% oxalic acid. (SEPemth Boiling Sodium in Inconel Loop E. E. Hoffman Metallurgy Division The results of operation of the first boiling- sodium—Incone! loop, which was terminated after 400 hr because of a pipe failure, were previously reported, Another such loop has now been operated for the scheduled test period of 1000 hr under similar thermal conditions. A slight vacuum was maintained in the system so that the sodium would boil tt approximately 1500°F. In the previous test of 400-hr duration no mass-transferred crystals were detected in the cold trap of the condenser lina,. but heavy intergranular cracking to a depth of 50 mils was detected in some areas. In this latest test, macroscopically visible quan- 4. E. Hoffman, ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 109. PERIOD ENDING DECEMBER 10, 1955 TABLE 5.11. RESULTS OF CREEP-RUPTURE TESTS OF NONPLATED AND RUTHENIUM.PLATED INCONEL Atmosphere: purified argon Stress: 3500 psi Test temperature: 1500°F Heat treatment prior to test: 100 hr at 1500°F Condition of Specimen Time to Rupture (hr) Final Elongation (%) Nonplated Plated Plated 746 14.0 873 13.4 728 13.6 tities of mass-transferred crystals were detected in the cold trap, as shown in Figs, 5.21 and 5.22. The various sections from the condenser line that were examined metallographically exhibited intergranular cracking similar to that found in the loop operated for 400 hr. In the hottest section of the condenser pipe, there was attack to a depth of 1 to 2 mils in the form of small subsurface voids. No intergranular cracking was detected in the hottest and coolest sections of the condenser pipe, but cracks were found in the straight section which connected the hot and cold traps. The cracks were to a maximum depth of approximately 25 mils (see Fig. 5.23), as compared with a depth of 50 mils in the loop operated for 400 hr. The temperatures in the areas where cracks were detected varied from 1150 to 1325°F. Metallographic examination indicated that a brittle grain-boundary phase was present in the areas where cracks appeared. Periodic thermal excursions caused by the condensing sodium led to severe thermal stresses in the pipe walls, which apparently caused the cracks to propagate. A boiling-sodium loop fabricated from type 348 stainless steel is now in operation. With this test system it will be possible to sample periodically the distilled sodium and to check its oxygen content. Decarburization of Mild S$teel by Sodium E. E. Hoffman Metallurgy Division It has been well established that various metals may be either carburized or decarburized while in contact with sodium at elevated temperatures. The direction and extent of this phenomenon depend both on the original carbon content of the sodium UNCLASSIFIED Y-16593 HOT TRAP A 750°C (1382°F) COLD TRAP 560°C (1040°F) Fig. 5.21. Sections Taken from an Inconel- BoilingSodium Loop After Operation for 1000 hr. and on the carbon content of the metal in contact with the sodium, Tests have been made to determine how far the decarburization of a type 1043 mild steel (0.433% C) will proceed in static sodium at 1830°F for 100 hr. In the two tests performed to date, the container materials were Armco iron and type 304 ELC stainless steel. The capsules were loaded and sealed in inert atmospheres. The mild-steel specimens in both capsules showed weight losses of 0.2 mg/in.2. The extent to which carbon transferred is illus- trated in Fig. 5.24. The decarburization of the type 1043 mild-steel specimen is particularly evident, The small areas of pearlite found in the 123 ANP PROJECT PROGRESS REPORT Fig. 5.22. Enlargement of Section of Cold Trap from the Inconel-Boiling-Sodium Loop Showing Mass- Transferred Crystals. 4X. wall of the Armco iron capsule after the test indicate that the Armco iron picked up carben, The type 1043 mild-steel specimen tested in the type 304 ELC stainless steel capsule was very similar in appearance to the mild-steel specimen tested in the Armco iron capsule; however, suf- ficient nickel had been picked up on the surface of the specimen in the stainless steel capsule to cause a phase transformation which extended to a depth of 1 to 2 mils. Table 5.12 shows the results of the carbon analyses of the various test components, } may be seen that the carbon content of the mild-steel specimens decreased and that the carbon content of the capsule walls in contact with the sodium increased. Static Tests of Special Stellite Heats in Lithium E. E. Hoffman Metallurgy Division Two special heats of Stellite, heats B and C, were submitted by the Union Carbide and Carbon 124 Metals Research Laboratories for corrosion tests in various liquid metals. The results of earlier tests of these materials in NaF-ZrF -UF, (53.5- 40-6.5 mole %) and in sodium were reported previ- ously.” The approximate compositions of these alloys were: Composition (wt %) Heat B Heat C Carbon 1 0.30 Chromium 35 28 Tungsten 5 Molybdenum 12 Manganese 0.50 0.50 Silicon 0.50 0.50 Nickel 3 Cobalt 58 55.7 SE. E. Hoffman et al, ANP Quar, Prog. Rep. Sept. 10, 1954, ORNL-1771, p 83. [TILEl 7] 020 2 iy — 223 | T IS(?X TITET Fig. 5.23. Wall of Condenser Pipe of Inconel. Boiling-Sodium Loop Operated for 1000 hr. Tem- perature at location shown was between 1150 and 1325°F. PERIOD ENDING DECEMBER 10, 1955 The alloys were unattacked by sodium but were attacked to depths of from 7 to 20 mils by the fused salt. The results of the tests of these materials in lithium are presented in Table 5.13. It is apparent that in these alloys there are isolated areas that are susceptible to very heavy attack. The variations in susceptibility to attack may be due to composition variations in the specimens, Solubility of Lithium in NaK R. Carlander Pratt & Whitney Aircraft Attempts were made to determine the solubility of lithium in NaK (44% K—-56% Na) by a differential thermal-analysis method and by using a solubility- temperature apparatus. In the differential thermal- analysis method, the temperature of the bath containing lithium was plotted on an automatic time-temperature recorder, and the difference in temperature between the NaK bath containing lithium and a blank bath containing only NaK was plotted on another automatic time-temperature recorder., When the lithium began to solidify, heat was given off, and a temperature difference was noted and correlated with the time-temperature chart to give the temperature at which solidifi- cation occurred. Tests were performed with 5, 6, and 10 wt % lithium added to the NaK, The results are given in Table 5.14. In the temperature-solubility apparatus, NaK with 5 wt % lithium added was heated to 200°C, agi- tated, and allowed to remain at that temperature TABLE 5.12, CARBON ANALYSES OF THE VARIOUS COMPONENTS OF SYSTEMS DESIGNED FOR STUDYING THE DECARBURIZATION OF MILD STEEL BY SODIUM Duration of tests: 100 he Test temperature: 1830°F Carbon Content* (wt %) Test System Material Analyzed Before Test After Test Type 1043 mild-steel specimen Mild-steel specimen 0.433 0.121 in Armco iron capsule Armco iron capsule 0.019 Bath zone 0.035 Vapor zone 0.019 Type 1043 mild-steel specimen Mild-steel specimen 0.433 0.100 in type 304 ELC stainless Stainless steel capsule 0.022 steel capsvule Bath zone 0.128 Vapor zone 0.022 *Two analyses performed on each sample. 125 ANP PROJECT PROGRESS REPORT UNCLASSIFIED Y-16943 WALL OF ARMCO IRON CAPSULE BEFORE TEST S20'§ T ofwen 70| | 2367 | 1 . eod ol 80 Tio ¥i0 o 30| e o [ors "~ 560 150 = - €2 a4 m L. ] 55T waLL OF ARMCO IRON CAPSULE AFTER TEST MILD STEEL SPECIMEN BEFORE TEST e ]E IE E ]§ I 2%0):: MILD STEEL SPECIMEN AFTER TEST Fig. 5.24. Armco Iron Capsule and Type 1043 Mild-Steel Specimen Before and After Exposure to Static Sodium for 100 hr at 1830°F. 200X. Reduced 18%. 126 PERIOD ENDING DECEMBER 10, 1955 TABLE 5.13. RESULTS OF STATIC TESTS OF SPECIAL STELLITE HEATS IN LITHIUM AT 1500°F FOR 100 hr Weight Change terial . Materia (g/in.z) Metallographic Netes Heat B -0.044 Uniform attack to a depth of 2 to 3 mils plus very heavy attack in several isolated areas Heat B -0.059 Approximately 90% of exposed surface attacked to a depth of 3 mils; very heavy attack on other 10% of surface Heat C -0.069 Uniform attack to a depth of 2 to 4 mils plus very heavy attack in some areas TABLE 5.14. RESULTS OF MEASUREMENTS OF THE SOLUBILITY OF LITHIUM IN NaoK Analysis Lithium Test Methods Temperature Content (°C) (wt %) Differential thermal 176 10 analysis 170 6 166 5 Temperature-solubility 25 0.014 for 2 hr. The bath was then cooled to 25°C, and a sample was obtained through a stainless steel filter. The results of chemical analyses of the samples for lithium are given in Table 5.14. it appears from the data that a small amount of lithium is soluble in NaK. Seesaw Tests of Titanium Carbide Cermets in NOF-Z!F4-UF4 W. H. Cook Metallurgy Division Some titanium carbide cermets with cobalt- or nickel-base alloys as binding material were sub- jected to screening tests in contact with NaF- ZrF,-UF, (53.5-40-6.5 mole %) in seesaw appa- ratus operated at 4.25 cpm. These cermets were submitted by the Sintercast Corp. of America. The hot and cold zones of the apparatus were at 1500 and 1200°F, respectively. Each specimen was held in the hot zone of its capsule during the 200-hr test period. The results are summarized in Table 5.15. The results of previous tests® of five of the same types of Sintercast Corp. specimens exposed to NaF-ZrF4-UF4 (50-46-4 mole %) for a 100-hr test period are also included in Table 5.15 for comparison purposes. The results of the earlier tests are considered to be less significant than those obtained in the recent tests, because at the time the earlier tests were made there were no untested specimens available for comparison, the specimen compositions were not known,’ involved techniques required for metallographically polishing these kinds of specimens. As shown in Table 5.15, only Sintercast speci- mens 1 and 7 were attacked by NaF-ZrF4-UF4 (53.5-40-6.5 mole %). The type of attack is illustrated in Figs. 5.25 and 5.26. In the metallo- graphic exomination of the specimen structures, it was noted that the titanium carbide particles in the cermets with cobalt-base binder were, in general, more globular than those in the cermets with nickel-base binder. It appears therefore that the cobalt-base alloys may have reacted more with the titanium carbide in the fabrication of the specimens than did the nickel-base ailoys. The typical extremes of the particle shapes are i llustrated in Figs. 5.25 and 5.26, and the average structure and typical appearance in the absence of attack are illustrated in Fig. 5.27. As may be noted, there was practically no mutual bonding between the titanium carbide particles, as was and little was known about the E. E. Hoffman et al., ANP Quar. Prog., Rep. June 10, 7C0mpositions and methods of fabrication of the Sintercast Corp. specimens are **Company Confidential Information?’’ and not available for general publication. 127 8¢l TABLE 5.15. SUMMARY OF THE RESULTS OF SEESAW CORROSION TESTS ON SOME SINTERCAST CORPORATION TITANIUM CARBIDE CERMETS Dperating cycle: 4.25 cpm Hot-zone temperature: 1500°F Cold-zone temperature: 1200°F Capsule material: Inconel Sintercast ' Attack (mils) i Metal Binder _ Specimen Alloy Base In NaF-ZrF -UF In NaF-ZrF ,-UF Metallographic Notes Ne. (50-46-4 mole %)* (53.5-40-6.5 mole %}** 1 Cobalt 0.5t02 4108 The attack was primarily along the boundaries between the TiC particles and the metal binder; the TiC particles were slightly smaller than those in specimens 2 and 4 and tended to be glebular in shape 2 Cobalt 0 The specimen was unattacked; the TiC particles were globular and intermediate in size between those of specimens 1 and 4 4 Cobalt 1 0 The specimen was unattacked; the TiC particles were slightly larger than those in specimens I and 2; the TiC particles were globular and sharply defined in the metal 3 Nickel 0 The specimen was unattacked; the TiC particles were not so globular as those bonded with the cobalt-base alloys, specimens 1, 2, and 4; the TiC particle size appeared to be the same as that for specimen 1 5 Nickel 0 0 The specimen was unattacked; the TiC particies were less globular in shape than those in specimen 3 and they appeared to be smaller 6 Nickel 0 The specimen was unattacked; the TiC particles were approximately the same size as those in specimen 4; they were less globular than those bonded with the cobalt-base alloys; there was more metal arec in a section than in any other of the specimens 7 Nickel 1.5t07 The attack was primarily aleng the boundaries between the TiC particles and the metal binder; the TiC porticles in this and specimen 10 were the most angular in shape of all specimens 8 Nickel 0.5t02 0 The specimen was unattacked; the structure was similar to that of specimen 6 except that there was less metal area 9 MNickel 0 The specimen was ynattacked; the structure was similar to that of specimen 8 except that there was a greater quantity of small TiC particles 10 Nickel 5 0 The specimen was unattacked; the TiC particles were angular like those in specimen 7 and were larger than those in any of the other specimens *100-hr tests. **200-hr tests, LY¥0OJIY SSIYI0YUd L1D3rodd ANV PERIOD ENDING DECEMBER 10, 1955 Fig. 5.25. Sintercast Specimen 1 (a) As-received and (b) After Exposure to NuF-ZrF4-UF4 (53.5-40-6.5 mole %) for 200 hr in Seesaw Apparatus with a Hot-Zone Temperature of 1500°F and a Cold-Zone Tem- perature of 1200°F. Specimen held in hot zone during test. Unetched. (Gwew with caption.} 500X, . Reduced 13.5%. CUNCLASSIFIED . L - Fig. 5.26. Sintercast Specimen 7 (a) As-received and (b) After Exposure to NaF-Zr F4-UF (53.5-40-6.5 mole %) for 200 hr in Seesaw Apparatus with a Hot-Zone Temperature of 1500°F and a Cold-Zone Tem- peratuwre of 1200°F. Specimen held in hot zone during test. Unetched. ({gemagywith caption.) 500X. Reduced 13.5%. 129 ANP PROJECT PROGRESS REPORT UNCLASSIFIED - Y-16889 Y-16890 Fig. 5.27. Sintercast Specimen 4 (a) As-received and (b) After Exposure to NaF-ZrF ,-UF , (53.5-40-6.5 mole %) for 200 hr in Seesaw Apparatus with a Hot-Zone Temperature of 1500°F aond a Cold-Zone Tem- perature of 1200°F. Specimen held in hot zone during the test, The holes were created by removal of TiC particles during metallographic polishing, Unetched. (Swem with caption,) previously reported for the more common types of cermets with low metal content.8:? Thermal-Cycling Tests of Inconel Yalve Disks and Seats Flame-Plated with a Mixture of Tungsten Carbide and Cobalt W. H. Cook Metallurgy Division Initial evaluation tests of valve disks and seats coated with a mixture of tungsten carbide and cobalt were made. The coating, which was being tested for suitability as a hard-facing material, was deposited by a commercial flame-plating method developed by Linde Air Products Co. 8E. E. Hoffman, W. H. Cook, and C. F. Leitten, Jr., ANP Quar. Prog. Rep. Mar, 10, 1955, ORNL-1864, p 84. 9E. E. Hoffman er al,, ANP Quar, Prog. Rep. Sept. 10, 1955, ORNL-1947, p 104, ey el 130 The coating, to be suitable, must be resistant to solid-phase bonding, galling, and wearing. The seating surfaces of four sets of Inconel valve disk and seat blanks were flame-plated by the Linde Air Products Co. with coatings 4 to 6 mils thick that were nominally 92 wt % tungsten carbide and 8 wt % cobalt. Three sets of the tiame-plated valve parts were then copper-brazed at a temperature of 2048°F to Inconel platens. Adequate wetting was obtained; however, cracks were created in the flame-plated coatings, as shown in Fig. 5.28. The cracks were thought to be caused by the difference between the coef- ficients of thermal expansion of the coating and the Inconel. Two tests were made on the fourth set of Inconel valve seat and disk blanks to further evaluate the usefulness of the flame-plated coating on Inconel. The purpose of the first test was to e e cmem e # UNCLASSIFIED Y-16856 Fig. 5.28. Inconel Valve Parts Flame-Plated with a Mixture of Tungsten Carbide and Cobalt and Then Copper-Brazed to Inconel Platens at 2048°F, Note cracks in the flame-plated coatings. determine whether the flame-plated coating was satisfactory for use at the proposed operating temperature. The disk and seat were slowly cycled in a vacuum from room temperature to 1500°F, which was approximately 100°F in excess of the proposed maximum operating temperature, There was a small amount of spalling of the plating where it was built up on corners. It the seating surfaces had been machined, as they would have been for a finished valve disk-and- seat set, there probably would not have been any spalling. There were no cracks that could be detected with a dye penetrant. In the second test, the disk and seat were heated in a vacuum from room temperature to 2012°F in 5 min, held at 2012°F for 15 min, and then allowed to cool to room temperature in approximately 3 hr. These thermal conditions were similar to those used in the brazing of the other flame-plated disks and seats, with the exception that the thermal shock was more severe. Ihe coatings on both the disk and seat developed cracks like those found in the coatings on the brazed disks and seats. The results of these two tests are illustrated in Fig. 5.29. PERIOD ENDING DECEMBER 10, 1955 Static Tests in NaF-Z¢F ,-UF, of Kentanium Cermet Yalve Parts Brazed to Inconel W. H. Cook Metallurgy Division A valve seat made from the Kentanium cermet KI151A (70% TiC=10% NbTaTiC-20% Ni) and a valve disk made from KI152B (64% TiC-6% NbTaTiC;~30% Ni) held in contact at a pressure of 10,000 psi (calculated) were exposed to static NaF-ZrF4-UF4 (53.5-40-6.5 mole %) at 1500°F for 150 hr. The surface-roughness values for the contacting surfaces were from 5 to 10 pin. before the test.'