MARATIN MARIETTA ENERGY SYSTEMS LIBRARIES 3 445k 0250997 O e - OLLECTION ORNL-1771 T his document consists of 197 pages. Copy? 7 of 254 copies. Series A, Contract Neo. W-7405-eng-26 AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT For Period Ending September 10, 1954 W. H. Jordan, Director S. J. Cromer, Co-Director R. I. Strough, Associate Director A, J. Miller, Assistant Director A. W, Savolainen, Editor DATE ISSUED OCY 25 1984 MARTIN MARIETTA ENERGY SYSTEMS LISRARIES A Division of Union Carbide and Carbon Corporation OAK RIDGE NATIONAL LABOQRATORY | i Pos?'foice Box P | Ouk Ridge, Tennessee 3 445k DESD:l:l? 0 CARBIDE AND CARBON CHEMICALS COMPANY 1. G. M. Adamson 2. R. G. Affel 3. C. R, Baldock 4, C. J. Barton 5. E. S. Bettis 6. D. S. Billingten 7. F. F. Blankenship 8, E. P. Blizard 9. G. E. Boyd 10, M. A. Bredig 11. F. R. Bruce 12. A. D. Callihan 13. D. W. Cardwell 14, C. E. Center 15. R, A, Charpie 16. G. H. Clewett 17. C, E. Clitford 18. W. B. Cottrell 19. R. G. 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Hood, Jr.) Technical Information Service, Oak Ridge Division of Research and Medicine, AEC, ORO vi ORNL-528 ORNL.-629 ORNL-768 ORNL-858 ORNL-919 ANP-60 ANP-65 ORNL-1154 ORNL-1170 ORNI.-1227 ORNL.-1294 ORNL.-1375 ORNL.-1439 ORNL-1515 ORNL.-1556 ORNL.-1609 ORNL-1649 ORNL-1692 ORNL-1729 Reports previously issuved in this series are as follows: 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, 195] 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 Pericd Ending December 10, 1953 Period Ending March 10, 1954 Period Ending June 10, 1954 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 all other ANP research at the Laboratory under its Contract W-7405-eng-26. The report is divided into three major parts: |. Reacter Theory, Componenf Design and Testing, and Construction, Il. Materials Research, and [l. Shielding Research. The ANP Project is comprised of about 300 technical and scientific personnel engaged 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 nafionf:l ANP effort. However, the bulk of the ANP research at ORNL is directed toward the development of a circulating-fuei type of reactor. The effort on circulating-fuel reactors was, until recently, centered upon the Aircraft Reactor Experiment. The equipment for this reactor experiment has been assembled, and the current status of the experiment is summarized in Section 1 of Part |. The design, construction, and operation of the Circulating-Fuel Reactor Experiment (CFRE), with the cooperation of the Pratt & Whitney Aircraft Division, are now the specific long-range objectives. The CFRE is to be a power plant system that will include a 60-Mw circulating-fuel reflector-moderated reactor and an adequate heat disposal system. 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-powered reflector-moderated aircraft reactor system. The design work, as well as the supporting research on materials and problems peculiar to the CFRE (previously included in the subject sections), is now reported as a subsection of Section 2, ¥ ““Circulating-Fuel Reflector-Moderated Reactor.’ The ANP research, in addition to that for the Circulating-Fuel Reactor Experiment, folls into three general categories: studies of aircraft-size circulating-fuel reactors, materials problems associated with advanced reactor designs, and studies of shields for nuclear aircraft. These" phases of research are covered in Parts |, H, and lll, respectively, of this report. vii CONTENTS FOREWORDHOII.II..II. ...... # 8% & % & B A E & % 2 B A& S B 0% E 8 o8 & & & ¥ & & 5 2 8 F F &5 8 TN Vii SUMMARY & o0 vvvn s, e DI B 1. PART I. REACTOR THEQORY, COMPONENT DESIGN AND TESTING, AND CONSTRUCTION CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT ..... e et e e e e 9 The Experimental Reactor System . .o v it e ittt iinnnnstanans e s n e e s 9 Characteristics of the Fuel and the Sodium Systems During the Water Tests ... ... o0y, 10 Preparations for Loadingthe ARE .. .. it i ittt i i it i s e onanso 1 Fuel-Concentrate injection Nozzle , . v . .. e e s s e e s s e e e 12 ARE Unloading Experiment. . « o v i v v oo s osanoosnenssassosnnse C et e e ceea. 12 ARE Fuel Recovery and Reprocessing « v v v e v st s sunouvasnsnsasanaasasnssssnnas 12 ARE Pumpss e v v s v o v us te et et ar e £ s ee b ias s s st A s et a0aa st e N an 14 Fabrication and Testing. « « v v s v v v s esosassnes C e e L i e e e s E s e 14 Seal-Gas Pressure-Balancing System . .o v v v ittt ittt e i i ittt it e 14 Radiation Damage to Pump Drives . v v v o v v v e v v oo e N e e e e 14 Radiation Damage to Shaft Seal . . .. .. i n e e et e r st e e s ettt e e b e 15 MOtor TSt v v m v o v v v o a s o s s o asnsconososoesonesssosssonsesssss P s e 15 Reactor System Component Test Loop. v v vt vt i v e ie i it e s neesataessanononssan 16 Operation of Loop. . ... e et e s e e e s s e n e s e 16 Examination of Hairpin Tube. « v v v vttt ittt v enasnoosaressonnososcnonasss 16 Reactor Physics v v v v v v v e v e e et s e st s et e e e h s ae e 16 REFLECTOR-MODERATED REACTOR ,..... T i e s s e 17 Designof the CFRE . .. .. ittt ittt ininans C e i e e s s s e e 17 CFRE Component Development Projects + o v v v et it i snn ottt etonnnennoanasenon 8 Reactor Physics. o v v s s v os s aisrnsnnotisontansnsoneraassssnnsosssansnsss 24 Reactor Calculations v v v v a v v v v sn v nn e Canre f e i a e s e s i ettt e e 24 Beryllium Thermal Stress Test . . v i it ittt ittt nssaaras s eena et eana 24 Beryllium-Inconel-Sodium Systems . . ..o v vttt i i i Cra s e e s e 29 Beryllium-Inconel-Sodium Compatibility Tests. « . ¢ . . . & b e h s e s s Ch e e 29 Mass Transfer Tests in Thermal-Convection Loops ... .. .. Ceeaas s h s e e e e 30 Mass Transfer Tests in Forced-Circulation Loops + v v v v s st vt evonnnvan e sonnans . 32 EXPERIMENTAL REACTOR ENGINEERING 4 v v e v vt v s v s v nsnnsantnoosnsansnanss 34 In-Pile _oop Component Development . . ..o v v v v v v e b e e e s e e s e as s ceen. M4 Horizontal-Shaft Sump Pump . 4 v ottt i ittt st v s s s e nsaanonas Car e e 34 Vertical-Shaft Centrifugal Sump Pump « v v v v s v i it i it i i st ii et i s i e e o s as 37 Hydraulic Motor Pump Drives + v v vvii i i oo e h e s n e e s eea e ae A es 38 Forced-Circulation Corrosion Loops v v v v v v o0 v v e au s s e ae e et e e e e 38 Inconel Loops « v vt vttt et s m e s santsasesoncssnsosenesnssenss L e e e e e 38 Dissimilar Metal Loops « v v v e e v v v e e e o e e e cenacssee s e 41 Gas-Furnace Heat-Source Development . ... oo vovvi v Cee e et e b e e 41 Study of the Cavitation Phenomenon .« o v v v vt it i i ittt e et osnasaronansneses 41 Expansion and Improvement of Thermal-Convection Loop Testing Facilities . .. ..o 0 v v vn v 42 CRITICAL EXPERIMENTS . . v v i et v as L et e e v e e e st e e s 44 Reflector-Moderated Reactor « v v o v v v s v s s s v s s nanessnnsonsoe s e e e s e R . Supercritical-Water Reactor. « v o v v v vt v s eaeeneesi et e s s et e e .. 47 PART Il. MATERIALS RESEARCH 5. CHEMISTRY OF MOLTEN MATERIALS .. ...... e R X | Solid Phase Studies in the NaF-ZrF -UF , System .o oivvin ooy e M Visual Cbservation of Melting Tflmperatur&s in the NaF-UF Sys'rem .................. . 35 Phase Relationships in UF -Bearing Systems. ... ... Gt e a e e es s esnnes e ¢ haanes 56 UF3iinZrFA-BeqrinQSystems.. ......... ettt e e e e e I UF, in NaF-KF-LiF Mixtures «oo0vvve it e e e Y UF, in NaF-RbF-LiF Mixtures . ....... .o i e e crr e 58 UF, in the Individual Alkali-Metal Fluorides .« v vvcviiv v i, 58 UF; in Binary Alkali-Metal Fluoride Systems .. ..o v, P 60 Purification of Rubidium Fluoride . ... oo i iv i ettt 60 Chemical Reactions in Molten Salis + . ... .. ... et et chee e verrseens 60 Chemical Equilibria in Fused Salts v v v v i i ittt ittt st anoss oo ssssaeas 60 Stability of Chromium Compounds in Molten Fluerides ........... < X Reduction of NiF, by H, in NaF-ZrF, Systems. . ..o vvviv i e N« Reduction of FeF, by H, in NaF-ZrF, Systems ..o uvn. et e e e 64 Preparation of Various Fluorides v v v v vttt v it i it it i st et i nansaoens 67 Fundamental Chemistry of Fused Salts v v v v i v v i vt vnenneasnsnrnesrnoesnssnsess O EME MeaSUrements « o v v v o s v oo vt o o vt oot s aoosussasennsesaneanennsssssess 67 Solubility of Xenon inMolten Salts. . . v v v v i v vt it e it e i ittt e i e i e 70 X-Ray Diffraction Studies in Salt Systems . v iviiiiiiriornnsersnsnnsesssssaess 70 Physical Chemistry. v o v o v v v v oo n o e et et e et e e e 72 Production of Purified Molten Fluorides . . o i vttt i ittt v in it en et oo s e 72 Use of Zirconium Metal as a Scavenging Agent + v v v v v v v v e e C e e e 72 Purification of NaF-ZrF, Mixtures by Electrolysis « oo vvvviviiiivacnen P e e /3 Preparation of UF3wBenring Fuels oo i v oo i n ot h e e e e Gt e e e 77 Alkali-Metal Fluoride Processing Facility. « v v v v a0 v v b e e e e seeeee. 78 Production Facility. v v v s v v v s v i v e i e oo et e e et e e 79 In-Pife Loop Loading v v v vt vt v v vnoennnns e e et et anes 19 Chemistry of Alkali-Metal Hydroxides . . . v v v v v v v v et e e e s ceresees 80 Purification. « v v v vt i vt i e et e e an et e e n s e r e e . 80 Reaction of Sodium Hydroxide withMetals . . ..o v oo i it i e ceseeras. 80 8. CORROSION RESEARCH v v it i ittt s i e nnse oo C et a et 81 Static and Seesow Corrosion Tests . v v v v v v e v ot v s v o s et e e e e e 82 Brazing Alloys in NaF-Zrt ,-UF and in Sodium ... ... Ce e T - 7. Special Stellite Heats in NaF ZrF -UF, and in Sodium. .. oo e T - < HasfelloyRinVariousMedio........... ............. C e ne e e st e 85 Incone!l in Molten Rubidium., v v oo v o e v vt Ce e e et et e e e e ae e 86 Carburization of Inconel by Sodium. « v v v v v v v i v v v v v e P -1 Special Tar-Impregnated and Fired Graphite . v v . . .. .. L Ceramics in Various Media oo ittt vn e I Fluoride Corrosion of Inconel in Thermal Convection Loops ... ... Cen e e e e 95 Effect of UF, in Z¢F ,-Base Fuels . ... ... f e et e e s e e 95 Effect of UF, in Alkali-Metal Base Fuels .. ..o vvvvv vy e e e ceeraaeaae. 96 Effect of Zirconium Hydride Additions to Fuel .. . oo v v v ittt it iiainon 98 Effect of Uranium Concentration « v v v v v v v v o v s oen oo e e 98 Effect of Inconel Grain Size . ... ... C e st s e rn e es e st e e s e e s e e e .. 98 Fluoride Corrosion of Hastelloy B in The rmc:l Convechon lLoops v vii v e vieeanes 98 Lithium in Type 316 Stainless Steel « v v i vt it it it i i i et ettt ettt a e 99 Fundamental Corresion Research o v o v v v v v i v v i v i v o s Gt s e 100 Mass Transfer in Liquid Lead . . o v v e v i n s v v v e 100 Flammability of Sodium A”oys e s st e s e s et . e h e et e e . . 102 Thermodynamics of Alkali-Metal Hydrox:des..........,....................... 102 Chemical Studies of Corrosion. « v v v v v v ettt vt i e erenan Cie v aneaseses 108 Effect of Temperature on Corrosion of Inconel ond Type 316 Stainless Steel . . .. ... vee... 108 Corrosion by Fission Products . ..o i i i en st st ennnesons s s 109 Controlled-Velocity Corrosion Testing Apparatus . .. . . e e e erenas 109 Reaction Between Graphite and Fluoride Melt. . v o v . i i e vt e it nnenenenanasaa. 