® The test was terminated when a pull rod failed. The lower valve part fell free from the upper valve part, and thus there was an indication that no solid-phase bonding had occurred. A subsequent low-power microscopic examination revealed no evidence of bonding. Visuval and microscopic examination of the 45-deg beveled seating surfaces of the valve disk and seat indi- cated that there was uniform seating during the test. This was also the first test in which the re- cently developed method of brazing Kentanium cermets to Inconel'! was tested in a fused-salt bath. There were no visible signs of attack, of distortion of the ]/“v-in.-thick nickel shim, or of cracking of the cermet pieces after the test. Effect of an Air Leak into an Inconel-Fused-Salt Test System W. H. Cook Metallurgy Division An Inconel—fused-salt tube-burst test capsule was found to be very heavily attacked above the level of the NaF-ZrF ,-UF , mixture following a 500-hr test at 1500°F. It was suggested that the attack (Fig. 5.30) was due to an air leak some- where in the system, and therefore several experi- ments were performed in which air was admitted to similar test capsules through a small tube. These test capsules failed within 24 hr at a temperature of 1500°F. The reaction products 10The usual 1.5 to 2 pin. used for solid-phase- bonding tests with bar specimens could not be ob- tained because this disk-and-seat configuration pro- hibited lapping. b, patriarca and R. E. Clausing, ANP Quar, Prog. Rep. June 10, 1955, ORNL-1896, p 145, and ANP Quar. Prog. Rep., Sept. 10, 1955, ORNL-1947, p 131. 131 ANP PROJECT PROGRESS REPORT UNCLASSIFIED Y-16584 UNCLASSIFIED Y-16659 UNCLASSIFIED - Y-16857 Fig. 5.29. Inconel Yalve Disk and Seat Flame-Plated with a Mixture of Tungsten Carbide and Cobalt. (a) As received. (b) After being cycled slowly from room temperature to 1500°F in vacuum. (c) After being heated in vacuum from room temperature to 2012°F, held at 2012°F for 15 min, and allowed to cool to room temperature in about 3 hr. 132 UNCLASSIFIED Y-16834 - Wl = Lt d I — O jh ! Q- 3 a ! ‘ - 100X o ]'o INCHES ‘ - l " } | . i Fig. 6.32. A 74.75% Ni=20% Mo=5% Cr—0.25% Ce Alloy Which Ruptured in 440 hr with an Elongation of 28% at a Stress of 8000 psi in Argon at 1500°F. 100X. Reduced 20%. 157 ANP PROJECT PROGRESS REPORT TINCHES 00X — L Fig. 6.33. An 80% Ni~20% Mo Alloy Which Ruptured in 91 hr with an Eiongation of 2% at a Stress of 8000 psi in Argon at 1500°F, 100X. Reduced 23%. UNCLASSIFIED 1 72 TNCHES 106X Fig. 6.34. An 80% Ni~20% Mo Alloy Which Ruptured in 763 hr with an Elongation of 5% at a Stress of 5000 psi in Argon at 1500°F. 100X. Reduced 23%. 158 [ ~ ‘\‘ {J"—-,._“ . }/ 5 .STRESSED " . A i PERIOD ENDING DECEMBER 10, 1955 : . . . \'\ ] UNCLASSIFIED ‘ Y-14258 Fig. 6.35. A 76% Ni-24% Mo Alloy Which Ruptured in 68 hr with an Elongation of 4% at o Stress of 8000 psi in Argon at 1500°F. 100X. Reduced 22.5%. A UNCLASSIFIED Y3 Y1261 3 v poe 14 —4 x < =z 0.02 Q.03 » e 10 2 Fig. 6.36. A 68% Ni-32% Mo Alloy Which Ruptured in 332 hr with an Elongation of 8% at a Stress of 8000 psi in Argon at 1500°F. 100X. Reduced 22.5%. 159 ANP PROJECT PROGRESS REPORT . UNCLASSIFIED | Y-14706 A o ! o o 1 ii: iI00X o o INCHES X ° 1 Fig. 6.37. A 68% Ni-32% Mo Alloy Which Ruptured in 715 hr with an Elongation of 33% at a Stress of 5000 psi in Argon at 1500°F, 100X. Reduced 23%. . 5 P S Nos . ‘ N UNCLASSIFIED ro S e 5 o ‘ : NG ; LR .“(r ‘l.\\-(-«"r . 3 . ’ . g " | STRESSED . [ L #‘ T 1 . - T I?@CHES S Fig. 6.38. A 77% Ni-20% Mo-3% Cr Alloy Which Ruptured in 674 hr with an Elongation of 8% at o Stress of 5000 psi in Argon at 1500°F. 100X. Reduced 23%. 160 PERIOD ENDING DECEMBER 10, 1955 UNCLASSIFIED Y-15689 | - STRESSED ' /} AT 4 - A A\ > VI"\_‘ i }; e .. - .\"\ i * ~ g . \t”’. s s e T w o e e } 14 | S LNl - oy L s T | & - \\' - H"a | - . \: . ‘..(L . th .V",‘ ) ‘\k ‘ " A i i 3 . i ) ‘ - T s :; “‘ \ A \ Yoo ™ T \ | § e f’/ ] . o 1 {3 le x Fig. 6.39. A 73% Ni=20% Mo=7% Cr Alloy Which Ruptured in 373 hr with an Elongation of 10% ata Stress of 8000 psi in Argon at 1500°F. 100X. Reduced 21%. Figo 6.40. N WNCLASSIFIED\ , % Y-14126 J A 75% Ni-20% Mo—5% Nb Alloy Which Ruptured in 530 hr with an Elongation of 8% at a Stress of 8000 psi in Argon at 1500°F. 100X. Reduced 21%. 161 ANP PROJECT PROGRESS REPORT TABLE 6.1, RESULTS OF EXTRUSION OF TUBE BLANKS Billet size: 3 in. in diameter Numb Soaking Reduction umber Composition of Billet Temperature ) Remarks E xtruded o Ratio ("F) 3 Nickel~type 316 stainless steel 2100 9:1 Good tube blanks obtained (duplex tube) 3 Inconel—type 316 stainless steel 2100 9:1 Good tube blanks obtained (duplex tube) 3 Monel~type 316 stainless steel 2050 9:1 Good tube blanks obtained (duplex tube) 3 Hastelloy B—type 316 stainless 2050 9:1 Surface of Hastelloy B was steel (duplex tube) somewhat roughened in all three extrusions; two extrusions showed evi- dence of cracking 2 Hastelloy B (flame —sprayed 2050 9:1 Surfaces of both extrusions with type 304 stainless steel) were somewhat roughened; one extrusion cracked 2 Hastelloy W (Inconel canned) 2050 7:1 Both extrusions shattered 1 Hastelloy W (uncanned) 2050 5,5:1 Did not extrude Six laboratory-prepared alloys were successfully extruded to rod form for corrosion studies, The compositions of the billets extruded and the re- duction ratios used are tabulated below (the soaking temperature used was 2150°F in each case): Billet Composition (wt %) Reduction Ratio 80 Fe-20 Cr 13:1 80 Fe-10 Cr—10 Ni 10:1 74 Fe—-18 Cr-8 Ni 10:1 60 Fe-20 Cr—-20 Ni 10:1 90 Ni—-10 Mo 10:1 80 Ni-10 Mo=10 Fe 10:1 Neutron Shielding Material for ART H. Inouye, Metallurgy Division M. R. D’Amore, Pratt & Whitney Aircraft Methods for producing B,C tiles to be used as neutron shielding for the ART are being studied. The Norton Company and The Carborundum Company are cooperating in this investigation. Since the neutron shield will be at a temperature of 1500 to 2000°F when the reactor is operating, studies are 162 also being made of the compatibility of Inconel and B,C in this temperature range. Previously reported studies indicate that carburization occurs at the grain boundaries and that metallic borides are formed. A coating, or barrier, material that is inert both to B, C and to Inconel will therefore be required. Tests of possible barrier materials are under way. Studies are also being conducted to determine the effects of irradiation on B, C at the expected service temperatures. Neutron capture by boron results in the production of helium and lithium, which are reported4 to be retained interstitially in the B C lattice at temperatures below 1650°F. There is evidence, however, that the helium and, possibly, the lithium are released at some higher temperature, Plates of B,C with a metallic binder are also being investigated, since they would have better thermal-conductivity, expansion, and shock-re- sistant properties. The plates can be fabricated 4p, Senio and C. W. Tucker, Jr., The Stability of Boron Carbide After 15 Atomic Percent Burnup, KAPL- 1091 (April 5, 1954). by wusing the picture-frame technique to hot roll cold-pressed, sintered, powder mixtures or by ex- truding warm-pressed billets of the powder mix- tures, A layer, equivalent to 0,010 in. of B9, that is thermally bonded to the inside wall of the Inconel shell will be required if the metal-bonded B C plates are used. ests of the reactions between B4C, binder, and Inconel shell at the service temperature are under way, and possible methods for applying the thin layer to the shell are being investigated. Inconel-Clad Niobium H. lnouye J. P. Pagg Metallurgy Division The search for a diffusion barrier to use between Inconel and nicbium in high-temperature applica- has been continued. Tantalum foils and copper—stainless steel foils were selected as the most promising of several barrier materials on the basis of bend tests and metallographic examination of pieces in the as-rolled condition and after 100- and 500-hr service tests at 900°C. These types of qualitative tests are being used to check earlier data obtained with molybdenum and titanium as the barrier materials and to obtain data on other barrier materials (iron, at present). A test pro- cedure is also being developed for obtaining quan- titative data on the mechanical properties of Inconel- clad niobium with tantalum and copper—stainless steel diffusion barriers. When a satisfactory test program has been devised, data will be obtained to determine the optimum diffusion barrier, tions Gamma«Ray Shield Material for ART Pumps J. H. Coobs J. P. Page Metallurgy Division A material having the following properties is needed for use as a gamma-shield heat dam around the impeller shafts of ART pumps: density of over 13.0 g/cm3, thermal conductivity of less than 0.12 cal/cm2:sec:(°C/cm) at 1400°F, brazeability to Inconel, and a coefficient of thermal expansion similar to that of Inconel, that is, between 10 and 20 pin./ins°C, There are no known materials that fulfill all these requirements, but it is thought that a satisfactory two-component material can be fabri- cated by powder-metallurgy techniques. A high- density ‘‘filler’’ material would be used, along with another material which would act as a binder and would ‘“‘insulate’’ the compact. From considera- PERIOD ENDING DECEMBER 10, 1955 tions of thermal conductivity, coefficient of ex- pansion, brazeability, and ease of fabrication, the nickel-base alloys appear to be the most promising for use as binder material, especially the Hastelloys and constantan. Calculations of the limiting cases of series and parallel heat flow indicate that the filler material must be either tungsten carbide or tantalum, Several tantalum-constantan compacts have been successfully hot-pressed to densities of greater than 13 g/cm3, The tantalum-base material is far superior to the tungsten carbide—base material from the standpoint of fabricability. However, metallographic examination indicates a strong tantalum-nickel reaction that results in virtual separation of the nickel from the constantan. This separation could cause variations in the thermal conductivity and could cause the thermal con- ductivity to be above the specified limit. Similar investigations of a tungsten carbide—constantan compact shows that the carbide is held in a homo- geneous constantan matrix. An apparatus is being fabricated for use in determining the thermal con- ductivities of these and similar mixtures. Control-Rod Fabrication J. H. Coobs H. Inouye Metallurgy Division Two core compacts for control rods were pre- pared by cold pressing and sintering 35 vol % (Gd-Sm), 0, in copper and in iron to densities of 79 and 82%, respectively. The iron mixture was sintered at 1100°C, and the copper mixture was sintered at 980°C, The core compacts were then canned in stainless steel, were evacuated, and were hot-swaged for a total reduction of 75% in seven passes, with the compact being reheated to 950°C between passes. The active sections of the finished rods were ]/2 in. in diameter and 7 in, in length. The cores were well compacted, with calculated densities of 97 and 98.5% for the copper and the iron mixtures, respectively. The core was thermally bonded to the stainless steel core, but it was slightly irregular in cross section., Metallographic examination showed the core components to be compatible under the sintering and hot-swaging conditions, No evidence was found of reaction be- tween the iron or the copper and the (Gd-S5m),0,. [t appears that, with slight modifications in the techniques, control rods with a finished size of 163 ANP PROJECT PROGRESS REPORT about 5/8 to % in. in diameter can be prepared, but for finished rods ¥ in. in diameter, and larger, extrusion would be preferable. A hot-pressed body with 30 wt % (Gd-5m),0, and 70 wt % Fe was extruded in an Inconel can at 2100°F, The original compact; 3/4 in. in diameter by 3 in, in length, with 30% porosity, was extruded to a core 24 in. in length by 0.230 in. in average diameter. The core-diameter variations were +0.015 in. from the average. Examinations of the cross section and the longitudinal section indicated that metallurgical bonds were attained. The density of the extruded core was 7.724 g/cm3, or 100% of theoretical. In tensile tests at room temperature the core elongated 3% before fracture, and the composite 42%. At 1650°F the composite uniformly elongated 17% before fracture. At room temperature the tensile strength of a 0.505-in.-dia rod of Inconel containing a 0.230-in.-dia rare-earth-oxide core was 66,500 psi, and at 1650°F the tensile strength was 11,800 psi. Several iron-zirconium alloys containing between 1 and 16% zirconium were studied to determine their physical and fabrication properties. The zirconium in the alloy was a stand-in for hafnium, which is o high-cross-section material that might be useful for control rods. The alloys were melted by induction heating in vacuum and were cast in a split-cast iron mold. The zirconium additions were contained in a drilled cavity in the iron melting stock. Chemical analyses of the ingots are presented in Table 6.2. All the ingots (except 527-16, which was unsatisfactory) were extruded from a 3-in.-dia ingot to 1-in.-dia rod at 2100°F, with no difficulty. Standard 0.505-in.-dia tensile-test bars of the various compositions were tested at room and ele- vated temperatures (Table 6.3). At all temperatures TABLE 6.2, ZIRCONIUM CONTENT OF IRON-ZIRCONIUM ALLOYS Zirconium Content (wt %) Ingot No. Top of Ingot Bottom of Ingot 52741 1.02 . 1.07 -3 3.51 3.30 -5 4.95 5.70 -8 8.36 8.93 -12 11.60 13.18 -16 14.40 14.50 164 TABLE 6.3. TENSILE TESTS OF IRON-ZIRCONIUM ALLOYS AS EXTRUDED AT 2100°F Alloy Testing Tensile Elongation Temperature Strength No. o ] (%) ("F) (psi) 527-1 Room 57,300 32.5 1300 10,500 80.0 1500 5,100 93.7 1650 5,800 13.8 -3 Room 64,500 21.3 1300 19,400 50.0 1500 7,900 75.0 1650 4,600 100.0 -5 Room 66,500 18.8 1300 24,900 55.0 1500 7,000 100.0 1650 5,100 102.0 -8 Room 89,600 2.5 1300 30,800 58.7 1500 11,900 68.8 1650 6,700 121.0 these alloys showed an increase in tensile strength with increased amounts of zirconium. A corre- sponding decrease in ductility accompanied the increase in strength, Data were not obtained on ingot 527-12 because the extruded rad was brittle and could not be machined, The data indicate that alloys with up to 12% zirconium can be formed while hot, and that alloys with 8%, or less, zir- conium can be formed at room temperature, NONDESTRUCTIVE TESTING R. B. Oliver J. W. Allen K. W. Reker R. W. McClung Metallurgy Division The development of the probe-coil eddy-current equipment for the inspection of small-diameter Inconel tubing is essentially complete. The in- strument is shown in Fig. 6.41 in simple block- diagram form. The exciting coil of the probe is supplied with a constant 80-kc current by a crystal-controtled oscillator-amplifier. The fre- quency of 80 kc was selected as one sufficiently high to induce a relatively large amount of eddy- current flow in the tube wall and sufficiently low o mm ek QSCILLATOR e AMPLIFIER - CBISIESINI%E’ A (A—-B) ————-@—————‘ AMPLIFIER PRCBE L—— EDDY CURRENTS PICKUP con_fl,[-—-—4——_——i- j-l— EXCITING COIL PERIOD ENDING DECEMBER 10, 1955 UNCLASSIFIED ORNL—LR—0OWG 11542 FLAW INDICAT!% TUBING PERSISTENT — SCREEN CATHODE -RAY QSCILLOSCOPE - DATA POTENTIOMETER / —— - d TUBING - ROTATION Fig. 6.41. Diagram of System for Eddy-Current Inspection of Tubing. b in. of Inconel or less. The pickup-coil voltage is changed in amplitude and phase by the changes in amplitude and distribution of the eddy current in the wall of the tubing. The quiescent portion of the pickup- coil voltage is ‘‘bucked out’* by the balance circuitry, and hence only the signal changes are to ensure effective penetration through measured., Since these changes are very small, amplification of 104 to 10° is necessary before they are fed to the cathode-ray oscilloscope for interpretation, It is important to note that, since a probe type of coil is utilized, the same frequency may be used on tubing of various outside diameters. Also, since the operating frequency is held constant, the sensivitity of the system remains constant even though the outside diameter of the tubing is varied. This system is in contrast to the system that uses an encircling coil in which the frequency of the encircling coil must be made to vary inversely with the square of the outside diameter of the tubing in order to maintain comparable sensitivities for various diameters of tubing. The operation of an eddy-current instrument at a constant frequency is vastly simpler than operation of one at a variable frequency. The amplified changes in pickup-coil voltage are displayed on a persistent-screen cathode-ray oscilloscope as a function of the tubing rotational angle. Because of the localized nature of flaws in tubing, the flaws cause abrupt changes in signal as they pass under the probe coil. Other variables in the tubing which affect the coil, such as slight changes in wall thickness, conductivity, and permeability, are not a function of the tubing ro- tational angle and hence are indicated by a gradual broadening of the cathode-ray-oscilloscope picture while the tubing is spiraling past. Eccentricity of the tubing causes a gradual change in signal amplitude as the tube is rotated under the probe and produces a wedge-shaped picture on the os- cilloscope. This type of data presentation allows positive, rapid interpretation of the flaw signals at high scanning speeds. Considerable effort has gone into the develop- ment of a probe that will allow a reasonably high scanning rate and, at the same time, permit minute flaws to be detected with ease. The smallest and 165 ANP PROJECT PROGRESS REPORT most sensitive probe made to date has an effective diameter of ¥, in,, and it was able to detect small cracks and pin holes, With this probe it was found that the tubing could be rotated at 300 to 400 rpm without loss in definition of signals, Because of the small diameter the longitudinal speed of the probe, even at 300 to 400 rpm, is quantity Therefore elongated probe coils are now being tested, and it is thought that a probe coil ]/2 in, long and ]/16 in, wide can be utilized without compromising the sensitivity to small flaws. A probe ]/2 in. long will allow scanning rates of 150 in./min, very slow for inspection, The Cyclograph is also proving to be an extremely useful instrument for cursory inspection of small- diameter Inconel and Hastelloy tubing., It is being used extensively to sort out sections of Hastelloy B tubing that contain large flaws, that is, flaws which penetrate 10%, or more, of the wall thickness, prior to more detailed inspection by other methods. With the use of the Cyclograph it has been possible to inspect all the Hastelloy B tubing and to reduce substantially the inspection time and cost, The Cyclograph indications obtained, to date, in the inspection of much of the small-diameter Inconel tubing have shown cyclic variation over the length of the tubing. The variations were, at first, thought to be the result of very small changes in wall thickness that were not detected by any other method. Subsequent metallographic examinations of representative sections located by these varia- tions in the indications revealed intergranular attack on the inside wall of the tubing. The attack penetrated as deep as 0.002 in. A similar, eyclic- defect signal has revealed, in the tube wall, radial cracks which had been sealed over by the grinding operation used to produce the finished surface. The use of ultrasound as a tool for the inspection of small-diameter tubing also appears to be very practical. The “B’’ scan® which presents the defect signal on a cathode-ray tube as a function of the tube rotation has been found to be the most satisfactory method for interpreting ultrasonic data (Fig. 6.42). A pilot model of the equipment for uitrasonic scanning, which will handle tubing in lengths up to 7 ft 6 in,, has been set up in a large tank and is now ready for operation. The mechanical design principles are very similar to those of the large- SR. B. Oliver et al., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 139. 166 scale production inspection unit now being fabri- cated, which will accommodate tubing in lengths up to approximately 28 ft. Experience has indicated that the position or alignment of the ultrasonic transducer with reference to the tubing being inspected does not critically affect the flaw indication. Thus the misalignment caused by the characteristic whipping and wobbling which occur when long lengths of thin-walled tubing are rotated at 200 to 400 rpm will not create diffi- culties. It appears that routine ultrasonic inspec- tion of tubing can be accomplished by laboratory- trained personnel with a minimum of setup and adjustment time. For high-speed production inspection, the speed with which the length of tubing can be scanned is limited by the length of tubing which may be in- spected during one revolution of the tube. The ratio of longitudinal scan movement per tube revolution must therefore be adjusted to ensure pitch complete inspection coverage. To increase the pitch, the %-in.-diq transducer was replaced with a 341-in.-dic| transducer. However, the %-in.-dia transducer covered the entire cross section of the V-in. tubing being inspected and produced a con- fused cathode-ray-tube pattern when the full 3/4-in. diameter In order to overcome this difficulty and yet maintain the greater pitch ratio, a collimator was fabricated that effectively blocked out approximately one-third of the transducer energy was used, pattern along a chord parallel to the tubing axis (Fig. 6.43). With the collimation a pitch of I/2 in, per revolution can be maintained. Correlation of known defects, as detected by visual, dye-penetrant, Cyclograph, and x-ray in- spection, has shown that all the defects are de- tectable by ultrasound inspection, including those those on the inside wall, as well as those on the outside wall, of the tubing. Investigations are continving in an effort to determine the smallest size of discontinuity that may be detected, but it is now believed that any detrimental defect can be found. Work is under way on the development of an audible warning system to supplement the con- ventional *“B’' scan cathode-ray-tube presentation. The warning circuit presently available has too slow a response rate to be adequate for the pro- posed rotation and scanning speeds. PERIOD ENDING DECEMBER 10, 1955 UNCLASSIFIED ORNL-LR-DWG 11543 TIMING SIGNAL REFLECTION FROM TUBING SURFACE —f——— REFLECT!ION FROM FLAW VIDEO SIGNAL TUBING SURFACE REFLECTION r— ———————————————— _ = I I | | | | ! | I I | <+ | | | ' | ' | [ | | i | Lo I'— ______________________________ T | FLAW REFLECTION | | | - | L] } | TIMER | | - {(FREE—-RUNNING | } MULTIVIBRATORY} || l | | 4 | i | | A | [ | | | | 1 RECEIVER THYR ‘I (TUNED TO CRYSTAL PUC\‘ST?F?N : | RESONANT FREQUENCY) | [ | | | | | | - -— | | | -y A IMMERSCOPE 1( j\_ _______ J 17-in. PERSISTENT SCREEN CATHODE-RAY OSCILLOSCOPE "B" SCAN ROTATIONAL SIGNAL el DATA POTENTIOMETER TUBING Fig. 6.42. Diagram of System for Ultrasonic Inspection of Tubing. The use of ultrasound techniques for inspecting welded Inconel plates is being investigated. To date, difficulty has been experienced in passing sound through the weld bead, and it appears that conditions within the weld bead are such that the sound energy is almost completely attenuated. The conditions that cause the difficulty may be excessive grain size, microporosity, or included foreign matter, An x-ray technique for the inspection of thick sections of beryllium is also being developed. Beryllium, which is a very light element, has a very large scatter coefficient, which makes it a difficult subject for high-resolution radiography. It is believed that collimation of the x-ray beam at the front surface of the object with a slit system will reduce the scatter and increase the contrast. The object and the film will be exposed by moving them under the slit (Fig. 6.44). CERAMIC RESEARCH C. E. Curtis J. R. Johnson J. A, Griffin A, J. Taylor Metallurgy Division Rare-Earth Cermet Fabrication A rare-earth cermet has been fabricated from Lindsay oxide code 920 and iron that is 40 vol % rare-earth oxide and 60 vol % iron, with 50% of the iron present as an equivalent amount of Fe,0,. A measured density of 6.14 g/cm?® was attained, which is to be compared with o theoretical density 167 ANP PROJECT PROGRESS REPORT 7 COLLIMATOR UNCLASSIFIED ORNL—LR—DWG 11544 —a————— ULTRASOUND TRANSDUCER COLLIMATOR TUBE N (= ROTATICN OF TuBE 3 ~ - 7 — = = / —— Fig. 6.43. Collimator Arrangement for Use with 3/“-in.-diu Ultrasound Transducer. UNCLASSFIED ORNL-LR~-DWG 11545 X -RAY TUBE 4/ !‘ /SLIT SYSTEM IO A NS\ —— MOTION —— BERYLLIUM OBJECT O\gROLLERSO \-FILM O Fig. 6.44. Slit System for X-ray Inspection of Thick Sections of Beryllium. 168 of 7.69 g/cm3. The apparent porosity (kerosene) was 21.4%, and the porosity obtained from the density valve was 20.8%. Metallographic examina- tion showed the structure to be uniform. The ductility was estimated to be considerably less than that of previously prepared specimens that contained 70 vol % iron, as the metal. Boron Carbide Shield Material Methods for fabricating the boron carbide tiles needed for shielding in the ART are being investi- gated. It is planned to produce metal cans in the shape of large tiles which would be vibration-packed with sized B, C grains, The cans would have lap joints at their edges to prevent neutrons from streaming between the cans. Experiments are under way to determine the best packing method and the best sizing of the grains to obtain maximum density. In addition, a study was made to determine the tendency of B4C particles to flow under pres- e sure. The B,C particles were packed in a brass sleeve (10-mil wall) which had steel plungers on each end, Pressure was applied to the plungers, and the bulge of the sleeve was taken as a measure PERIOD ENDING DECEMBER 10, 1955 of the flow of the B,C particles, Bingham-type flow was observed, with a yield point of gbout 60 to 80 psi for the least dense powder and 120 to 140 psi for the most dense powder. 169 ANP PROJECT PROGRESS REPORT 7. HEAT TRANSFER AND PHYSICAL PROPERTIES H. F. Poppendiek Reactor Experimental Engineering Division Heat-transfer studies with the fuel mixture NaF- KF.LiF-UF , flowing in a heated Hastelloy B tube and the construction and assembly of a heat-trans- fer system that includes a pump were completed. The ART fuelsto-NaK heat exchanger test system was modified for further experiments, and the fric- tion characteristics of heat exchanger tubing were determined. Hydrodynamic studies of the ART core have been continued, and a study of flow in the sodium cooling passages around the core was initiated. Detail designs were completed for the experi- mental study of the temperature structure to be expected in the ART core, and fabrication of the apparatus was started. An analytical study is under way of the temperature structure in a volume- heat-source system in which the fluid is flowing turbulently between curved channels. A general study of the problem of heat removal from the fuel dump tank of an ART-type aircraft reactor after shutdown was initiated, The prediction of tem- perature distribution in helical pipes was corre- lated with data obtained from operation of the MTR in< UNCLASSIFIED ORNL-LR-DWG 10822 v ] — — GM COUNTER / FLOWMETER ;fl ~\ |.>fl |_—— FREEZING SOLUTION Fig. 8.1. Apparatus for Investigating Holdup Times of Fission Gases by Charcoal Traps. CREEP AND STRESS CORROSION J. C. Wilson Solid State Division In-Pile Tube-Burst Creep Tests W. W. Davis N. E. Hinkle J. C., Zukas Solid State Division The first in-pile, tube-burst, creep-test specimens were irradiated for two weeks in the LITR. In these tests four tubular specimens were stressed to rupture at a circumferential fiber stress of 2000 psi at temperatures of 1500°F for two and 1550 and 1450°F for the other two. Postirradia- tion measurements of the total strain in the uni- formly bulged regions are under way in the hot cells. Corresponding bench tests under the same conditions are in progress. A second in-pile rig is complete and ready for LITR irradiation. Additional fuel-containing tube-burst rigs, pre- viously described, have been assembled and are to 184 be filled with fused-salt fuel. Several stress- corrosion rigs in which static sodium is used as a coolant for a tubular fuel-containing specimen that is stressed in bending are ready for welding. These rigs will be used for control tests of a simi- lar apparatus already irradiated in the LITR, The MTR creep-test apparatus, in which the specimen was tested at 1500°F and 1500 psi, has been returned to ORNL, and the water jacket is being removed for postirradiation measurements of specimen elongation. Alternate In-Pile Apparatus W. E. Brundage C. D. Baumann Solid State Division An altemate apparatus for in-pile stress-corro- sion experiments has been designed. In this rig, as in the rigs now being used,® the molten fuel 3). C. Wilson et al,, ANP Quar. Prog. Rep. Sept 10, 1955, ORNL-1947, p 165. ORNL~-LR-DWG 10823 1000 . 500 TOTAL TIME TRAP WAS ACTIVE ~ 200 100 N O TIME (min) ™ o TIME TO PEAK ACTIVITY IN TRA < e —— TIME TO BREAK-THROUGH OF ACTIVITY 1 ‘ -130 —-110 -90 -70 —-50 -30 —~f0 +10 CHARCOAL BED TEMPERATURE (°C) Fig. 8.2. Holdup Time vs Temperature of Char coal Trap for Radiokrypton Carried in Nitrogen. contacts one wall of the annular container, and solidified fuel contacts the other wall. This permmits the heat generated to be removed and, at the same time, a small surface-to-volume ratio to be maintained and the maximum fuel temperature to be limited to that of the sample. The heat for maintaining the fuel in the liquid state will be supplied by fission heating during normal operation, but supplemental heating can be supplied by a cylindrical heater around the outside of the container if necessary. Heat will be re- moved by flowing water being passed through the central tube of the fuel container. Stress loading witl be supplied by gas pressure over the fuel, with the outer portion of the fuel container used as the specimen in hoop tension. PER!IOD ENDING DECEMBER 10, 1955 ORNL—LR-DWG 10824 10,000 5000 2000 : TOTAL TIME TRAP WAS ACTIVE 1000 S 500 200 TIME TO BREAK- THROUGH OF ACTIVITY [ B 100 TIME (min) 50 20 ———— e TIME TC PEAK ACTIVITY IN TRAP 10 . s = 1 -130 {10 =30 =70 =50 —-30 —10 +10 GHARCOAL BED TEMPERATURE (°C) Fig. 8.3. Holdup Time vs Temperature of Char- coal Trap for Radiokrypton Carried in Helium. Calculations have been made of the flux depres- sion outside the sample, the flux distribution in the fuel, and the resulting heat distribution through the fuel. The maximum specific power is generated in the outer portion of the annulus, with the power generated at the inner surface being ap- proximately 40% of the maximum, The flux.depres- sion calculations are being checked by flux meas- urements, in the LITR, with the use of a cadmium- magnesium alloy of the same size as the fuel volume and with a thermal-neutron cross section the same as that of the fuel to be used. 185 ANP PROJECT PROGRESS REPORT L. 9. ANALYTICAL CHEMISTRY OF REACTOR MATERIALS C. D. Susano J. C. White Analytical Chemistry Division An apparatus was modified so that the r-butyl bromide method could be applied to the determina- tion of oxygen in sodium sampled at operating temperatures. Studies were continued on the dis- tillation method for the determination of oxygen in sodium, A method for determining the concen- tration of rare-earth elements in stainless steel was developed. Aluminum was determined in NaF-ZrF4-UF4 by a modification of the Aluminon method, In connection with hydrofluorination studies a rapid method of analysis was developed for the determination of hydrogen fluoride and water in effluent gases. Studies were continued on the determination of oxygen in metallic oxides by the bromination method, A method was de- veloped for the determination of tantalum in NaF-KF-LiF. Methods were developed for the determination of ftrivalent iron fluoride salts. in mixtures of DETERMINATION OF OXYGEN IN SODIUM A. S. Meyer, Jr. G. Goldberg W. J. Ross Analytical Chemistry Division n-Butyl Bromide Method Studies were undertaken to ascertain whether positive errors are inherent in the determination of oxygen in sodium by the n-butyl bromide method,! Investigations of the errors introduced by the con- tamination by oxygen during sampling and by traces of water in the organic reagents were reported previously.2 Additional positive errors may result from the presence of alkaline-earth metals in the alkali-metal samples. The n-butyl bromide method is based on the con- version of the electropositive metals to neutral bromide salts by their reaction with butyl bromide and the subsequent acidimetric titration of the alkali-metal oxides. In the pure state the alkaline- earth metals calcium, barium, and strontivm are essentially inert to n-butyl bromide, while mag- ‘J. C. White, W. J. Ross, and R. Rowan, Jr., Anal Chem. 26, 210 (1954). 