110 Lithium Fluoride Castings ...... e s r e e e e P 110 . METALLURGY ... .. iiiinineenan, e e s s e senavaranas 111 Stress-Rupture Tests of Inconel . . . .. .. v v v v co ceee . bee coes 112 High-Conductivity-Fin Sodium-to- Alr Rudsator s a e e s . veeraeseass 116 Investigations of FinMaterials + « v v v v s e i in i it i s nennesrennonooenanssnseess 117 Development of Brazing Alloys for Use in Fabricating Radiators . v v v 4. .. cha e 118 Radiator Fabrication .+ . . . . G r e e e e “a e e a e e e . 120 Nozzles for the Gas-Fired qumd Mefal Heoter System .« . v v s O (0] Special Materials Fabrication Research. . ..... ch s e I vaw e 121 Stainless-Steei-Clad Molybdenum and Columbium. e h et st e e e e 121 tnconel-Type Alloys + v vv v v e u v v e e e e e s e e e s e 122 Nickel-Molybdenum-Base Alloys .+ . v v e v vt vt o vt netevnonnnnaen e e et 122 Duplex Tubing o v o vt e sttt a s oo noneeesanosanotonsnonsnsossnennssesss 123 Sigma-Phase Alloys . . . v ittt i vt ittt ttoncanonseenonss e 124 Boron Carbide Shielding. v v v 0 v s v v v v v v v vt e . Cee s e e e 124 Tubular Fuel Elements & . o it i i it ittt it i it ittt et et entnesasanan 124 Metallographic Examination of a Fluoride-to-Sodium Heat Exchenger . ... ... ... caeees 124 . HEAT TRANSFER AND PHYSICAL PROPERTIES . .o e . ‘e 127 Physical Properties Measurements « v v o v v v v st o st v neaasens e RPN e e 127 Heat Capacity v v o v oo v o st v i s s s o nnoessonaosassssnassssrensnnse . 127 Density and Viscosity v o v o v st st s vn oo tasessnsonnesosssnss . A VY Thermal Conductivity & v s o v e v it s v i e st s o s v aononsroorsorosssesssanosanes 129 Electrical Conductivity . oo v v v et a s e e e . . e . . 129 VapPOr PressSure o o v v o o o s s 0 s 60t asiesorenonssenses c. . e h e e . . 129 Fused-Salt Heat Transfer v « v v v o vt v v s e st tnonesaonssontosnoasonsssannoasss 130 Transient Boiling Research . ...... C e s e et e e e 130 Fluid Flow Studies for Circulating-Fuel Reactors . .. ........ ettt et 131 Heat Transfer Studies for Circulating-Fuel Reactors .. v v v v v vt o vt s e v sennannasas 131 . RADIATION DAMAGE + v v v v s v vt s i et st st et o et ensonnsnnesanonansness . 134 MTR Static Corrosion TestS « v v s evesssasnssssssnsnossosssssesarenaasssanes 134 Fission Product Corrosion Study + o v cu e v ve v on v Ces se e e e creeaes 135 Facilities for Handling Irradiated Capsules & o v v v v v v s v i e it s vt e s s st ansnonnasons 135 Analysis of lrradioted Fluoride Fuels for Uroniom . . v oo o0 . caes . . . . 136 High-Temperature Check Valve Tests o v v v v v v v v Pean e s e e . 137 Miniature In-Pile Loop — Bench Test v v v o s vt o i entntnntatossasososnnssssases 137 Life Tests on an RPM Meter, Bearings, and a Small Electric Motor Under lrradiation. . . . . s ees 137 Removal of Xenon from Fused Fluorides « v v v v s v v 0o e e e e e sa s s e aaaas 140 CoF UF Imadiation, v u v v s ss v i st it o s i i nnans v o ea s ... 140 LITR Horrzonmf Beam-Hole Fluoride-Fuel Loop s « v v v v v v v e v v s s et nonnenns e ea 141 ORNL Graphite Reactor Sodium-Inconel Loop ...... e s e eas 142 Creep and Stress- Corrosnon Tests v o v nn et tanenan e e . . ere. 142 xi Remote Metallography v v i v i v o v i ittt te it et s e s ne st o onnasnnsstsanassssnas 144 Fission-Fragment Annealing Studies . v v v v s v v v v et e s e e e 145 High-Temperature, Short-Time, Grain-Growth Characteristics of Inconel. .. v v oo v v v i v v in 145 BNL Neutron Spectrum ~ Radiation Damage Study « v o0 v i it it i it ittt i e e 145 MTR Neutron-Flux Spectra — Radiation Damage Study . v v v v v s v v st v st st i v c oo 147 10. ANALYTICAL STUDIES OF REACTOR MATERIALS ... i ittt ittt e s s annnas 148 Analytical Chemistry of Reactor Materials o v v v v i i it e i i i it e it i s i i e st enaan 148 Determination of Oxygen in Fluoride Fuels. . . v o i ittt it i i et it e v i s i i s aenn s 148 Oxidation-Reduction Titrations inFused Salts . . . oo v v s i i it i i i i e i en e 150 Polarographic Studies in Fused Ammonium Formate + v v v v v i v vt vt i i i st i e v nnnas 150 Conversion of UF ; and UF, to the Respective Chlorides with BCly. .o vu v v vvvovonsy 151 Solubility of Tri- and Tetravalent Uranium Fluorides in Fused NaAICl, v oo vvv i onn e, 151 Oxidation of UF 5 with Oxygen . v v v v v v vi i vns et e s s e ceee. 152 Determination of Lithium, Potassium, Rubidium, and Fluoride lon in NaF-KF(RbF)-LiF-Base Fuels . v v v v v v v v it i v v v vt v a e C e e e e e 152 Oil Contamination in ARE Helium., v« o v v v i it i i i s it it e s e s s e s veanane 154 Petrographic Investigations of Fluoride Fuels. i i e v i vt ittt s e vt v o esans 154 Summary of Service Analyses. ¢ v v v i ittt i i et c i fe st e s assuae e 154 PART I, SHIELDING RESEARCH 11, SHIELDING ANALY SIS Lo ittt ittt i it et a st a st st oot aossoannsnans 157 Slant Penetration of Composite Slab Shields by Gamma Rays ... ... e et e e 157 Air Scattering of Neutrons. v v v v v vt v vt et v i et osnaasn et e et 158 Single Anisotropic Air Scattering in the Presence of the Ground (Shielded Detector) .. ... .. 158 Single Isotropic Air Scattering in the Presence of the Ground (Unshielded Detector) . ... ... 