2p. S, Meyer, Jr., W. J. Ross, and G. Goldberg, ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 172. 186 nesium may react with n-butyl bromide to form a Grignard reagent, The alloys of these alkaline- earth metals with alkali metals are probably more to n-butyl bromide than are the pure alkaline-earth metals, and any residual metal or Grignard reagent will be converted to hydroxide upon dissolution of the reacted sample in water and will thus be titrated as oxygen. Since the lower-atomic-weight alkaline-earth metals in concentrations below the range of spec- trographic determination may be sufficient to intro- duce significant errors in the oxygen determination, chemical methods for their detection were tested. Calcium and magnesium were determined by flame- photometric measurement after separation from the bulk of the sodium metal. Successful separa- tions were carried out either by precipitation of reactive the calcium and magnesium from ammoniacal solu- tions with 8-quinclinel or by precipitation of the sodium as sodium chloride by saturating aqueous solutions of the samples with HCl gas. After separation by either of these methods, the calcium and magnesium concentrates from a 10-g sample contained less than 10 mg of sodium, Typical samples of the sodium used in corrosion and heat-transfer studies were found to contain approximately 100 ppm of calcium and less than 50 ppm of magnesium. These concentrations corre- spond to a maximum error in the determination of oxygen of about 80 ppm, and appropriate correc- tions will have to be determined. Since the alkaline-earth elements that are alloyed with sodium are at least partially converted to bromide salts during dissolution of the samples, the error introduced by them is not significant if the con- centration of oxygen in the sodium is high. The increase in the reactivity of the alkaline-earth metals upon dissolution in sodium was demonstrated by the determination of oxygen in sodium which contained relatively high concentrations of altkaline- earth metals, for example, 0.2% barium. Since the molar concentration of oxygen as determined by the n-butyl bromide method was less than that of the alkaline-earth metal, a portion of the metal must have reacted with the n-butyl bromide. In further studies of the n-butyl bromide method it has been shown that, if the reaction is contained in a sealed vessel, the conversion of the alkali metals to bromide salts can be carried out by the reaction of the sample with pure n-butyl bromide rather than with the moderated solution of n-butyl bromide in hexane, which is used in the usual ana- lytical reagent. The reaction time is reduced from about 4 hr to less than 5 min when the pure reagent is used. The procedure is applicable both to NaK and to sodium, When the reaction is carried out with the pure reagent under pressure, the possibility of con- tamination during reaction is materially reduced and it is possible to effect the direct transfer of the test portion of the sample from the bulk sample. A sampler and on apparatus for the defermination of oxygen in sodium by the amalgamation pro- cedure have been modified, as shown in Fig, 9.1, for use with n-butyl bromide, In this apparatus a sample of molten metal is added, from a transfer line, to approximately 150 ml of n-butyl bromide reagent contained in a 3 x 18 in., glass reaction vessel. The transfer is carried out under helium at a pressure of 1 atm. Samples have been taken from sodium at temperatures as high as 1000°F. A pressure below 10 psi was developed even when samples which contained 9 g of sodium were re- acted. This procedure is now being used to com- pare the results of determinations of oxygen by the n-butyl bromide method and by the distillation method, Distillation Method The distillation method for the determination of oxygen in sodium, which was described in the previous report,® was applied to the determination of the oxygen in sodium contained in forced-circu- lation corrosion-testing loops and in sodium-transfer reservoirs. Samplings have been made at tempera- tures from 600 to 1400°F. For the oxygen determination the sample is introduced through the top of the distillation appo- ratus into a calibrated, hemispherical cup which is fitted with a preshaped liner of 1-mil nickel foil. A sufficient quantity of the liquid sodium is intro- duced to flush the transfer line and the sample cup thoroughly. When the temperature of the sodium is less than 600°F, the distillation apparatus is evacuated before the sample is transferred. At higher temperatures, a pressure of 1 atm of helium must be maintained to prevent flash distillation of 31bid., p 173. PERIOD ENDING DECEMBER 10, 1955 the sodium during transfer. After the cup has been filled with the test portion of liquid metal, the pressure in the distillation chamber is reduced to less than 10 y, and the sample is heated to distill off the metallic sodium. After the distillation is completed, the nickel liner is transferred to a beaker, in which the residual oxides are titrated with a dilute solution of standardized acid. In a check of the effect of time and temperature made by the distillation method, analyses were made of sodium which had been deliberately contaminated by the addition of an amount of Na O, equivalent to 700 ppm of oxygen before introduction into the corrosion-testing loop. When the distillation was carried out at above 900°F, that is, near the maximum of the recom- mended4 range of distillation temperatures of 350 to 550°C (660 to 1030°F), the oxygen values were extremely low., The effect of distillation time was also evident; for example, after a 1-hr distillation period at 900°F, the oxygen concentration of the residue was less than 100 ppm, and after 4 hr the oxides had been almost completely volatilized. Distillation of another sample of the contaminated sodium at 850°F for 4 hr gave a residue which corresponded to 400 ppm of oxygen. Since part of the oxygen could have been trapped in cold legs of the loop, the 400-ppm value represents a reason- able analysis of the oxygen in the sample that was originally contaminated with 700 ppm of oxygen. on analyses The accuracy of the distillation method is being studied by comparing the results with those ob- tained by the n-butyl bromide method and by testing the quantative recovery of added oxygen. A distillation apparatus and an apparatus for analyses by the n-butyl bromide method have been attached to a reservoir which contains sodium that was deliberately contaminated by exposure to air during locding. Repeated determinations after the dissolution of samples of the sodium in the n-butyl bromide reagent have established the concentrgtion of oxygen as 600 * 50 ppm. An analysis by the distillation method showed 620 ppm of oxygen, In the study of the quantitative recovery of added oxygen, reproducible results which correspond to 35 = 5 ppm of oxygen were obtained after dis- tillation of as-received sodium for 4 hr at 850°F. Essentially the same results were obtained after 4). Rr. Humphries, personal communication, June 24, 1955. 187 ANP PROJECT PROGRESS REPORT distiflation for periods of 1 and 2 hr at this tem- perature. Analyses of the residues indicated that Na,O is the predominant residual oxide. The effect of metallic impurities in the liquid metal on the CALROD HEATER. ~ OVERFLOW TRAY>/ e Wt Gy "O" RING VACUUM GLASS PIPE FLANGE VALVE ~ GLASS PIPE— | A _ww OVERFLOW REACTION RECEIVER VESSEL distillation method is to be studied in detail, After the oxygen content of a measured quantity of sodium has been established, a calculated weight of NaOH will be added to raise the concentration of oxygen UNCLASSIFIED ORNL-LR-DWG 10825 SODIUM INLET HELIUM VIEWING WINDOW /A I S E BUTYL BROMIDE ANHYDRONE GLASS FRIT TEFLON GASKET 1 Q 1 2 \_/ SCALE IN INCHES Fig. 9.1. Apparatus for the Determination of Oxygen in Sodium by Use of the n-Butyl Bromide Method. 188 ek e o e to 300 ppm, and the determinations wiil be re- peated to test the accuracy of the distillation method. DETERMINATICN OF TRACES OF RARE-EARTH ELEMENTS IN STAINLESS STEELS A, S. Meyer, Jr. B. L. McDowell Analytical Chemistry Division The concentrations of the rare-earth elements gadolinium, europium, and samarium must be limited to a few parts per million in reactor con- struction materials because of the high absorption cross sections of these elements for thermal neu- trons. Therefore a method was developed for de- termining the quantities of the rare-earth elements in stainless steel, Inconel, and nickel, In order to concentrate the rare-earth elements in a sample, 25 mg of spectrographically pure yttrium oxide is added to a 2- to 5-g sample of stainless steel. The ytirium oxide is used as a carrier for the precipitation of the rare-earth elements and also serves as an internal standard for their spectrographic determination. The sample, fo- gether with the Y 0,, is dissolved in HCI and then fumed with Hé|04. After the hydrated oxide precipitate of silicon, niobium, tantalum, etc,, has been removed by filtration, the major con- stituents of the sample, iron, chromium, and nickel, are separated by electrolysis at a mercury cathode, The volume and acidity of the solution are reduced by evaporation, and the rare-earth elements, to- gether with the yttrium carrier, are precipitated as hydrous oxides by saturating the solution with ammonia, The alkaline-earth elements remain in solution, together with a portion of the amphoteric elements, The precipitate is dissolved in HCI, and the rare-earth concenfrate is precipitated by the addition of an oxalic acid. The precipitate is filtered and then ignited to the oxide, The concentration of rare earths in the precipi- tated Y, 0, is determined spectrographically, and the concentration in the stainless steel sample is calculated from the ratio of carrier to sample. When 2.5-g samples of stainless steel are taken, concentrations of the rare-earth elements of 1 ppm can be determined, Between 75 and 90% of the added carrier is recovered. The final precipitate is essentially pure Y203, with SiO2 as the prin- cipal contaminant, The method offers several advantages over the procedure of Spitz et al.,® in which the rare-earth PERIOD ENDING DECEMBER 10, 1955 group is precipitated with an iron carrier and measured against uranium as an internal standard. Fewer precipitations are required, and, since the carrier is added before the separation step, quanti- tative recovery of the rare earths is not essential, The Y, 0, provides a much less complex spectrum than that provided by the uranium and iron, The method has been used for the determination of gadolinium in types 347 and 304 stainless steel, Quantitative recovery of a standard addition of gadolinium was obtained. The method will also be tested with other types of stainless steel, Inconel, and nickel. With minor modifications the procedure should be applicable to all ferrous and nonferrous alloys that do not contain as major constituents elements such as zirconium, vanadium, and other metals which would not be deposited at the mercury cathode, SPECTROPHOTOMETRIC DETERMINATION OF ALUMINUM IN FLUORIDE SALTS WITH AURIN TRICARBOXYLIC ACID A. S. Meyer, Jr. C. R, Williams Analytical Chemistry Division The Aluminon (aurin tricarboxylic acid) method for the spectrophotometric determination of alumi- num was applied to the determination of aluminum in fluoride salt mixtures such as NaF-ZrF ,-UF ,. Aluminon forms a stable, red-colored lake with solutions of pH 5 to 5.4. While this reaction is utilized as one of the more sensitive colorimetric methods for the detection and determination of aluminum, it is subject to interference by many elements,® particularly those which form hydrous oxide precipitates at pH 5. In the determination of aluminum in samples of NaF-Zer-UFA‘, the principal interfering element, zirconium, is removed by extracting it as zirconium cupferrate (nitrosophenylhydroxylamine) from a 4 M solution of H_SO, with chloroform containing hy- drogen cupferrate. Other interfering elements, such as iron, which is present in trace concen- trations in these samples, are also removed by this extraction. Any uranium which may be present in the tetravalent state is quantitatively extracted, while hexavalent uranium remains in the aqueous phase, Since only a small part of the uranium is 5E. W. Spitz et al., Anal. Chem. 26, 304 (1954). 6F. D. Snell and C. T. Snell, Colorimetric Methods of fl‘lnalysis, 3d ed., vol Il, p 248, Van Nostrand, New York, 949. 189 ANP PROJECT PROGRESS REPORT oxidized to the hexavalent state during dissolution with H2504, its residual concentration is not suffi- cient to interfere in the determination of aluminum. When cupferron is added to acidic solutions which contain zirconium, a very slightly soluble precipi- tate of zirconium cupferrate is obtained. Since this precipitate is not readily extractable by organic solvents,” the zirconium cupferrate is formed in the organic phase by extracting the aqueous solu- tions with a solution of hydrogen cupferrate in chloroform., Aluminum is not extracted from a 4 M H,S0, solution if the concentration of the hydrogen cupferrate in the chloroform phase does not exceed 3%. An aliquot of the extracted solution that con- tains 6 to 50 ug of aluminum is taken for color development. The color is developed by heating the solution with Aluminon for ¥ hr at a tempera- ture of 80 to 90°C; gelatin is added to stabilize the lake. Mercapto acetic acid is added to complex any iron that is introduced in the reagents which are added after the cupferron extraction. The absorbance of the final solution of 50-ml volume is measured at 530 my in l-cm cuvettes 2 hr after the addition of the chromogenic reagent. The number of determinations made by this methed is not yet sufficient to establish the pre- cision, but, on the basis of calibration datq, it appears that the coefficient of variation of the results is approximately 5%. The concentration of aluminum in typical samples of NaF-ZrF , eutectic mixtures and NaF-ZrF -UF, (50-46-4 mole %) fuels was found to be approximately 100 ppm. DETERMINATION OF WATER IN HYDROGEN FLUORIDE GAS A. S. Meyer, Jr. W. J. Ross Analytical Chemistry Division A method was developed for the determination of the concentration of water relative to hydrogen fluoride in the effluent gases from a hydrofluorina- tion reactor. The measurement of the relative water content was needed to ascertain when the hydrofiuorination reaction of oxides in zirconium- base fuels, that is, ZrO2 + 4HF —> Zrl":4 + 2H20 was complete. Pyridine was found to be an effi- cient absorber for both hydrogen fluoride and water, ’N. H. Furman, W. B. Mason, and J. 5. Pekola, Anal. Chem. 21, 1327 (1949). 190 The gases from the reactor were bubbled through approximately 5 cm of pyridine contained in a simple absorber constructed of 30- by 200-mm test tubes. Two absorbers were connected in series to measure the efficiency of the absorption. After a quantity of gas had been passed through the pyridine, the test tubes were removed from the assembly, stoppered immediately with serum-type bottle stoppers, and replaced with tubes filled with fresh pyridine, Test portions of the absorber solutions were taken with 1-ml| tuberculin syringes. The concentration of water in the pyridine solu- tions was determined by titration with coulometri- cally generated Karl Fischer reagent.® The con- centration of hydrogen fluoride was determined by titrating an aqueous solution of the pyridine absorbate with 0.1 N NaOH solution to a phenol- phthalein—thymol-blue mixed-indicator end point. Since pyridine is a relatively weak base (K, = 1.8 x 10-9), the hydrogen fluoride can be titrated directly in aqueous solutions which contain pyri- dine. The hydrogen fluoride and water were almost completely removed from the gases in the first absorber., When the concentration of hydrogen fluoride in the first absorber did not exceed 50 mg/ml, the concentration of hydrogen fluoride in the second absorber was negligible. Increases in the concentration of water in the second ab- sorber were limited to a few tenths of a milligram per milliliter and may have been introduced by contamination with atmospheric moisture during loading of the pyridine into the test tube. The determinations made by this method are reproducible to about +0,5% water in hydrogen fluoride. Since in the concentration of water from 1% in commercial hydrogen fluoride to 10 to 30% in the effluent gases were observed during test runs, the reproducibility of the determinations is suffi- ciently high to permit monitoring the reaction. The over-all precision of the resulis is limited primarily by the precision of the determination of the water, which is approximately 2%. The pre- cision of measurement of small increases in the concentration of water is much lower because of a relatively high original concentration of water in the pyridine. The precision of the method can be improved by the use of anhydrous pyridine. increases 84, s. Meyer, Jr., and C. M, Boyd, The Determination of W)ater by Coulometric Titration, ORNL-1899 (June 5, 1955). DETERMINATION OF OXYGEN IN ZIRCONIUM OXIDE BY BROMINATION J. P. Young M. A. Marler Analytical Chemistry Division The study of the methods of determination of combined oxygen in metal oxides and fluoride salts by the bromination method of Codell and Norwitz? was continued. The reactions involved in the process include the reaction at approximately 1000°C of bromine vapor in a carrier gas of helium with a mixture of a metal oxide and powdered graphite to form CO and the metal bromide. After removal of the excess bromine and the metal bromide in cold traps, the CO is oxidized to CO by reaction with hot CuO. The resultant CO2 is absorbed in Ba(OH), and estimated quantita- tively by titration with standard acid. Several additions and modifications were made to the apparatus, which is described in previous reports. 1011 A tube containing copper and copper oxide, which is maintained at 400°C, was inserted between the helium tank and the apparatus in order to remove any oxygen or hydrocarbons that might be present in the gas. Also, the Tygon tubing that had been used to comnect some parts of the apparatus was replaced with glass and short pieces of gum-rubber tubing. As recommended by Codell and Norwitz,? granular zinc maintained at 350°C was inserted between the cold traps and the CuO tube to remove the last traces of bromine that might escape the cold traps. Further, the thin, platinum liner, described in the previous report,12 which is inserted inside the reaction tube during the analysis of fluoride salts, was replaced with a geld liner that extends into the first cold trap. Although gold is attacked by bromine, it is more inert to oxygen than is platinum, and therefore lower blank corrections for the appa- ratus are required. The sample container has also been fabricated from gold. For this determination, mixtures of ZrO, and graphite are ground together in a mortar by use of a pestle and then allowed to react with bromine M. Codell and G. Norwitz, Anal. Chem. 27, 1083 (1955). 10y C. White, J. P. Young, and G. Goldberg, ANP Quar. Prog., Rep. March 10, 1955, ORNL-1864, p 161. 1y, p, Young and G, Goldberg, ANP Quar. Prog. Rep. june 10, 1955, ORNL-1896, p 178. 12, P Young and M. A. Marler, ANP Quar Prog. Rep. Sept. 10, 1955, ORNL-1947, p 176. PERIOD ENDING DECEMBER 10, 1955 vapor under varying conditions. The residual amount of ZrO, is determined by weighing the residue after ignition of the graphite. Bromination of Zr0,, according to the equation ZrD2 + QBr2 + 20— ZrBr4 + 2CO was incompiete after 4 hr of reaction time at tem- peratures as high as 1100°C. Approximately 20% of the ZrO, remained unreacted under these con- ditions. When 30 mg of FeF, was added to 10 mg of ZrO, and 50 mg of graphite and the intimate mix- ture was brominated at 950°C for 2 hr, complete removal of ZrO, from the sample container was achieved, The order of mixing was found to be critical. In order to achieve the complete removal of ZrOz, it was necessary to mix the ZrO, with FeF, in a mortar with a pestle, and then mix the graphite with the ZrO -FeF, mixture in the same manner., |t is assumed that the reactions involved are: 3Zr02 + 4FeF3—> 3ZrF4 + 2Fe203 F3203 + 3Br2 + 3C — 2FeBr3 + 3CO The amount of CO formed during the bromination of the sample was, however, much less than the expected theoretical amount, The CO was de- termined by oxidizing the gas to CO, and absorbing the CO, in standard Ba(OH), solution. In all determinations of the resultant CO, the amount was only 30 to 40% of that expected. In order to check the possibility that the method of absorbing CO, was not efficient, several samples of CaCO, were thermally decomposed in the apparatus and the resulting CO, was determined. Quantitative recovery of CO, was achieved in all cases. The study on the bromination of ZrO, and FeF, in the presence of graphite is still in progress, but, at present, it appears that the reactions postulated do not adequately describe the mechanism of the removal of ZrO_ from the sample container, In order to establish the quantitative recovery of the CO produced in the bromination of the oxide, the determination of oxygen in ferric oxide by bromination was investigated. Quantitative re- covery of the oxygen as CO, was obtained, but it was necessary fto mix the ferric oxide and graphite in a mortar with a pestle. Bromination of a 15-mg sample of ferric oxide was complete in 4 hr at a temperature of 950°C. 191 ANP PROJECT PROGRESS REPORT DETERMINATION OF TANTALUM IN FUSED FLUORIDE SALTS J. P. Young J. R. French Analytical Chemistry Division A search was made for a suitable colorimetric method for the determination of trace amounts of in NaF-KF-LiF and NaF-KF-LiF-UF . Dinnin'3 has reported the use of pyrogallol, which forms a colored complex with tantalum, but uranium and fluoride ions are known to interfere in this determination, A satisfactory modification of Dinnin's method has been developed, however, for application to the fuel solvent NaF-KF-LiF. Preliminary tests revealed that the usual method of removing fluoride ion by heating in an H,SO, solution was not satisfactory, since tantalum was volatitized as TaF_. In order to avoid this diffi- culty, the sample was digested in 0.1 M H_SO until almost all the water had volatilized; another volume of 0.1 M H2SO4 was then added, and the solution was evaporated to dryness, |n this method, TaF, is hydrolyzed to Ta,0, before the tempera- ture of the solution exceeds the volatilization temperature of TaF _. tion, including the tantalum The residue of the evapora- 020 , is then treated according to the procedure described by Dinninl3 for the colorimetric determination of tantalum with pyro- gallol. The residue is fused with K25207, and the melt is dissolved in ammonium oxalate. The pyrogallol complex is formed in a solution that is 0.175 M ammonium oxalate and 4 M HCI, and its absorbance is then measured at 330 mp. This method was found to be satisfactory for the concentration range of from 0.15 to 0.8 mg of tantalum in a 50-ml volume. It is essential that the H 504 evaporation and the K,$,0, fusion be made in Vycor crucibles rather than platinum, since even very small quanti- ties of platinum interfere with the pyrogailo} method for determining tantalum. The attack of fluoride ion on Vycor has not been of any consequence. The method of Milner, Barnett, and Smales, 4 which was developed for the separation of tantalum in tantalum-uranium alloys, was examined for pos- sible application to samples of fluoride salts which contain uranium, |n this method tantalum is ex- tracted with hexone from an aqueous solution 134, 1. Dinnin, Anal Chem. 25, 1803 (1953). s, w. C. Milner, G. A, Barnett, and A, A, Smales, Analyst 80, 380 (1955). 192 which is 10 M in HF, 6 M in H2$O4, and 2,2 M in NH4F. The sample is first fumed with H2SO4, and then HF and NH ,F are added. The tantalum is extracted into hexone and then extracted from the hexone phase with a 15% solution of H,0,. The efficiency of this extraction was evaluated by determining the amount of tantalum which had been extracted in the solution of H,O,. The pyrogallol method, described above, was used for these analyses. Quantitative recovery of small amounts of tantalum has not been achieved, as yet. DIRECT DETERMINATION OF TRACES OF Fe(lll) IN NuF»KF-LiF-UF4 J. C. White Jo W, Miles!? Analytical Chemistry Division A direct spectrophotometric determination of Fe(lll) in Nc:l"'-KF-LiF-UF4 was developed in which the sample is dissolved in a solution of H3P04, H,50,, and H,BO,. Potassium thiocyanate is added to form the iron(}l)-thiocyanate complex, which is then extracted with isobutyl alcohol. The H,PO, concentration must be maintained above 8 M to prevent reduction of the Fe(ill) by U(IV) during dissolution of the sample. The coefficient of variation for the method was found to be 3% for the range 0.25 to 1.5 mg of Fe(lll) per gram of sample. Further details of this investigation are available in a separate report.!é DETERMINATION OF TRACES OF Fe(lil) IN MIXTURES OF ALKALI-METAL FLUORIDE SALTS J. W. Miles J. C. White Analytical Chemistry Division A method, which is based on the absorption of the ferric phosphate complex in the ultraviolet, has been developed for the determination of Fe(lll) in the presence of Fe(ll) in mixtures of alkali- metal fluoride salts. Iron(l}l}) in concentrations of 1 to 5 pug/ml in 0.5 M H2504 and 1 M H,PO, solu- tions may be determined accurately by observing the absorbance at 260 mu. The alkali metals, Ni(il), GCe(Hl), Fe(ll), ond the fluorides do not interfere, but zirconium and uranium interfere ]SResearch participant, University of Medical School, Louisville, Ky. 16, c. White, Determination of Traces of Iron(lll) in NaF-LiF-KF-UF4 with Thiocyanate, ORNL CF-55-9-96 (Sept. 20, 1955). Kentucky seriously, The details of this method have also been reported separately,17 ANP SERVICE LABORATORY W. F. Vaughan Analytical Chemistry Division Uranium analyses were performed for the high- temperature ART critical experiment. This service was provided so that other experimenters could follow the increase in uranium concentration as criticality was approached by the addition of the fuel concentrate, N02UF6, to the fuel carrier, NaF-ZrF, (50-50 mole %). The samples, weighing approximately 60 g each, were ground in a plastic dry box prior to analysis. Test portions (1 g) were fused with K;5,0,, and the melt was dissolved in 5 vol % H2504. The entire portion was passed through a zinc reductor, and the vranium was titrated with standard oxidant. Appropriate correc- 17J. C. White, Determination of Iron(Ill) in Mixtures of Alkali Metal Fluoride Salt, ORNL CF-55-7-103 {July 22, 1955). PERIOD ENDING DECEMBER 10, 1955 tions were made for the iron present, the concen- tration of which was determined colorimetrically. The elapsed time for analysis of each sample was of the order of 30 to 45 min. During the period, 1276 samples were analyzed for the ANP project, and a total of 6609 determina- tions was made. The backlog consists of 46 sam- ples. A breakdown of the work is shown in Table 9.1. TABLE 9.1. SUMMARY OF SERVICE ANALYSES REPORTED Number of Number of Samples Determinations Reactor Chemistry 746 3963 Experimental Engineering 487 2503 Miscellaneous 43 143 1276 6609 193 ANP PROJECT PROGRESS REPORT 10. RECOVERY AND REPROCESSING OF REACTOR FUEL F. R. Bruce D. E. Ferguson M. R. Bennett F. N. Browder G. |. Cathers W. K. Eister H. E, Goelier J. T. Long R. P. Milford S. H, Stainker Chemical Technology Division PILOT PLANT DESIGN Engineering design of the pilot plant for re- covering fused-salt fuels is .to be completed by December 31, 1955, and construction is to be completed by March 31, 1956, The ORNL Engi- neering Department has estimated that the facility will cost $346,000, Process, project, and instrumentation engineering is expected to raise the estimated over-all cost to $435,000. The single-bed absorber containing granular NaF which was to have been operated at 650°C has been replaced on the engineering flowsheet by a pair of absorbers! containing NaF, These ab- sorbers will both be at 100°C during absorption of UF , and some fission-product fluorides in the first bed, The beds will then be raised to 400°C, and the UF, and a few fission-product fluorides will be desorbed from the first bed with F_ and passed to the second absorber, where the fission products will remain. The large two-section furnace re- quired to heat the single-bed absorber was returned to the manufacturer for conversion into two separate furnaces. excluding all engineering. ENGINEERING DEVELOPMENTS Contactor A percolator type of contactor was found to pro- duce more violent agitation than that produced by a sieve-plate contactor during gas contacting of the molten fused salt in tests made in the fluori- nator designed for use in the pilot plant. In these tests, 270 b of molten NaF-ZrF, was sparged with nitrogen, and the percolator, or air-lift, type of contactor was operated at gas rates of 1 to 3.5 cfm. Visuval observations made during these tests showed violent pumping of the salt and a large amount of salt collecting on the cooler upper wall. The entrainment and subsequent deposition 1F. R. Bruce et al.,, ANP Quar, Prog. Rep. Sept. 10, 1955, 0RNL"]947, Figu }0.2, P ]800 194 of salt caused by percolation was greater than that caused by sieve-plate contacting. An inverted funnel-type splash plate installed over the perco- lator discharge contained the salt splash and thus reduced salt entrainment. An unpredictable in- crease in sparging-gas pressure during percolator tests was a major problem., Replacement of the ]/4-in.-dia tubing, which dispersed the gas through ]/16-in.—dic| holes, by a i/4--in.-dic:| open-end tube did not affect the pressure buildup. The sieve- plate contactor was operated at gas rates of | to 7 cfm. Salt agitation was gentle and there was a slight splash of salt on the vessel wall. Tests of these contactors with uranium-bearing fuel mixtures will be required in order for a decision to be made as to which to install in the pilot plant. Freeze Yalves A 4-in.-0D, 8-in.-long cone-body prototype of the freeze valves designed for the pilot plant re- tained a pressure of 20 psig in 15 of 20 cycles of melting (350°C), freezing, and pressurizing. Ex- amingtion of the salt seal around the downcomer revealed a porous structure but no open gas channels (Fig. 10.1). It is believed that slight contraction of the fused salt permitted gas leakage around the downcomer. The design of the valve is being modified. Resistance Heating of Transfer Pipes and Waste-Discharge Nozzle Resistance heating was tested as a means of heating and of maintaining the temperature of salt transfer pipes at 1200°F. An uninsulated 7-ft length of Y-in. sched-40 Inconel pipe was heated to 1200°F in 5 min by passing a current of 600 amp through the pipe. The voltage drop through the pipe was 1.2 v/ft, Resistance heating will also be used to heat the nozzle which will discharge salt from the fluori- nator waste pipe to the waste carrier., The nozzle is designed with a hood section to contain and LOWER CONE BODY : NICKEL BARSTOCK - NO DIRECT GAS CHANNEL HERE METAL @ CUTTINGS POROUS SALT SEAL, a4 t—in.~dic DOWNCOMER - AND DIAPHRAGM REMOVED =~ - THERMOWELL 1 Ya-in-dia NICKEL & SALT OUTLET PIPE !/o=in. SCH-40 INCONEL S%ST i PERIOD ENDING DECEMBER 10, 1955 UNCLASSIFIED | PHOTO 15686 UPPER DISENGAGING SECTION NICKEL PLATE PART OF _ DIAPHRAGM ~ ENTRANCE LIQUID » " LEVEL PROBE _ e SALT INLET PIPE Yo—in. SCH-40 INCONEL Fig. 10.1. Sectioned 4-in.-0OD, 8-in.