158 Formulas for Multiple Scattering in a Uniform Medium o « o v oo v v v v o e v C e e 159 Ground Scattering of NeUtrons 4 v v v v v v it i st it ittt st s e nonsossassnoens 162 Focusing of Radiation in a Cylindrical Crew Compartment « v v v v v v v v v v s Cea e e 163 12. LID TANK SHIELDING FACILITY 4 it it ettt vt c s s s aaton s snssaossnonasasas 164 Reflector-Moderated Reactor and Shield Mockup Tests v v v v v i i i it i i et i e i e e en o 164 Effective Removal Cross Sectionof Carbon & v v v v vt it i i it it it i i e s i e st e s s 164 GE-ANP Helical Air Duct Experimentation « « v v « v v e s v s o s nnonssensettnennenss 166 13, BULK SHIELDING FACILITY & v v ittt i e ivaean st astnntoosooenassns P e 168 Reactor Radiations Through Slabs of Graphite. s v v v v v e v vt e i e v i st n e e bene e e 168 Reactor AIr Glow v v v v v v oo v oo s oo oot tnnsnnsennsesnssssotensossnsansessses 170 Fuel Activation Method for Power Determination of the ARE . .o v v v v v i v v v v i v i v an 173 14, TOWER SHIELDING FACILITY & v v vttt ittt vt s e st s aotoossossoasnesnasssns 175 Fast-Neutron Ground and Air Scattering Measurements v v v v v v vt v v v oo v o s e e a e e 175 Calorimetric Reactor Power Determination « v v v v v s v v e v v vt c o st ton oot tonasnsas .. 175 GE-ANP R-1 Divided-Shield Mockup Tests ¢ « v v v v v v e e st e s a oo osonasosansons .. 178 PART V. APPENDIXES 15. LIST OF REPORTS ISSUED DURING THE QUARTER v it it i s v v s v st s s snnsononos 183 ORGANIZATION CHART OF THE AIRCRAFT NUCLEAR PROPULSION PROJECT .......... 185 x11 ANP PROJECT QUARTERLY PROGRESS REPORT SUMMARY PART |. REACTOR THEORY, COMPONENT DESIGN AND TESTING, AND CONSTRUCTION Circulating-Fuel Aircratt Resctor Experiment The water tests of the fuel and the sodium cir- cuits of the ARE system at room temperature were completed (Sec. 1). The sodium circuit was pres- sure-filled with water from the sodium fill tanks, while the fuel system was vacuum-filled to ensure the elimination of gas pockets in the fuel system. Both systems were found to function properly, and the filling, circulating, and draining operations were effected with a minimum of difficulty. After the water was drained, the system was dried by heating it to approximately 600°F. The electrical heating system was found to be satisfactory in the check thus afforded. The final system completion work is now under way, that is, removal of the sodium system reactor bypass, completion of the fuel-enrichment system installation, completion of thermocouple and insu- iation installation, and other minor modifications. When this work is completed, charged to the system and the high-temperature checkout phase of the experiment will be initiated. The neutron source was put into the reactor, and the nuclear checked out. Also, mechanical checks were made of the per- formance of the safety and control rods. The building electric and helium systems were made ready to accommodate the loading facilities, and final arrongements were completed for attaching the fuel-sampling equipment. Radiation damage experiments indicated the de- sirability of providing gamma shielding at several points where elastomers were used (belts, dia- phragms, etc.}. Where shielding was impractical, the composition diaphragms were replaced with sodium will be Cinsfrumentation was metal diaphragms. The results of operation of the reactor system component test loop at K-25 are encouraging in that the loop has now been operated without major difficulties for more than 1800 hr. None of the minor difficulties encountered would indicate that serious problems might arise in operation of the ARE. Reflector-Moderoted Reactor The program for the development and construc- tien of the Circulating-Fuel Reactor Experiment (CFRE) has been outlined, .and many of the de- velopment projects are under way. Tentative design data have been compiled and o flow sheet has been prepared (Sec. 2). The first priority development project, the test of beryllium in con- tact with sodium and Inconel under thermal stress, has been completed. The results of this test were needed in the determination of the dmcunf of poison to be expected in the reflecter. The test indicated that beryllium will not crack under the thermal stresses involved in the temperature range 1000 to 1300°F. Since corrosion and mass transfer, as well as thermal stress, will be important in the beryllium-lnconei-sodium system, many static and dynamic tests under varieus conditions have been made. There is considerable evidence to indicate satisfactory compatibility in the beryllium-Inconel- sodium system at temperatures up to 1200°F. The temperature coefficient of reactivity for the CFRE was computed on the UNIVAC and, for rapid - temperature changes, was found to be -3.5 x 10-°/°F. The critical mass computed for the rhombicuboctahedral critical assembly was rede- termined because of errors found in the originaol data. The redetermined value agreed closely with the experimental value, but, since the criticol mass is not very sensitive to errors in detail, further evaluation of the agreement must await additional experimental results, Experimental Reactor Engineering The emphasis in the engineering work is now on development of components for an in-pile loop for insertion in a horizontal beam hole of the MTR and the design ond construction of forced-circu- lation corrosion testing loops (Sec. 3). The in-pile loop for insertion in the MTR is o joint ORNL and Pratt & Whitney Aircraft Division project. It is to circulate proposed fuel mixtures in the high-flux of the MTR so that the extent of radiation damage to materials of construction and the effect of radi- ation on the fuel can be determined. Two types of pumps have been developed for in-pile use: a L | ANP QUARTERLY PROGRESS REPORT vertical-shaft centrifugal sump pump for instal- lation external to the reactor shield and o hori- zontal-shaft sump pump for insertion inside a beam hole. A turbine-type impeller is being considered for the