-Long Cone-Body Freeze Valve After 20 Cycles of Melting (350°C), Freezing, and Pressurizing. remove any gases evolved from the waste (Fig. 10.2). The hood vacuum connection is to be nickel, and the salt transfer pipe and hood section are to be Inconel. These materials and the large cross- sectional area of the hood section will minimize self-resistance heating in the hood section and in the vacuum connection and will permit a tempera- ture of 1200°F to be reached by the salt transfer pipe. The voltage drop across the entire unit will be 4.25 v with a current of 600 amp. PROCESS DEVELOPMENT Fused-Salt Fluorination Studies Additional evidence of the induction period? in UF, volatilization from fused-salt fuel was ob- tained (Fig. 10.3) in further work on the fluorina- 2p, E. Ferguson et al., ANP Quar. Prog, Rep. March 10, 1955, ORNL-1864, p 164, tion step. This work was directed toward full evaluation of the effect on the fluorine efficiency of the fluorine-to-nitrogen mole ratio, the fluorina- tion rate, and the method of contact of the gas with the fused-salt phase. In Fig. 10.4 the same data are plotted to show how the efficiency of fluorine utilization varies during the course of fluorination and to illustrate the deviation from the ideal case, in which there would be no solu- bility effect to cause an induction period. A comparison of sampling methods showed that agreement to within 3% was obtained in uranium analyses of samples taken during the course of fluorination experiments with the use of a dip ladle or by immersion of a solid rod into the fused salt to obtain a quick-freeze sample and samples taken after fluorination by grinding and sampling the entire batch of salt, This study was made 195 ANP PROJECT PROGRESS REPORT SALT TRANSFER, Ya-in. SCH 40 INCONEL PIPE Yg-in. INCONEL‘L UNCLASSIFIED ORNL-LR-DWG 10826 Yo-in. SCH-10 NICKEL PIPE VACUUM CONNECTION N & 2in. HOOD SECTION T _ 1/8 in. o 5-in. DIA with three different uranium concentrations in the Nc:F-ZrF4 salt, namely, 8, 2, and 0.5 wt %. NaF Absorption Capacity and UF6 loss on Desorption The absorption capacity of NaF for UF6 was determined to be about 0.9 g of uranium per gram of salt in the case of one lot of Harshaw Chemical Co. material under a variety of conditions (Table 10.1) The capacity of Baker & Adamson Co. reagent-grade NaF was 1.89 ¢ of uranium per gram of salt. Absorption of UF‘5 on NaF under an initial vacuum indicated a capacity range of 0.80 to 0,90 for different mesh sizes, while absorption values obtained when the excess UF6 was being removed with fluorine as a sweep gas varied from 0.72 to 0.86. Practically the same capacity values were 196 ] Fig. 10.2. Waste-Discharge Nozzle. obtained at 70, 100, and 150°C and at two different pressures, The results indicated that the capacity of 0.9 in the case of the Harshaw Chemical Co. material is independent of the particle size to which it is degraded. The 1.89 value for the capacity of the Baker & Adamson Co. material corresponds closely to the molecular ratio in the complex UF6-3N0F. Fluorine was much more satisfactory as a sweep gas than was nitrogen in the desorption of UF from NaF, perhaps as a result of some moisture in the nitrogen, Two runs were made with 12- to 40-mesh NaF which had been well conditioned by being alternately evacuated and exposed to fluorine at a pressure of 15 psia before sufficient UF , was introduced to completely saturate the NaF. run the UF , was desorbed with a stream of nitrogen In one L] ORNL—LR~DWG 10827 100 50 20 URANIUM [N FUSED SALT (% OF INITIAL) 05 0.2 0 1 2 3 4 5 MOLE RATIO OF FLUORINE TO TOTAL URANIUM o4 Fig. 10.3. Amount of UF6 Remaining in 375 g of NaF-ZrF .UF , (50.-46-4 mole %) Fluorinated at 600°C at a Rate of 100 ml/min as a Function of Amount of Fluorine Introduced. while the temperature of the NaF was being raised from 100 to 400°C; 1.4% of the uranium stayed on the NaF, The total desorption period covered 1 hr, although the temperature of the bed had reached 400°C in gbout 15 min. In the second run the UF6 was desorbed with fluorine, and only 0.0075% of the initial uranium load remained on the NaF. UF‘5 Decontamination in NaF Absorption Step In each of four consecutive fluoride volatilization runs (20 to 40 g of uranium in each) in which the two-bed NaF decontamination system was used, the activity of the product UF, was less than the UX,-UX, activity associated with natural vranium. There was no evidence of decreased decontamina- tion with repeated re-use of the bed (Table 10.2). The over-all beta or gamma decontamination factor PERIOD ENDING DECEMBER 10, 1955 -rev ORNL-LR~—DWG 10828 1.0 B P\ “ ‘ Lt \ | | | 0.8 - *— — _ T ") T s _______1__\ [ R = \6 | R = 08 .LJKJ | ] " \ - | Lol N a i \\ ‘ | S oa N 5 * N 2N 5 ] 7\.! . Hop o’ \ T S * 0 L i o ! 2 3 4 5 6 MOLE RATIO OF FLUORINE TO TCTAL URANIUM Fig. 10.4. Efficiency of UF Volatilization as a Function of Amount of Fluorine Introduced. was about 10°, with 102 being attributable to the fused-salt fluorination step and 103 to the NaF absorption-desorption step. The process flowsheet used ! provides for volatilizing UF‘5 from the molten fused-fluoride-salt fuel with fluorine, absorbing the UF . on an NaF bed at 100°C, desorbing the UF6 at 100 to 400°C in a stream of fluorine, and passing the desorbed UFé through a second NaF bed. As in previous work, much of the decontami- nation from ruthenium occurred during the absorption period; that is, the UF , and niobium were absorbed on the first bed ot 100°C, while the ruthenium passed through (Table 10.3). Almost all the de- contamination from niobium occurred in the de- sorption period, when the UF , was being regener- ated from the NaF at 100 to 400°C; the niobium was left on the NaF. For all four runs the average uranium loss in the first cold trap of the process was 0.06%. This was perhaps due to the low NaF-to-uranium weight ratio of 2/1 in the first bed. The uranium loss in 197 ANP PROJECT PROGRESS REPORT & - TABLE 10.1, CAPACITY OF NeF .EOR UF6 Conditions i Capacity Material UF6 Pressure Temperature Removal of (g of U per g of NaF) (psia) (°C) Excess UF6 Harshaw Chemical Co. 12 to 40 mesh 15* 100 Not removed 0.86 15* 100 By vacuum 0.86 7.5% 100 By vacuum 0.80 15+ 70 By vacuum 0.88 15* 150 By vacuum 0.90 T5* 100 By fluorine 0.86 15%* 100 By fluorine 0.86 8 to 12 mesh 15%* 100 By fluorine 0.81 %-in. pellets 15%+ 100 By fluorine 0.72 15* 100 By vacuum 0.81 Baker & Adamson Co. Powder 15* 100 By vacuum 1.89 *UF, introduced to NaF under vacuum, **UF6 introduced at atmospheric pressure to displace fluorine. TABLE 10.2. SUMMARY OF FOUR CONSECUTIVE RUNS IN TWO-BED FUSED-SALT FLUORIDE-YOLATILITY PROCESS Conditions: Total of 128 g of uranium in Nch-ZrFA-UF4 (52-44-4 mole %) with gross beta activity per milligram of uranium of 5 X 10° counts/min. Each run fluorinated with 1:1 F:;_)N2 mixture for 1.5 hr and then with pure l'_'2 for 0.5 hr; UF6 in F2N2 gas stream absorbed on NaF, with some volatile activity passing into cold trap; UF6 desorbed at 100 to 400°C into a second cold trap Average F2/U mole ratio in absorption pericd: 4/1 Average F2/U mole ratio in desorption period: 1/1 Absorbent beds: 60 ml of 12- to 40-mesh NaF in 1-in.-dia tubes NaF/U weight ratio after four runs: 1/1 Uranium Loss (%) Product Gamma Product Yield Waste Activity per Milligram Run First Cold First NaF Second NaF (%) Trap Bed Bed Salt of Uranium* (counts /min) 1 83 0.0 0.02 3.6 2 35.2 0.10 0.05 3.1 3 151 0.07 0.08 1.0 4 43.8 0.08 0.02 2.1 Over-all 70.1 0.06 0.5 5.1 0.04 2.5 *Gamma activity per milligram of natural uranium is 8 counts/min. 198 PERIOD ENDING DECEMBER 10, 1955 TABLE 10.3. DISTRIBUTION OF VOLATILIZED ACTIVITY Amount* (% of total) Activity Fission-Product First Second Cold Trap NaF Bed NaF Bed Gross beta 81 18 0.85 Gross gamma 11 89 0.14 Ru gamma 97 1.6 1.1 Zr-Nb gamma 2.2 98 0.04 Total rare-earth beta 4,3 92 3.6 *Collected and actually found by analyses of cold trap and NaF beds. the fused-salt fluorination step was about 0.04%, but losses in the two NaF beds were 0.5 and 5%, respectively, These losses were due to some back- pressure buildup, which was evident in all runs, as the result of either partial plugging of the NaF during the absorption period (probably because of a volume change) or plugging of the UF cold-trap inlet. Occasional tapping of the absorption bed prevented the NaF-plugging problem from being very serious in all runs except the second. Here the plugging was so severe that excessive nitrogen pressure had to be used, with considerable loss of product. Plugging of NaF upon absorption of UFé was completely avoided in two subsequent runs, without activity, by using a sieve plate at the enfrance to the first bed for dispersion of the gas instead of just a 1/4-in. inlet line as in the above four runs. 199 Part lli SHIELDING RESEARCH 11. SHIELDING ANALYSIS F. H. Murray C. D. Zerby Applied Nuclear Physics Division S. Auslender H. S. Moran Pratt & Whitney Aircraft AIR SCATTERING OF Co%® GAMMA RAYS: THEORY vs EXPERIMENT H. S. Moran The air-scattered gamma-ray dose rate predicted by theory' for a Co%® source at a source-detector distance of 15 meters was compared with experi- mental measurements in a similar geometry.? After an appropriate correction® for ground scattering was applied, the two results were found to be in sub- stantial agreement. The details of the comparison were published in a separate report.? ENERGY ABSORPTION RESULTING FROM GAMMA RADIATION INCIDENT ON A MULTIREGION SHIELD WITH SLAB GEOMETRY C. D. Zerby S. Auslender fiémzoding of a Monte Carlo calculation of heat generation resulting from transport gamma radiation through shields with stratified slab geometry has been completed.’® Preliminary results are in good agreement with experimental results obtained for lead.’ The code has been extended to produce data on the gamma-ray dose and energy flux as well as energy deposition. Parameter studies on a water- 'E. P. Blizard and H. Goldstein (eds.), Report of the 1953 Summer Shielding Session, ORNL-1575 (June 14, 1954), p 170-203. 28, L. Jones, J. W, Harris, and W, P. Kunkel, Air and Ground Scattering of Cobalt 60 Gamma Radiation, CVAC-1707T (Marcfi 30, 1955). 3M. L. Coffman and B. T. Kimura, Gamma Ray Ground Scattering for Co60 and GTR Sources, NARF-55-16T (May 30, 1955). 14, s. Moran, Air Scattering of €080 Gamma Rays: Theory vs Experiment, ORNL-2019 (Jan. 6, 1956). S5c. D. Zerby and S. Auslender, ANP Quar. Prog. Rep. Mar, 10, 1955, ORNL-1864, p 173. 6¢. D. Zerby and S. Auslender, ANP Quar. Prog. Rep, June 10, 1955, ORNL-1896, p 192, 7€, S, Kirn et al., Oblique Attenuation of Gamma-Rays from Cobalt-60 and Cesium-137 in Polyetbylene, Con- crete, and Lead, NBS5-2125 (Dec. 23, 1952). Data from these studies will be presented as dose and energy buildup factors for monodirectional, iso- tropic, and cosine monoenergetic sources. Also, curves on heat deposition will be obtained for all slabs in the parameter study. lead slab and an iron slab will be made. INTEGRAL EQUATIONS FOR THE FLUX DENSITY NEAR A THIN FOIL AND FOR NEUTRON SCATTERING IN AIR IN THE PRESENCE OF THE GROUND F. H. Murray In calculations for a thin foil it appears that very often it is possible to carry out calculations of the flux in the foil interior as if the flux due to sources outside a part of the foil were constant for the small part of the foil considered. With this assumption, known formulas may be employed fora slab so that upper and lower surface source densi- ties for any part of the foil can be determined in terms of given sources and other parts of the foil. The sum of the upper and lower surface densities becomes the effective surface source density for each small part of the foil, and a set of integral equations is obtained for the coefficients of a harmonic expansion, giving the outward flux from any part of the surface. The analysis needed for neutron scattering in air with ground present is very similar to that employed for a foil. In this case there is no lower surface source density to be considered, and the flux in the ground near any surface point is calcu- lated as if the surface flux were constant over the infinite plane and equal to its value at the surface point. This assumption leads to a set of integral equations for the harmonic coefficients of the sur- face source density, as before. have been published separately.® These equations 8. H. Murray, Integral Equations for the Flux Density Near a Thin Foil and for Scattering in Air in the Presence of the Ground, ORNL CF-55-9-108 (Sept. 23, 1955). 203 ANP PROJECT PROGRESS REPORT 12, SHIELD DESIGN J. B. Dee C. A. Goetz H. C. Woodsum Pratt & Whitney Aircraft R. M. Davis The Martin Company D. R, Otis Consolidated Yultee Aircraft Corp., San Diego The neutron-induced activation of sedium in the heat exchangers of circulating-fuel reactors has been calculated. In this calculation, both core neutrons and delayed neutrons were considered. CALCULATION OF THE SODIUM ACTIVATION IN THE HEAT EXCHANGERS OF CIRCULATING-FUEL REACTORS J. B. Dee D. R. Otis The activation of sodium caused by core and delayed neutrons in the heat exchangers of circu~ lating-fuel reactors was calculated for a range of reactor dimensions and a power of 300 Mw. Calcu- lations for a 60-Mw reactor that corresponds to the ART were also made, Calculation of Activation by Core Neutrons The activation from core neutrons was deter- mined directly from LTSF experimental data ob- tained during static source tests with mockups of a circulating-fuel reflector-moderated reactor (CFRMR)'=3 and various reactor-shield configura- tions. The resuvits for each configuration were averaged over the thickness of the heat exchanger to obtain the average specific activation. The specific activation for each configuration was then plotted as a function of the thickness of various reactor regions (Figs. 12.1 through 12.5). The data were corrected for minor differences between the experiment and the reactor, such as cladding thicknesses and air gaps. The wvariation of the sodium activation with respect to the beryllium reflector thickness (Fig. 12.1) appears to be independent of the boron cur- tains, and it has an apparent relaxation length of . T. Chapman et al.,, ANP Quar. Prog. Rep. June 10, 1955, ORNL-1896, p 194. ZG, T. Chapman et al., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL.-1947, p 197. 3R. VW. Peelle et al., Sec. 13, this report. 204 5.7 em. The small variation with respect to heat exchanger thickness (Fig. 12.2) implies that self- shielding is offset by neutron moderation in the heat exchanger, so that a sodium layer between the reflector and heat exchanger would not be effective in reducing the activation. Increasing the thick- ness of the first boron curtain beyond 2 in. (0.66 g of B'® per square centimeter) indicates that further reduction of the sodium activation (Fig. 12.3) may be obtained by increasing the thickness 4 in. OF HEAT EXCHANGER, Y in. OF BORAL 5 [ CONFIGURATION: Be, 2in OF BORAL, e . 2—01—059—27 10 - T 5[ & CONFIGURATION: Be, Ypin. OF BORAL, N 4in. OF HEAT EXCHANGER, 2 in. OF BORAL —— | i © CONFIGURATION: Be, 2in. OF BORAL, - 2 Bin OF HEAT EXCHANGER, 2 in. OF BORAL o5 L & CONFIGURATION: Be, 2 in. OF BORAL, 4in. OF HEAT EXCHANGER, 2 in. OF BORAL . | 4'— SPECIFIC ACTIVATION PER g OF Na (activations/min} N T W A=57cm ¥ 40? - N = — | —— - — < - \\ i 5 ~ 2 N 'NOTE: DATA CORRECTED FOR BORAL THICKNESS o 2 T AND COMPOSITION B - LS s 0% 8 10 12 14 16 BERYLLIUM THICKNESS (in ) Fig. 12.1. Mean Sodium Activation from Core Neutrons in the Heat Exchangers of LTSF CFRMR Mockups — Effect of Beryllium Thickness. @’""“' 105 2—01—-059—-28 { : | ] [ = ) | J I = © CONFIGURATION: 8in. OF Be, 2in. OF BORAL, __ | E HEAT EXCHANGER, 2 in. OF BORAL w § 5 % A CONFIGURATION: 12 in. OF Be, ‘% in. OF BORAL, | B HEAT EXCHANGER, 2 in. OF BORAL g | ® CONFIGURATION: 12 in. OF Be, 2 in. OF BORAL, 5 HEAT EXCHANGER, 2 in. OF BORAL =z | T : | O U" . e 2 ! ] wl ! o. z \ Q : = I 2 T i ; ! - A . L E ‘\":"--—— [ G . I ul t o 0 5 — km " . . | 2] «® - 3 o < — “ | 1 N 27 S \ S NOTE: DATA CORRECTED FOR BORAL THICKNESS — . AND COMPOSITION | O I . | I ‘ | : 5 I 0 { 2 3 4 5 6 7 HEAT EXCHANGER THICKNESS (in.) Fig. 12.2. Mean Sodium Activation from Core Neutrons in the Heat Exchangers of LTSF CFRMR Mockups — Effect of Heat Exchanger Thickness. of the enriched boron curtain in present designs. The variation with respect to the thickness of the second boron curtain (Fig. 12.4) is small. The re- placement of natural boron with B'? has still to be investigated. This would reduce the moderation (largely by B for equivalent absorption. The decrease in activation occurring with the introduction of a hydrogenous layer (polyethylene) between the beryllium reflector and the first boron curtain is shown in Fig. 12.5 in terms of its beryl- lium replacement effectiveness. Since polyethyl- ene cannot be used in the reactor, the computed effectiveness of zirconium hydride, which has high thermal stability at reflector operating tempera- tures, is shown for comparison. The contribution of prompt core neutrons to the sodium activation in the heat exchanger of a CFR was calculated from the L.TSF experimental data by application of conventional shielding transfor- PERIOD ENDING DECEMBER 10, 1955 2—-01-059-29 1C T l I | a jo CONFIGURATION: 8in. OF Be, BORAL, *t‘ 4in, OF HEAT EXCHANGER, % in. OF BORA & CONFIGURATION: 8in. OF Be, BORAL, 4in. OF HEAT EXCHANGER, 2 in. OF BORAL ‘\\ L] NOTE: DATA CORRECTED FOR BORAL THICKNESS AND COMPOSITION ~ =~~~ ! ! . f 8°8 SPECIFIC ACTIVATION PER g OF Na(activations/min) 5 2] O = | = ~ : ; ~ I { N 2 [~ A CONFIGURATION: 12 in. OF Be, BORAL, Co E 6in. OF HEAT EXCHANGER, 2 in. OF BORAL | - ® CONFIGURATION: 12 in. OF Be, BORAL, . 