horizontal-shaft pump because it would have the advantage that both the inlet and dis- charge could be at the bottom. Hydraulic motors of suitably small dimensions have been found to be satisfactory drives for these pumps. Two series of Inconel forced-circulation cor- rosion loops for circulating fluoride mixtures are being developed to meet the following require- ments: (1) a Reynolds number of 10,000 with temperature gradients of 100, 200, and 300°F and (2) a temperature gradient of 200°F with Reynolds numbers of 800, 3,000, and 15,000. The maximum fluid temperature is to be 1500°F, A study is under way of the cavitation phe- nomenon associated with operating liquid metal systems at elevated temperatures, high flow rates, and high pump speeds. A correlation of fluid-flow- noise infensity with pressure data was noted. The number of stations available for convection- loop testing was increased from 18 to 31 and the basic design of the loops was simplified. Various means of heating the loops and of making operation of them more automatic are being studied. Critical Experiments The first step of the present critical experiment program was the construction of a small two-region reflector-moderated reactor to provide experimental data on a system of simple geometry and materials for use in checking the calculational methods being used (Sec. 4). The core consists of alter- nate sheets of enriched uranium metal and Teflon and is surrounded by a beryllium reflector. The uranium loading can be varied, within the specified dimensions, to make the system critical. The assembly was looded as prescribed by the multi- group calculations but was not critical. However, when the calculations had been corrected to take into account errors in the original data, a new attempt to achieve criticality with the prescribed loading was made. The corrected prescribed loading was 20.9 to 22.75 |b of U??® and the experimental loading was 24.35 1b of U235, A larger critical assembly of the same shape is to be constructed that will consist of three regions, with the beryllium island and the reflector sepa- rated by the fuel annulus. A further check on the calculational methods will be obtained. 2 oo PART H. MATERIALS RESEARCH Chemistry of Molten Materials Studies of the fluoride systems of interest as reactor fuels were continued, with particular em- being given to systems in which the uranium-bearing component is the less corrosive UF, or a mixture of UF, and UF, rather than UF, alone (Sec. 5). Recent attempts to correlate the anticipated reduction of UF, in the UF -bearing melts with wet chemical analysis for UF, and UF, and results of petrographic examination show some surprising anomalies. When UF, dissolved in LiF, in NaF--ZrF4-UF4 mixtures, or in NaF-LiF mixtures is treated under flowing hydrogen at 800°C with excess uranium metal, 90% or more of the UF, is reduced to UF,. However, when this technique is applied to UF, in NaF-KF-LiF mixtures, the reduction is only 50% complete at 800°C and, perhaps, 75% complete at 600°C. Petragraphic examinations of the specimens reveal no complex compounds of tetravalent uranium; it is possible that the UF, is “hidden’” in solid solutions or in complex UF4-UF3 compounds in which it is not at present recognizable. Solid phase studies of the NaF-ZrF -UF, system were initiated following completion of studies of the NaF-ZrF, system, and to date no ternary com- pounds have been discovered. A tentative equi- librium diagram was prepared. phasis A methed for large-scale purification of rubidium fluoride has been developed that can be used if material sufficiently free from cesium cannot be obtained from commercial sources. Fundamental studies of the reduction of NiFF, and FeF, by H, in NaF-ZrF, systems were made as a means of determining possible improvements in purification techniques. Also, methods for preparing simple structural metal fluorides were studied. Large quantities of puritied ZrF ,-base fluorides were prepared for engineering tests at ORNL and elsewhere, and the demand for purified fluorides of other types is rapidly increasing. Therefore preparation and purification methods have been studied intensively in an effort to lower production time and costs. Developments indicate that the price of purified NaZrF, and NoF-ZrF ,-UF, mix- tures may be halved in the next few months. Corrosion Research The static and seesaw corrosion testing fa- cilities were used for further studies of brazing - o alloys, special Stellite heats, Hastelloy R, In- conel, graphite, and various ceramics in sodlum flucride fuel mixtures, and other mediums (Sec. 6). The brazing alloy 67% Ni-13% Ge~11% Cr-6% Si—-2% Fe~1% Mn was found to have good corrosion resistance in fluoride fuels and fair corrosion resistance in sodium and therefore will be useful for the fabrication of many reactor components. In the thermal-convection loop studies, UF,- bearing fuels were tested in Inconel. Hot-leg attack is not found in Inconel loops in which ZrF - base fluoride mixtures with the uranium as UF, are circulated. A deposit is, however, found on the hot-leg surface. Only preliminary information is available, but it appears that neither attack nor a hot-leg layer is found with alkali-metal-base fluoride mixtures containing UF,. Mixtures of UF, and UF, result in a reduction in attack from that found with only UF,, but some attack is present, and in high-uranium-content systems the attack may be significant, Several Hastelloy B loops have now been suc- cessfully operated in both the as-received and the over-aged