4in. OF HEAT EXCHANGER, 2 in. OF BORAL ~ ' . w3 ! | 1 | i 0 0.5 1.0 1.5 2.0 2.5 3.0 35 BORAL THICKNESS (in.) Fig. 12.3. Mean Sodium Activation from Core Neutrons in the Heat Exchangers of LTSF CFRMR Mockups — Effect of First Boral Curtain Thickness. mations.? This implies that the neutron captures in sodium in the heat exchanger may be expressed in terms of a kernel involving approximately ex- ponential neutron attenuation through the material thicknesses along the roy connecting a small fis- sioning region with a sodium volume element. In terms of these transformations the prompt-neutron contribution was expressed, as follows: (1) o o z k ZBrB dp.c =—H(a,z)g ayr.c i e , where @p.c = core-neutron-induced activation per gram of sodium in a CFR heat ex- changer (atoms/min}, 4E. P. Blizard, Introduction to Shield Design, ORNL CF-51-10-70, Part | rev. (Jan. 30, 1952) and Part |} rev. {March 7, 1952). 205 ANP PROJECT PROGRESS REPORT 5 2-01-059-30 10 : E E T — — —JP - - - — [ — g A CONFIGURATION: 8 in. OF Be, 2 in. OF BORAL, 5 | 4 in, OF HEAT EXGHANGER, BORAL = S 1 —t—1 — 4t 8 ¥ CONFIGURATION: 12 in, OF Be, 2 in. OF BORAL, 5 P 4 in. OF HEAT EXCHANGER, BORAL z | : | l | L ® CONFIGURATION: 2 in. OF Be, % in. OF BORAL, o> ’ 6 in. OF HEAT EXCHANGER, BORAL § ! ‘ 5 2 ) | _ ; ‘ : _ | g | = _-_‘—-—.}-_____ 2 T q | = i 5 10t R | = correction allowing for the sodium z - Az displacement outward in the experi- ment because of the bulk of alumi- num in the boral and the air gaps which are not present in the CFR (plotted in Fig. 12.6),% Az = thickness of boral and air gaps {cm), e = allowance for the difference between the neutron attenuation through boral in the LTSF mockup and that through B,C-Cu in the CFR, EB = activation-reducing cross section for the aluminum in the boral = 0.0164 ¢cm~!, based on multigroup calculations at Pratt & Whitney, tg = thickness of boral (cm). Calculation of Activation by Delayed Neutrons The activation induced by delayed neutrons emitted in the heat exchanger was estimated from the experimental data taken at the LTSF with the use of a moving fission belt (see Sec. 13). Since the delayed-neutron emission from any point on the belt is proportional to the time of exposure to the circular opening (core hole) of the ORNL Graphite Reactor, and since the emission is con- stant at any point during the belt rotation (be- cause the belt transit time is less than the aver- age half life of the delayed-neutron emitters), the distribution of delayed-neutron emission for the belt can be expressed, as follows: (2) nLT()’) = ”LT(O) VAR (y/d)2 K, where nLT(y) = number of delayed neutrons emitted per second from a square centimeter of the belt at a distance y (ecm) from the center line (neutrons/cm?.sec), ”LT(O) = emission rate on the belt center line (neutrons/cm?-sec), PERIOD ENDING DECEMBER 10, 1955 UNCLASSIFIED 2-01-052-38 30 : ‘ | | \ ! 2.0 o4 0 1 20 30 40 5C 60 70 80 90 100 z— Az Fig. 12.6. Plot of £ vs (z — Az). a = half width of the emitting area of the belt (equal to the radius of the core hole of the ORNL Graphite Reactor), K = ratio of emission from belt completing circuit in finite time to emission in zero transit time. decay of delayed-neutron emitters that complete one-half the circuit is K ~ 0.96.) (The correction for Normalizing the emission rate over the total belt area to the total power from fissions in the belt gives 2P, ..d 3 ny p(0) = —=L ®) LT Tal where P, + = total power from fissions in the belt (5.2 w obtgined by normalizing to the LTSF static tests), d = number of delayed neutrons emitted per watt of fissions (neutrons/w), ! = length of belt (cm). 5Under the assumption of exponential ray attenuation, the following expression was used to determine &: In E,q [(z - Az)/?\n} - E-‘{(z - Az)/)tn V1t [az/(z - Az)z] } £,z =82/ ] - By {(z = Dap/\ V1 + (Y22} In [z/(z - Az)] 207 ANP PROJECT PROGRESS REPORT An approximate value for the age of delayed neu- trons moderated from 500 to 3 kev in the heat ex- changer was calculated, by means of the following expression, to aid in interpreting the experimental results in terms of the CFR geometry. (4) T =f50 ) , E 1 fET E which was approximated by In (Eo/E ) 3 2 No, ZS‘ ENO; 7 (4a) T = = 216 cm? F = neutron energy (kev), D = diffusion coefficient (1/3X N7 ), 1 2, = fotal neutron macroscopic cross section for the heat exchanger {assumed to be homogeneous), Nz. = number of atoms per cubic centimeter of the ith kind, &, = average logarithmic energy decrement for atoms of the ith kind, O . = total neutron microscopic cross section for atoms of the ith kind averaged over the lethargy range. The average composition {in g/cm?) of the LTSF heat exchanger mockup was: nickel, 2.23; sodium, 0.65; fluorine, 0.54, Since /7 (14.7 cm) is small compared with the belt length in the heat exchanger (120 cm) but significant compared with the belt half width (35.5 cm), the ratio was obtained of the slowing- down density from a source of infinite length but finite width (with sources distributed according to Eq. 2) to the slowing-down density from an infinite plane source representing the uniform re- The slowing-down density from an infinite plane was taken to be e_,02/11'7' — = 0.0192n 4T actor heat exchanger. (5) qmp[ = no:)p[ opl ! where ep) = delayed neutron emission per square cen- timeter from an infinite plane (neu- trons/cm?.sec), p = distance from infinite plane to sodium sample (2.54 cm). 208 The slowing-down density from the LTSF fis- sion belt was taken to be “ ~V(p? —y%)/ar e (6) = 2 n (y) dy LT LT 4T 1 0 since q; = slowing-down density from an infinite line source 2 = nle—r /47/4777', n; = line source strength, taken to be n; .(y) dy (neutrons/cm?.sec), r = distance from the line to the sodium sample 2 (czm)' 2 r = p — y - The integral was evaluated numerically to ob- tain the following resuit: (7) qor = 0.01937, .. , where P d — LT (8) LT = 2al For equivalent source strengths the ratio of slow- ing-down density (‘prl/qLT) at 3 kev for the LTSF belt compared with an infinite plane is thus seeo to be approximately unity. The fuel in the heat exchanger of a CFR is approximated by an infinite plane with a delayed neutron source strength of PRd (9) Popl T, Taxtux where ! P, = reactor power (watts), V, = total fuel volume of system (em3), fyx = fraction of heat exchanger volume con- taining fuel, tx = heat exchanger thickness (cm). If it is assumed that the sodium activation is proportional to the slowing-down density of 3-kev neutrons, then the specific activation in the heat exchanger of a CFR is given by 0.019210,; iy 0.019372'LT Ty 4R-D Topl (10) L AL T-D LT It sodium in {atoms /min), delayed neutron activation per gram of a CFR heat exchanger delayed neutron activation per gram of sodiumin a LTSF mockup (atoms/min), (For the calculations presented in FPERIOD ENDING DECEMBER 10, 1955 Calculated Total Activations for Several Reactors The total activation of sodium in the heat ex- changers from core and delayed neutrons was caleculated for four circulating-fuel reactors (Table 12.1). The power for three of the reactors was 300 Mw, and the fourth reactor corresponded to the ART, with a power of 60 Mw. In addition to the total saturated activities, the total activity after Table 12.1,a; +.p = 1500 atoms/min.) 30 hr of operation was determined. TABLE 12.1. SODIUM ACTIVATION IN THE HEAT EXCHANGERS OF CIRCULATING-FUEL REACTORS Configuration ] 2 3 4* Island radius (in.) 4 4 4 Core radius (in.) 8.59 10.16 9.74 10.5 Beryllium thickness (in.) 8 12 16 1 First B'0 curtain (atoms/cm?) 3.52 x 1022 3,52 x 1022 3,52 x 1022 *x Heat exchanger thickness (in.) 8 6.5 5 2.1 Second B0 curtain (atoms/cm?) 1.12 x 1022 .12 x 1042 1,12 x 1022 * PEWai density (kw/cm>) 5.5 2.75 2.75 0.72 Weight of sodium in heat exchanger (g) 118,000 144,000 137,000 42,000 Reactor power (Mw) 300 300 300 60 Saturated activity from core neutrons (curies) 19,300 3,350 430 470 Saturated activity from delayed neutrons (curies) 1,970 1,530 1,250 61 Total saturated activity (curies) 21,270 4,880 1,680 531 Total activity after 30 hr (curies) 16,060 3,680 1,270 400 *Configuration 4 comresponds to the ART configuration, **First and second B]0 curtains consist of 0,375 in. of natural B4C. 209 ANP PROJECT PROGRESS REPORT 13. LID TANK SHIELDING FACILITY R. W. Peelle J. M. Miller Applied Nuclear Physics Division W. J. McCool J. Smolen Pratt & Whitney Aircraft Analysis of the static source tests in the second series of experiments with mockups of a circulating- fuel reflector-moderated reactor and shield (RMR- shield) in the Lid Tank Shielding Facility (LTSF) was completed. A summary of the tests is pre- sented. The dynamic source tests were completed, and the tests are being analyzed. Two tests with the dynamic source are described here. STATIC SOURCE TESTS WITH MOCKUPS CF A REFLECTOR-MODERATED REACTOR AND SHIELD H. C. Woodsum! The static source tests in the recent series of experiments with RMR-shield mockups were sub- divided into sections according to the parameters being investigated. These parameters, descrip- tions of the various mockups tested, and an index J. Smolen to the results of the measurements are given in Table 13.1. Much of the information has been reported previously,?:3 as indicated in the table, Most measurements not previously published are presented in this report, Effect of Varying Lead Thickness The thickness of lead was varied in mockups having 8-, 12-, and 16~in.-thick beryllium reflectors. Gamma-ray dose-rate measurements for the 8- and 16-in,-thick reflectors are reported here (Figs. 13.1 and 13,2), but the data are not directly comparable because the ‘‘core shell”’ materials differed. With the 8-in. reflector, Li-Mg (¥ in, thick) was sub- stituted for the Inconel (% in. thick) used to mock up the core shell. Inconel gives a hard capture gamma ray, approximately 9 Mev. Thus, in order to correlate the gamma-ray data from these two beryllium thicknesses with data from 12 in. of beryllium,2 two configurations were measured to 1Shield Design Group, Pratt & Whitney Aircraft at ORNL. 2, T. Chapman, J. B, Dee, ond H. C. Woodsum, ANP Quar. Prog. Rep. June 10, 1955, ORNL-1896, p 194. 3G. 1. Chapman et al., ANP Quar. Prog. Rep. Sept, 10, 1955, ORNL-1947, p 197. 210 determine the effect on the gamma-ray dose rate of interchanging Inconel and Li-Mg (Fia. 13.3). The effects on the thermal-neutron flux of varying the thickness of lead in mockups with 8 and 16 in. of beryllium are shown in Figs. 13.4 and 13.5, respectively. Measurements beyond mockups with 12 in, of beryllium were previously published.? Fast-neutron dose-rate measurements beyond mockups with various lead thicknesses and 16 in. of beryllium (Fig. 13.6) appear to be self-consistent and to be consistent with data for 12 in. of beryl- lium.2 The measurements for 8 in. of beryllium (Fig. 13.7), however, do not appear to be self- consistent, This was apparently due to inter- mittent instrument malfunctioning. The results Sunane 3 2-01—-057—-66—-169 GAMMA-RAY DOSE RATE {mr/hr) o CONFIGURATION 2-A, NG LEAD " a CONFIGURATION 2-B, 1'% in. OF LEAD ® CONFIGURATION 2-C, 3n.OF LEAD A CONFIGURATION 2-D, 4% in. OF LEAD 60 7O 80 90 100 HC {20 130 140 150 160 z, DISTANCE FROM SOQURCE (cm) Fig. 13.1. Gammo-Ray Dose Rate Beyond RMR- Shield Mockups — Effect of Varying Thickness of Lead Behind 8 in. of Beryllium, TABLE 13.1. SUMMARY OF REFLECTOR-MODERATED REACTOR AND SHIELD MOCKUP TESTS AT THE LTSF Note: The water region behind the lead was borated to 1,95 wt % unless otherwise indicated PERIOD ENDING DECEMBER 10, 1955 Configu- Components and Thicknesses Index to Plots of Data (&) ) Nickel Lead a Investigation ration S(.I:orlel se:r”ium Boral HNQF';' :anks Boral (Pressure (Gamma T‘:TGI Other Deviar ;:2,(.")) Gamma-Ray Thermal- Fast o No. ‘e {(Re ‘ector) in.) (Heat 'xc anger) (in.) Shell) Shield) ir ther Deviations Dose Rate Neutron Neutron Activations (in.) (in.) (in.) (in.) (in.) (em) Flux Dose Rate Effect of varying water bora- 1-A Plain water in Lid Tank (i. e., no RMR tank instalied) o 0 ORNL-1896 ORNL-1896 tion 1-B RMR tank installed and filled with plain water 4,55(€) 5.50(4) " " ORNL-1896 1-C RMR tank installed and filled with 0.45%-borated water 4.55(6) 5.50(d) x X X 1.D RMR tank installed and filled with 1.95%-borated water 4.35(°) 5.3 ORNL-1896 ORNL-1896 X Effect of varying lead thick- 2-A Li-Mg, Y% 8 2 4 1 0 6.47 52.2 Fig. 13,1 Fig. 13.4 Fig.13.7 ness behind 8 in, of beryllium 2-B i " " " n " ],1/2 7.36 56.9 " " " 9.C 1 " " I 1" 1" 3 6.55 59.9 ' " " 2.D 1" "t n 1 n " 41/2 6.84 64.0 1" 1 n Effect of varying lead thick- 3-A Inconel, ]/8 12 2 4 2 1 0 4.30 59.6 ORNL-1896 ORNL-1896 ness behind 12 in. of 3-B " " y " ! " 1% 4.89 64.0 " § beryllium 3-C " " " " I " 3 4,78 67.7 " n ORNL-1896 3-D " n rn n 1 1 41/2 9.83 76.1 x x x 3-D° " " " " " " " 6.07 72.8 ORNL-1896 x x Au, ORNL-1896 3-D” " " " " " " " 4,77 71.5 " ORNL-1896 ORNL-1896 3.E " ] 1 Ii " 1 6 5.06 75.6 n 1" 3-F L I d y iy d 7% 555 79.9 t " ORNL-1896 1-G 1 " " 1 " " 9 6.14 84.3 I 1" x Effect of varying lead thickness 4-A Inconel, 1/8 16 2 4 2 1 3 6.45 79.8 Fig. 13.2 Fig. 13.6 Au and Na behind 16 in. of beryllium 4-B " " " ! " " al, 6.0 82.9 ! Fig. 13.5 . 4-C " | " 1 I 1 6 6.93 87.7 I n Inconel core shell and capture 5-A Inconel, ]/8 8 2 4 2 1 41/2 5.03 61.6 Fig. 13.3 x Au,Ng, ORNL-1947 gamma-ray dose 58 LiMg, % 12 " " " ' " 5.15 72.2 I x Effect of Inconel cladding be- 6-A Inconel, }’8 12 2 4 2 1 4]/2 5.72 0.020 in. of Inconel behind beryllium 72.5 ORNL-18%6 tween beryllium and first 6-B " " i " " " " 5.85 0.125 in. of Inconel behind beryliium 72.9 " boral curtain E ffect of boral thickness 7-A Inconel, }é 8 !/é 4 2 1 4],6 4,24 57.0 X X Na behind 8 in. of beryllium 7-B Li-Mg, ) n 1%, " 1Y E n 7.68 62.3 x x x 7.C 1 1" 2 1" 1,2 1 " Na 7.D n 1 3]/2 1 I it 1" 7.44 64.6 Ng Effect of heat exchanger thick- 8-A Li-Mg, 1/‘" 8 1 6 1 1 3 x Na, LiCl ness and boral distribution 8-B " It n " " " n 6.11 I/2--in.-thick boral slab behind first and 62.0 x Na, ORNL-1947, behind 8 in, of beryllium second NaF-Ni tanks (total of three tanks) MgCI2 8-C Inconel, '/8 " 2 " 2 " 41/2 Ma, ORNL-1947 211 ANP PROJECT PROGRESS REPORT TABLE 13.1 (continued) Components and Thicknesses Index to Plots of Dafu(b) Configy- Core Beryllium NaF-Ni Tank Nickel Lead Total 2, Investigation ration Shell (Ref:ecfor) Boral (Hec:lf E |chc: sr) Boral (Pressure {Gamma :i: Other Deviations (2'“) Gamma-Ray Thermal- Fost . Na. (in) (i (in.) (ixn )u ge (in.) Shell) Shield) (cm) ' Dose Rate Neutron Neutron Activations : . . (in.) (in.) Flux Dose Rate Effect of boral thickness be- 9.A Inconel, !/B 12 0 4 ) ] 4}'2 6.55 68.2 ORNL-1896 x hind 12 in. of beryllium 9-B " " ‘,5 n n " I 5.98 68.9 " x Na, ORNL-1896 9-C " " ] " " " " 5.81 70.0 x x Na, ORNL-1896 9-D " " 1% " 1Y " " 8.57 72.3 x x x 9-E Li-Mg, Y " 2 " % " " 5.16 68.4 x Na Effect of heat exchanger thick- 10-A Inconel, }8 12 ]/2 2 2 1 4]/2 4.56 62.4 x x x Na ness and boral distribution 10-B " " 2 6 " " " 6.09 77.9 x x Na, ORNL-1896 behind 12 in. of beryllium 10-C ' i Y " " " " 5.50 73.5 ORNL-1896 x x " 10-D " n n n ]é n 1 5.71 69.9 X X n Effect on secondary gamma-ay 11-A Li-Mg, ]& B ]]/2 4 11/2 1 4]/2 7.74 ]/z-in.-thick boral slab on each side of last 64.9 Fig. 13.8 X dose of distributing boral in 11/2-in.-thick lead slab lead Effect on secondary gamma-ray 12-A Li-Mg, }:‘ 8 2 4 2 1 4!/2 7.05 Lead-water spacing as shown in Fig., 13.9 60.4 Fig. 13.9 dose of lead spacing 12-B " " " " " " " 8.65 I 62.0 3 12-C " " I I 1 " " 8.75 I 62.1 1" 12-D " 1 E t 1 1" 1" 8.55 " 61.9 n 12-E n u 1" 1" i " 6 7.94 n 65.1 t 12-F n 1 n n 1" " " 7.05 1 60.4 " Capture gamma-ray dose from 13-A Inconel, }é 12 2 4 2 ] 9 6.40 z-in.-fhick Li-Mg slab behind lead 85.2 x shield and effect of gamma- 13-B " " " " ” i n 6.00 k‘-in.-thick boral slab behind lead 84.8 x(€) x{€} Na, GORNL-1896 ray streaming 13-C n " " " " " I H Same as configuration 13-B with bismuth on 84.8 x(€) x sides of RMR tank 13-D n " 1" 1 " n " " Same as configuration 13-C with !fi-in.-fhick 84.8 x boral slab just inside water bag 13-E n " " " " " I 6,14 Bismuth and lead shields on sides of 84.3 x RMR Tank Effect of varying pressure shell 14-A Li-Mg, !:1 8 2 4 2 0 41/2 7.48 62.1 x thickness 14-B Inconel, ‘/8 12 " " " 2 i 4.93 74.2 ORNL-1896 x Effect of heavy metals in re- 15-A Li-Mg, ]fi 8 2 4 2 1 41/2 7.42 3-in.-thick bismuth slab behind beryllium 72.2 Fig. 13.13 Fig. 13.14 flector 15-B " " i " ) " " 7.56 2-in.-thick copper slab behind beryllium 69.8 " " Na, ORNL-1947 Effect of replacing borated 16-A inconel, I/é 12 1]/2 4 'Ilé 1 4]/2 9.27 Borated water replaced by plain water 73.0 Fig. 13.15 Fig. 13.16 water with plain water behind configuration (@)Thickness of configuration measured from swface of source plate to first water layer. (8)y indicates that z traverse measurements have been taken but not published to date. Configuration numbers cited in ORNL-1896 should be disregarded because the numbering system has been changed. 