conditions. In both coses @ consider- able increase in hardness occurs during operation. With ZrF, -base mixtures containing UF,, very little attack is found, even after 1000 hr. Thermal-convection loop tests of molten lithium in type 316 stainless steel were operated for 1000 hr. There were no signs of plug formation, and only a small amount of mass transferred material was found in one loop. Alloys of 45% Cr—55% Co, Ni-Mo alloys, and the Fe-Cr-base stainless steels have been shown to be more resistant to corrosion and mass transfer in liquid lead than are the pure metals. Their resistance to mass transfer can probobly be related to the formation of interme- tallic compounds. | Metallurgy Creep and stress-rupture testing by the tube- burst method has been studied intensively (Sec. 7). In the tube-burst tests, a tube that is closed at cne end is stressed with an internal gas pressure. The stress pattern introduced into the specimen in this fest approaches the stress pattern that will be found in ANP-type reactors. Apparatus for the tests has been constructed, and a theoretical analysis has been made with which o check en the experimental results can be obtained. In the investigation of high fhermql-conductwlty PERIOD ENDING SEPTEMBER 10, 1954 fins for sodium-to-air radiators, stress-rupture and creep tests were made on copper fins with various types of cladding at stress levels between 500 and 2000 psi at 1500°F. The tests show that for o 1000-hr exposure in air, stresses greater than 500 psi and less than 1000 psi are tolerable; that s, in this stress range there is no indication of brittleness in the core or oxidation of the core due to cladding failure of type-310 stainless-steel- clad copper fins. From the over-all considerations of melting point, oxidation resistance, dilution of fin and tube wall, formation of low-melting eu- tectics, and flowability, it was found that Coast Metals alloy 52 was the best brazing alley for use in the construction of radiators with high- conductivity fins. A sodium-to-air radiator with 6 in. of type-430 stainless-steel-clad copper high- conductivity fins was fabricated by use of a combi- nation heliarc welding and brazing procedure. Packed-rod nozzle assemblies were fabricated for the 100-kw gas-fired liquid-metal-heater system, and work was started on the formation of duplex tubing. An attempt is being made to prepare tubing that will have good corrosion resistance on the inner surface and oxidation resistance on the outer surface. Attempts are being made to find new alloys in the nickel-molybdenum system that will have better high-temperature strength and fluoride corrosion resistance than Inconel has. Hastelloy B satisfies these requirements, but it has poor fabrication properties and oxidation resistance; it also loses its ductility in the temperature range of interest for application in high-temperature circulating-fuel reactors. lInvestigations are under way to find a suitable melting practice and heating treatment that will increase the ductility of Hastelloy B in the temperature range of interest. Heot Transfer and Physical Properties The enthalpies and heat capacities of NaF-ZrF - UFJ (65-15-20 mole %) and of LEF-N0F~UF4 (57.6- 38.4-4.0 mole %) were determined (Sec. 8), The thermal conductivity of NaF-KF-UF, (46.5-26.0- 27.5 mole %) was found to be 0.7 Btu/hr-f’rz(ol:/ff) that for KF-LiF-NaF-UF , (43.5-44.5-10.9-1.1 mole %) was 2.0, and that for LiF-KF—UFd (48.0-48.0- 4.0 mole %) was 1.4. A new electrical conduc- tivity device has been constructed and has been successfully checked with molten salts of known conductivity. ANP QUARTERLY PROGRESS REPORT A device for studying the rates of growth of tube-wall deposits has been successfully tested with a simple heat transfer salt. Also, a hydro- dynamic flow system for studying the reflector- moderated reactor flow structure has been tested. A mathematical study of the temperature siructure in converging and in diverging channel systems that duct fluids with volumetric heat sources and a study of wall cooling reguirements in circulating- fuel reactors were made. Rudiation Damoge Additional irradiations of Inconel capsuies con- taining fluoride mixtures were carried out in the MTR (Sec. 9). UF,-bearing fue! has been examined, and it shows no corrosion, in confrast to that found previously with UF -bearing fuels. Inspection of the LITR in-pile loop, which failed prior to startup, disclosed that the failure was caused by ¢ bregk in the weld connecting the pump discharge nipple to thz fuel line. Design revisions and refabrication of some of the parts are in prog- ress. Developmental work continued on a smaller loop for operation in an L.ITR A-piece. Only one capsule containing a Detailed exomination of an Inconzsl loop which circulated sodium at high temperature in the ORNL Graphite Reactor showed no evidence of radiation- induced corrosion. were corried out in the hot cells on irradiated fue! plates for Frait & Whitney Aircraft Division, and studies were made on annealing-out of fission- fragment dumage. Work continued on examination of wire and multiple-plate-type units for GE-ANP, Metallogrophic examinations Anclytical Studies of Reactor Materials The primary analytical problem continugs to be the seporation ond determination of trivalent and tetravalent uranium in hoth NqF-ZrFS and NaF- KF-LiF-bose fuels (Sec. 10). tentiometric titration of UFJ with metallic A successful po- in molten Naldrf zirconium was gaccomplished by means of polarized platinum g¢lectrodes. The sclu- bility of UF, in NaAICl, was determined to be 18 mg/g ot 200°C, in contrast to a solubility of UF, of less than 1 mg/g. NaAIClA is expected to dissolve tetravalent ura- Therzsfore molten nium selectively from the fuels. Calibration measurements have been completed on the apparatus for the determination of oxygen as metallic oxidas in reactor fuels. The reaction involves the hydrofluorination of the oxide and measurement of the increase in conductivity of liquid HF as o function of the water formed. In- vestigations are being made of the oxidation of Ut', and of UF, with oxygen at slevated tempera- tures, An improved method for the determination of lithium in NaF-KF-LiF-base fuels was developed. Also, studies were made of the solubilities of potassium, rubidiym, and cesium tetraphenyl- borates in various organic solvents to ascertain differential solubilities. PART ill, SHIELDING RESEARCH Shialding Analysis Application of the Monte Carlo method to the calculotion of gamma-ray penetration of crew shie!d sides has been worked out, and the method appears to be quite satisfactory for this type of problem (Sec. 11). Considerable progress has been made on understanding the Tower Shielding Fa- cility (TSF) mecsurements of ground and air scattering. Expressions have been derived which describe the effect of ground interference with the air-scattered flux, and thus it is now possible to estimate ground-scaitered radiation both «t the TSF and in on airplane at landing and ftakeoff, Calculations have been set up for evaluating multiply scattered radiation in air. The values obtained will be of considerable importance in studies of the highly asymmetric but light shields of several current designs. Lid York Shislding Facility Preparations for a second series of shielding tests for the reflector-modercted reactor ot the Lid Tank Shielding Facility (LTSF) have included the construction of a large tank which will hold all the components of the mockups and the irradi- ation of the UF C,F, solution which may be used to simulate the reactor fue! in mockup ex- periments (Sec. 12). Measurements of the removal cross section of carbon made in a continuous have in a value of o, = 0.750 barn. This is to be compared with the previous value of 0.81 1 0.05 barn measured behind a solid slab of graphite. Thermal-neutron flux measurements have been mode bevond two con- carbon medium resulted figurations of GE-ANP helical air ducts, a single duct and a triangular array of three ducts. Meas- urements beyond g 35-duct arroy will begin soon. ‘Bulk Shielding Facility The use of a graphite reflector as a shieid com- ponent has been investigated at the Bulk Shielding Facility (BSF) with measurements of attenugtion through various thicknesses of the material (Sec. 13). The fast-neutron spectrum of the BSF reactor from 1.3 1o 10 Mev through 1 ft of graphite was also measured. Removal cross section values for the carbon were determined to be 0.82 barn for a 1-ft slab, 0.84 barn for a 2-ft slab, and 0.80 barn for a 3-ft slab. An experiment has been performed at the BSF to provide an experimental basis for future esti- mates of the amount of visible light around o nuclear-powered aircraft. Measurements in an air- filled tube placed against the reactor have indi- cated that the maximum glow will occur at!a pressure correspondmg to an altitude of ubout 30,000 ft. in a proposed method for the determination of the power of the ARE, the relative octivity of the PERIOD ENDING SEPTEMBER 10, 1954 fuel samples exposed in the ARE ond in the known flux of ancther reactor is meosured. The method has been tested at the BSF, Tower Shielding Facility Measurements of the ground and goir scattering of neutrons have been made at the Tower Shielding Facility (TSF) (Sec. 14). A preliminary anolysis indicates that the contribution at the maximum altitudes (around 200 ft) is between 2 and 5% of the total scattered neutrons for differential ex- periments. A new procedure for a calorimetric determination of the power of the TSF reactor has been devised which is based on the relationship between the rate of temperature rise in the water of the reactor tank and the reactor power. The resuits of three experiments were consistent to within 1%. The next series of tests at the TSF will be on the GE-ANP R-1 divided-shield mockup. Part | REACTOR THEORY, COMPONENT DESIGN AND TESTING, AND CONSTRUCTION 1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT E. S. Bettis J. L. Meem Aircraft Reactor Engineering Division THE EXPERIMENTAL REACTOR SYSTEM A general revision was found to be necessary in the ARE ‘in that shadow shielding had to be installed, at several points where elastomers were used (belts, diaphragms, etc.),to provide protection from gamma fields that would: have produced pro- hibitive degrees of radiation damage. At some points, where shielding was impractical, the compo- sition diaphragms were replaced by metal dia- phragms. The entire ARE system has now been completely checked out as far as rocometemperature preoper- ational tests are concerned. Both the fuel and the sodium circuits have been operated simultaneously with water as the circulated liquid in each circuit. The moderator volume of the reactor was bypassed for these tests to keep water away from the be- ryllium oxide blocks. The sodium circuit was pressure-filled with water from the sodium fill tanks, while the fuel system was vacuum-filled to ensure the elimination of gas pockets in the fuel system. Both systems were found to function quite satisfactorily, and the filling, circulating, and draining operations