212 () Air bag inserted between source plate and aluminum window of RMR tank. ({)Represents total thickness of air and tank wall of RMR tank. (€)|ln configuration 13-B, x and y traverses were also made; in configuration 13-C, y traverses were also made. — 2—01-057—66-161 **Jw PERIOD ENDING DECEMBER 10, 1955 Wrs. kW 2—01—-057—-66—-176 e S S — j:i;’ £I—]:;f T e - | TS ] 5 _ o CONFIGURATION 2-A, NO LEAD — +— & CONFIGURATION 2-B, ii/z'!n‘ OF LEAD — - GAMMA —RAY DOSE RATE (mr/hr}) o © CONFIGURATION 4-A, 3in. OF LEAD A CONFIGURATION 4-8, 4l5in. OF LEAD O CONFIGURATION 4-C, & in. OF LEAD -2 10 80 20 100 Ho 120 {30 140 150 160 2, DISTANCE FROM SOURCE (c¢m) Fig. 13.2. Gomma-Ray Dose Rate Beyond RMR- Shield Mockups =« Effect of Varying Thickness of Lead Behind 16 in. of Beryllium. S 2-01-057—66-175 2 o 10" -0 CONFIGURATION 5-A, INCONEL, 8 in. Be o CONFIGURATION 2-D, Li-Mg, 8in. Be — GAMMA-RAY DOSE RATE {(mr/hr) o CONFIGURATION 3-D INCONEL, 12 in. Be & CONFIGURATION 5-8, Li-Mg, 12in.Be 70 80 90 100 HO 120 130 140 150 160 z, DISTANCE FROM SOURCE {cm) Fig. 13.3. Gamma-Ray Dose Rate Beyond RMR- Shield Mockups — Effect of Inconel Core Shell. 0 CONFIGURATION 2—C, 3in.OF LEAD —— — \\ ® CONFIGURATION 2-D, 4% in OF LEAD 1ot D\ : ! . | | Lo C— )T - ; - — t THERMAL - NEUTRON FLUX (nvm) 50 60 70 80 30 100 1o 120 130 2, DISTANCE FROM SOURCE (cm} Fig. 13.4. Thermal-Neutron Flux Beyond RMR- Shield Mockups = Effect of Varying Thickness of Lead Behind 8 in. of Beryllium, for 0 and 3 in. of lead are thought to be in error by a factor of 2, but the remaining data are probably more accurate, |t is hoped that this set of meas- urements can be made again. In the meantime, the data obtained to date are presented in lieu of more exact measurements. Study of Secondary Gamma-Ray Production In order to analyze the gamma-ray dose rate from secondary gamma rays produced in the shield, 213 ANP PROJECT PROGRESS REPORT L 2-01-057-66-162 S n n CONFIGURATION 4-8 THERMAL-NEUTRON FLUX (nvfh) 10 5 2 1072 70 80 90 100 1o 120 130 7, DISTANCE FROM SCURCE {(cm) Fig. 13.5. Thermal-Neutron Flux Beyond RMR- Shield Mockups — Effect of 41/2 in. of Lead Behind 16 in. of Beryllium, two sets of tests were run, The first was a study of the effect of distributing boral in the lead region, and the second was a study of the effect of spacing a portion of the lead out into the borated water. In a configuration that had a 4'/2-in.-thick lead region, a Y-in.-thick boral slab was inserted on each side of the last l'/z-in.-thick lead slab. No large decrease in the gamma-ray dose rate resulted (Fig. 13.8). Thus it appears that the secondary 214 Sy 2-01-057-66-163 FAST-NEUTRON DOSE RATE (mrep/hrl ® CONFIGURATION 4-A, 3-in. OF LEAD A CONFIGURATION 4-B, 4% -in.OF LEAD O CONFIGURATION 4-C, 6-in.OF LEAD 8O 20 100 110 120 Zz, DISTANCE FROM SQURCE (cm} Fig. 13.6. Fast-Neutron Dose Rate Beyond RMR- Shield Mockups « Effect of Varying Thickness of Lead Behind 16 in. of Beryllium. gamma-ray dose rate is not mainly due to thermal- neutron captures in the lead. This is confirmed by calculations of the lead capture gamma-ray dose rate based on the thermal-neutron flux measure- ments immediately behind the lead and on the thermal-neutron capture cross section in the lead. Spacing the last 1Y%-in.-thick lead slab out into the borated water showed that the total gamma-ray dose rate decreased as the water thickness pre- ceding the slab was increased from 13 to 7 in. (Fig. 13.9). However, as the water thickness was increased beyond 7 in., there was no detectable change in the total gamma-ray dose rate. This is Gonrtr i 2 2-04~-057-66-172 10 5 2 10 5 < i . o 2 € w2 W \ = > | E 10° A R , L | < — - : . o ; . o .\LT_ ] L | i § TEST. TRANSIT TIME {sec} CONFIGURATION ) a 1 @ L o S = ] 50 Etf S 2.00 R | o 3 150 B B R | = 4 1.25 o I = I 2020 @ 5 ® B X i ® ' -in- THICK BORAL SHEET ' 2 - [ 1 2-in-THICK NaF TANK o ‘ 1-in- THICK PLASTIC SHEET ‘ B FUEL PLATES | | S | 35 40 45 50 55 60 65 z, DISTANCE FROM FISSION SOURCE (cm) Fig. 13.19. Sodium Activation in the Heat Ex- changer Region of the RMR-Shield Mockups for Yarious Fuel Transit Times, originating in the core, so that if the stationary uranium piates between the heat exchanger tanks affected sodium resonance neutrons at all, the effects would tend to be more pronounced in the second heat exchanger tank, Therefore the com- parison between the two test implies that the uranium in the heat exchanger had little, if any, effect in the static tests. Fission-Product Gamma-Ray Spectrum R. W. Peelle W, Zobel3 T. A. Love’ For a circulating-fuel reactor (CFR) a large fraction of the gamma-ray flux outside the reactor 3Bulk Shielding Facility. PERIOD ENDING DECEMBER 10, 1955 shield originates from fission-product decays® in The first ex- perimental energy spectrum has now been measured the primary heat exchanger region. for the gamma rays from the gross fission-product mixture, For these measurements a multiple- crystal 78 was used in conjunction with the circulating solid fuel belt (described above). Although the spectrometer was installed in the LTSF, the associated electronic instrumentation remained at the Bulk Shielding Facility. Neutrons from the ORNL Graphite Reactor were allowed to impinge upon the solid fuel belt shown in Fig. 13,17, Figure 13.20 is a sketch of the configuration used in the LTSF in combination with the spectrometer and fuel loop. The boral, lead, and water placed within the perimeter of the loop were provided to reduce as much as possible the counting rate in the spectrometer caused by sources other than the section of the fuel belt viewed by the spectrometer collimator, A photograph of the experimental assembly with the spectrometer region partially dismantled is shown in Fig. 13.21. The fuel loop (1) is at the left of the picture, hidden by the remainder of the configuration. The water tank (2) and the lead shielding (3) within the loop, as well as the major portions of the spectrometer assembly, show plainly. The scintillation spec- trometer (4) is surrounded by 6 to 8 in. of lead, except for a ]],é-in.-dic gamma-ray collimator pointing at the fuel loop. Bricks (5) made from a combination of LiF and paraffin were stacked around the lead spectrometer shield to a thickness of 8 to 12 in. During operation the top of the assembly was plugged with lead and LiF-paraffin shields (not shown in the photograph). Lumber was stacked in the regions between the lithiated paraffin and the plywood wall (6) protecting the rotating fuel belt. Measurements of the gamma-ray spectrum were made while the fuel loop was operated in such a manner that any one fuel plate completed a revo- lution every T seconds. The observed counting rate was a function of time after the start of acti- vation of the rotating belt, gamma-ray spectrometer In principle, it would 6R. W. Peelle, T. A. Love, and F. C. Maienschein, ANP Quar. Prog. Rep. June 10, 1955, ORNL-1896, p 203. ’F. C. 'Maie,nschein, Multiple Crystal Gamma-Ray Spectrometer, ORNL-1142 (April 14, 1952). BT, A. Love, R. W. Peelle, and F. C. Maienschein, Electronic Instrumentation for a Multiple-Cry stal Gamma- Ray Scintillation Spectrometer, ORNL-1929 (Oct. 3, 1955). 223 ANP PROJECT PROGRESS REPORT t—-¥17’/2in. Afl Y/2in. OF BORAL Y4—in. BORAL SHUTTER WALL OF TANK 9 5}/3 in. # * o, iy " 2-01-057-66-179 73/, in, OF BORATED WATER Yo in. OF BORAL GAMMA-RAY / COLLIMATOR SPECTROMETER SHIELD 8in. OF LEAD FRAME SUPPORTING ROTATING FUEL BELT Fig. 13.20. Expetimental Arrangement for Measurement of the Fission HEAT EXCHANGERS AND SPECIAL EQUIPMENT E. MACPHERSON J. C. AMOS J. W, COOKE M. H. COOFER L. H. DEVLIN L. R. ENSTICE F. R, J. D. o FERRIGNO D. PEAK G. TURNER R. WARD W. F. BOUDREAU J. PARKER, SEC. GENERAL EQUIPMENT %. R. C580RN F. M. LEWIS D. T. STOREY 1. R. EVANS D. L. GRAY E. MAEYENS T. H. MAYES 5. R.ASHTON W. 0. CHANDLER G. C, JENKINS E. B. PERRIN WELDED EQUIPMENT R.B. CLARKE B. E. BLACK E. B. EDWARDS H. W. HOOVER E. A. JAGGERS T. A, KING 1 W.TEAGUE SPECIAL EQUIPMENT W. D. GOOCH L.w. LOVE T. K. WALTERS D. R, WESTFALL IN-PILE LOOP TESTS B. TRAUGER C.C.BOLTA L. P. CARPENTER J. A, CONLIN C. W. CUNMINGHAM P. A GNADT D. M. HAINES T . BENDER M. L. OVERTON, SEC. CORE STUDIES WHITHAN M. SMITH J. STELZMAN Ero . PEACH . TUNMNELL o O PROCUREMENT CODRDINATION REACTOR CONSTRUCTION ENGINEERING TEST UNIT PWA ARE ARE PWA ARE FWA FWA PWA ARE ARE PWA PwWA PWA PWA PWA PiA PWA ARE ARE ARE ARE ARE TCARE ARE ARE ARE ARE ARE ARE ARE ARE ARE ARE ARE ARE ARE ARE ARE ARE PWA ARE ARE ARE PWA ARE ARE ARE ARE PWA ARE ARE ARE ARE ARE ARE ARE ENGINEERING DESIGN ARE ARE TEST SERVICES W. B. MCDONALD ARE D. P, HARRIS, SEC. ARE A. N. MONTGOMERY ARE TEST FACILITIES ASSEMELY COORDINATION E. $STORTO ARE J. L. CROWLEY ARE W, H. KELLEY ARE W. ). MUENZER ARE H. C. SANDERSON Pwa CORROSION LOOPS C. P. COUGHLEN ARE R. A, DREISBACH Pwa P, G, SMITH ARE ELECTRIC SERVICES E. M. LEES ARE D. L. CLARK ARE C. E. MURPHY ARE TEST ENGINEERING 4.1 MILICH Pwa . 1. W KINGSLEY ARE TECHNICIAN GROUP R.HELTON ARE J. 5. ADDISON ARE T. ARNWINE ARE G. 5. CHILTON ARE J. M. COBURN ARE 1. R. CROLEY ARE J. M. CUNNINGHAM ARE R.E.DIAL ARE J. R. DUCKWORTH ARE W. H. DUCKWORTH ARE W. K. R. FINNELL ARE H. FOUST ARE R. H, FRANKLIN ARE W. D. GHORMLEY ARE C. ). GREEN ARE T.L. GREGORY ARE F. M. GRIZZELL ARE R. A. HAMRICK ARE P. P, HAYDON ARE €. G. HENLEY ARE B, L. JOHNSON ARE J.R. LOVE ARE D. E. MCCARTY ARE G. E. MILLS ARE B. H. MONTGUMERY ARE C. €. NANCE ARE J. 1. PARSONS ARE D. G. PEACH ARE H. E. PENLAND ARE R. RED ARE F. J. SCHAFER ARE J. R. SHUGART ARE C. E. STEVENSON ARE A. G, TOWNS ARE C. A. WALLACE ARE B. T, WILLLAMS ARE OFFICE SERYICES J. P. LANE ARE H. C. GRAY Pwa 1. ZASLER ARE L. E. FERGUSON, SEC, ARE REACTOR DESIGN D. C. BORDEN PWa J. M. CORNWEL L PWA R. C. DANIELS ARE H. J. DICKINSON Pwa J. Y. ESTABROOK ARE W. E. LEUTLOFF Pwa C. MANTELL PWA ¥. ). NELSON PWA 1%, TUMAVICUS PwA PUMP DEWGN W. G. COBB ARE I. T. DUDLEY ARE W._ S, HARRIS PWA R. E. HELMS ARE A, D, JARRATT PWA F. L. MAGLEY ARE W, H. POND Pwa, L. ¥. WILSON ARE W. C. GEORGE ARE ELECTRICAL DESIGN T. L. HUDSON ARE A. H. ANDERSON ARE J. KERR ARE B. C. GARRETT PwA J. 0, HICHOLSON ARE GENERAL DESIGN A.A. ABBATIELLOD ARE E. F. JURGIELEWICZ Pwa J. T. MEADOR Pwa W. A, SYLVESTER Pwa N, E. WHITNEY PWA Wl GALYON ARE G. R HICKS ARE J.R. LARRABEE Pwa G. G- MICHELSON ARE C. A, MILLS ARE C. F. SALES ARE RECORDS AND PRINTING §. J. FOSTER ARE J. 1. PLATZ ARE CONSULTANTS F. A, ANDERSON, UNIVERSITY OF MISSISSIPPI L. F.BAILEY, UNIVERSITY OF TENNESSEE R. L. MAXWELL, UNIVERSITY OF TENNESSEE W. K. STAIR, UNIVERSITY OF TENNESSEE 7503 AREA CONSTRUCTION W. G. PIPER ARE S, M. JANSCH, SEC. ARE R. CORDOVA ARE W. F, FERGUSON ARE V. J. KELLEGHAN ARE F. R. McQUILKIN ARE A M. MILLS, JR. ARE G. C. ROBINSON ARE R. D. STULTING ARE C. F, wEST ARE THE AIRCRAFT NUCLEAR PROPULSION PROJECT AT THE OAK RIDGE NATIONAL LABORATORY DECEMBER 4, 4955 SUPPORTING RESEARCH W. H. JORDAN A, J. MILLER STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT STAFF ASSISTANT W. K. ERGEN ARE E. P. BLIZARD AP . R. GRIMES MC W. D. MANLY M D. 5. BILLINGTON S5 H. F. 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AC 5' fiAESSEDER Pwfi B. J. REFCE M .V KLAUS s D.P. GREGORY PA F. H., MURRAY AP G. GOLDBERG AC D. H. JANSEN M HOT LAB FACILITY ;' : E%Ex"“ EEE CONSULTANT R. B, MURRAY AP PHASE EQUILIBRIUM STUDIES B. L. MCDOWELL AC I E POPE " & NULLER A L. A, NCHEL, GEORGIA TECH. g. ZE%:;BEY :Ip: C. 1. BARTON e W. J. ROSS AC L R. TROTTER " M. é-.’;?_t?SMAN gg b BALNER e - B L. M. BRATCHER MC DEVELOPMENT E. J. MANTHOS 55 J. L. WANTLAND REE SHIELD DESIGN R. 1. SHEIL MC PYNAMIC CORROSION GROUP CERAMIC RESEARCH W. B. PARSLEY 55 R. M. BURNETT REE R. E. THOMA ve 1. P. YOUNG AC J. H. DEVAN M M. WARDE* " R. A RAMSEY 55 J- LONES REE 1. B.DEE PwA R. E. CLEARY PWA J.R. FRENCH AC E. A, KOVACEVICH M - . E. 5. SCHWARTZ s R. L. MILLER REE M. A. MARLER AC A. HOBBS,™ SEC. M G. M. WINN REE CRITICAL EXPERIMENTS R. DAVIS GLM H. A. FRIEDMAN MC G. D. BRADY M C. A, GOETZ PWA B. A. SODERBERG MC E. J. LAWRENCE M RADIATION METALLURGY A, D, CALLIHAN® AP 0.R.OTIS cv R. E. MOORE MC SERVICE M. A. REDDEN M C.E. CURTIS® M J. C. WILSON* 55 PHYSICAL PROPERTIES M, L. RUEFF,* SEC. AP H. €. WOODSUM PwaA H. DAVIS PWA W. F. VAUGH T . « P~ YAUGHN AC FABRICATION GROUP L. M. DONEY M C. D. BAUMANN 5 . 1. COHEN REE LID TANK SHIELDING FACILITY R B MEADOWS e R.F.APPLE AC J. A. GRIFFEN" M ¥. E. BRUNDAGE 5 W. D. POWERS REE R. P. METCALF MC D. E. CARPENTER AC 1. H. CooBS M A.J. TAYLOR* M W. W. DAVIS 58 " 5. €. BLALOCK REE M. K. ALBRIGHT AP R.W.PEELLE AP FUEL PREPARATION RESEARCH R C. BRYANT AC M, R. D'AMORE PWA G. D. WHITE* M N E. HINKLE 55 5. J. CLAIBORNE REE W Faobe oA wiecoor o o . waTsON w L F. oo a e ouh £ ALEXADER: ¥ b5 oLy, 5 7% Jones R. GWIN AP 1. MILLER AP M SLotD e A. H. MATTHEWS AC R. MCDONALD PWA J. C. ZUKAS 58 V. G. HARNESS* AP ). SMOLEN PWA o C. E. PRATHER AC 3. P, PAGE W T. PRICE PWA 1. J. LYNN AP " E. BECKHAM AR E &. zfl}EESY :g A. D. WILSON AC T. K. ROCHE M CONSULTANTS E. R. ROHRER* AP 1. FRANCIS AP - C. M. WILSON AC R W. JOHNSON M RADIATION CHEMISTRY E. V. SANDIN PwA D. R. HENDRIX AP CHEMICAL EQUILIBRIA D LN L CoMPANY D. SCOTT, JR, ARE H. JARVIS c NONDESTRUCTIVE TESTING GROUP SRfNZ'l'ETf SPECIALTIES CORPORATION G g- EFNI;E?LTZ ig FUEL REPROCESSING 5. SNYDER . . 5.V, P WILLIAMS Fup R TAYLOR A “E . KETCHEN e R.B. OLIVER M T-N. Mcvay b T BROVNING 5 F- R BRUCET o7 BUL 3. D. REDMAN MG J.W. ALLEN " 5. T. ROBINSON, SANDERSON & PORTER D. E. GUSS USAF K SHIELDING FACILITY 3. 1 STURM e SPECTROGRAPHIC ANAL YSIS R, W, MCCLUNG M D. SCHLUDERBERG, BABCOCK & WILCOX H. L. HEMPHILL 55 F. C. MAIENSCHEIN ap s K. W. REBER " T. S. SHEVLIN, OHID STATE EXPERIMENT J. KRAUSE PWA CHEMICAL DEVELOPMENT €. BOUNDS, SEC. ap CORROSION STUDIES 4R MCNALLY" sl 0. E. CONNER M STATION M. F. 0SBORNE s D. E. FERGUSON* T 1. ANNO BMI J. A. NORRIS* AND OTHERS sl W, J. MASON M H. E. ROBERTSON 5 G. I. CATHERS o1 T.V. BLOSSER AP F. KERTESZ G M. T. ROBINSON 5 M. R. BENNETT cT W. R. CHAMPION LAC ? i' i‘d&“" mg WELDING AND BRAZING GROUP CONTRACTOR 5: g: :EBEITSER gg R. M. DUFF cF G. ESTABROOK AP - A P. PATRIARCA M NORTH CAROLINA STATE UNIVERSITY J.D. FLYNN AP G. F. SCHENCK PWA UNIT OPERATIONS 25 e b. FUCKER e R. E. CLAUSING M ENGINEERING PROPERTIES T R PWA 1 V. DIDLAKE e A. E. GOLDMAN M W. K. EISTER® T M. P. HAYDON* AP - M. MASS SPECTROMETRY G. M. SLAUGHTER M 0. SISMAN* 55 < K. M. HENRY AP : BA MCDOWELL M R. M. CARROLL 55 J. T. LONG CcT £. B. JOHNSON AP PRODUCTION OF PURIFLED FUELS C. R. BALDOCK* Sl C' E. SHUBERT M 1. G. MORGAN s S. H. STAINKER CcT T. A LOVE AP J.R.SITES sl C-E o " T MORGAN = G. IONES, IR. cT E G SILVER AP G. j EEgiLAEKELY ::g R g ififfl%? " METALLURGY CONSULTANTS s D. W. LEIGH cT W. ZOBEL AP C.R.CROFT NG T N. CABRERA, UNIVERSITY OF VIRGINIA D. J. KIRBY AP A DOSS e PHYSICAL CHEMISTRY OF N. 3, GRANT, MASSACHUSETTS DESIGN C. O. McREW c J. E. EORGAN MC LIQUID METALS GROUP INSTITUTE OF TECHNOLOGY H. E. GOELLER* cT J. SELLERS ic v SHITH s J. L. GREGG, CORNELL UNIVERSITY "R, P. MILFORD T R. M. SIMMONS AP 1L TRUITT e G. P.SMITH M W. D. JORDAN, UNIVERSITY OF ALABAMA F. N. BROWDER cT D. SMIDDIE IC W. T. WARD C. R, BOSTON M E. F. NIPPES, RENSSELAER o G. G. sTOUT [« "R K. BAGWELL mg J. V. CATHCART M POLYTECHMIC INSTITUTE PILOT PLANT H. WEAVER AP 1 P EUBANKS o H, W, LEAVENWORTH PWA W. F. SAVAGE, RENSSELAER J. W. WAMPLER AP B F. HITCH o J. ). McBRIDE M POLYTECHNIC INSTITUTE H. K, JACKSON* cT W, JENNINGS NG G. F. PETERSEN M G. SISTARE, HANDY & HARMON W, H. CARR* cT TOWER SHIELDING FACILITY G A PALMER s M. E. STEIDLITZ M P. C. SHARRAH, UNIVERSITY OF W. K, LEWIS* cT B C. THOMAS e L. L. HALL M ARKANSAS D e o R. G WILEY Me J. L. GRIFFITH M F. G. TATNALL, BALDWIN-LIMA-HAMIL TON i E-HELSE- . e E. C. WRIGHT, UNIVERSITY OF ALABAMA E e s MOLTEN SALT THERMODYNAMICS MECHANICAL PROPERTIES GROUP F. J. MUCKENTHALER AP E.F. BLANKENSHIP Me D. A. DOUGLAS M G. RAUSA GLM M. BLANDER MC C. W. DOLLINS M F. N. WATSON AP 5. CANTOR MO C. R. KENNEDY PWA E. C. CARROLL ic B. H. CLAMPITT MC J.R.WER, JR. M |. D. CONNER AP 5. LANGER NC 1. W. WOODS M METALLURGY CONTRACTORS J. N, MONEY AP M. B. PANISH M K. W. BOLING M BATTELLE MEMORIAL, INSTITUTE g‘ g ‘L\Jf::JGEHRTwooo Alg L E. TOPOL NG j g. 5GSD;ON M BRUSH BERYLLIUM COMPANY L E. .D. M FERROTHERM SUPPORTING STUDIES V. G. LANE M GLENN L. MARTIN COMPANY B. MCNABB, JR. M METAL HYDRIDES, INC. CONSUL TANT P. A, AGRON E B. C. STOWERS, JR. M NEW ENGLAND MATERIALS TESTING J. E. SUTHERLAND C. K. THOMAS M LABORATORY H. A, RNELL LNIVERSIT BETHE, COl UNIVERSITY C. W. WALKER M RENSSELAER POLYTECHNIC INSTITUTE SUPERICR TUBE COMPANY CONTRACTORS CONSULTANTS INSPECTION GROUP UNIVERSITY OF TENNESSEE METAL HYDRIDES, INC. H. INSLEY A. TABOADA M NUCLEAR DEVELOPMENT CORPORATION T. H. Mcvay R. L. HEESTAND PWA OF AMERICA R. M. EVANS M CONTRACTORS AMES LABORATORY BATTELLE MEMORIAL INSTITUTE CARTER LABORATORIES A, R, NICHOLS, SAN DIEGO STATE COLLEGE UNIVERSITY OF ARKANSAS *PART TIME 229 ‘.&fl Seo W nwu.f..s;‘ xR, nfi.fln“dt.