were effected with a minimum of difficulty. During these operational shakedown tests, it was found that the gas lines to the sump tanks could not be kept pressure tight because the Swagelok fittings developed leaks. All these joints have therefore been silver-soldered to correct this situation, The pumps performed satisfactorily, and useful system curves for actual future operation were obtained. After the water tests of the system had been completed, it was necessary to remove oll water from the system. This necessitated heating the entire system to o temperature of approximately 600°F. This operation provided the first check out of the electrical heating system. performance was gratifying. Such troubles as arose were corrected rather easily, and, in general, the heating of the system was effected with very little ditficulty. A CO, cold trap was incorporated in the off-gas line while the system was being heated to collect all moisture that was driven from the system. Dry Here again the air was admitted to the fill tanks, and both the fuel and the sodium systems were flushed with this dry air, the air leaving the system through the CO, cold trap. When the dew point of the exit air from this cold frap reached ~30°F, the system was considered to be dry and the electrical heat was turned off. Befare the heat was removed from the system, however, the temperature on the thermal barrier doors around the heat exchangers was raised to approximately 1000°F and operation of the doors was checked., Even though the doors had been reworked to eliminate sticking, it was found that the doors were still binding. When the barrier doors were removed, it could be seen that the side guides whichrun the height of the door were binding in the guide ways of the door frames. This binding was resulting from the bowing of the guide rails caused by the thermal gradient which exists in the door. The door does not require these side guides, since the door housing provides ample guides for the normal functioning of the doors. Therefore the side guide rails are being removed so that there will be no further binding of the barrier doors. It was also noted while the system was hot that a considerable volume of kerosene was being driven out of the beryllium oxide moderator blocks by the high tempercture. The kerosene had been absorbed by the blocks during the cutting operations. To remove the kerosene, a vacuum was pulled on the moderator volume of the reactor. A CO, cold trap was inserted in the vacuum line to the moderator, and while the reactor wos maintained at a temper- atyre of approximately 450°F, the vacuum pump was run continvously. This operation was continued until no more condensate of the kerosene distillation was being collected in the cold trap. Approxi- mately 2 gal of kerosene was removed by this procedure. |t is possible — in fact, fairly certain - that some residue was left in the moderator blocks. It was the consensus of the chemists that this would not have an adverse effect on the sodium coolant, and therefore nothing further is to be done about the small amount of tarry residue remaining in the reactor moderator blocks, ’ ANP QUARTERLY PROGRESS REPORT After the system had been thoroughly dried, all heat was turnad off and the final system completion The major work in this category consists in removing the sodium system reactor bypass, removing the flanged filter pots in the sodium purification system, and completing the Ther“ mocouple installation also has to be completed, as well as thermal insulation of the fuel pump bowls. work was bequn. enrichment system tie-in fo the fuel circuit, The decision to remove the flanged filter pots and o substitute weld-sealed filter pots was mads It wos decided at that time, however, to postpone this substifution until after the woter test. The original filters were useful in cleaning up the water used to test the operation of the sedium circuit, Completion of the fue! enrich- ment system in no way affected the water test, and hence this work was postponed in the interest of the woter test as expeditiously as The fuel injection system is to be woter-tested independently of the fuel system. Insulation of the fuel pump bowls was postponed several months ago. completing possible. to allow more room for access to the pumps during the water tests. Since the pumps have riow been checked out, the insulation job will be completed, The system should be ready for charging with sodivm early in September, and the high-tempera- ture check out phase of the experiment will be initiated. While the sodium checks arc being run, the hot.gas leck test of the fuel system will be performed. This test involves loading the fuel system with a mixture of helium and krypton at a pressure of about 15 psi. While this gas is being circulated at 1300°F by the fuel pump the annulus circuits will be monitored by mass spectrographic methods for the presence of krypton in the annulus. The nsutron source was put in the reactor, and the nuclear instrumentation was checked out. Al three (the twe regular plus the spare) fission chambers were checked, and the count-rate vs chamber-voltage curves were plotted. The data obtained provided the necessary information for establishing the operational plateou for chamber. Also, mechanical checks were made of the performance of the safety and control rods. A modification of the rod-cooling circuit was required in order to minimize the heat loss from the center of the reacter. lation of propertioning orifices to get coivect helivm flow around ths fission chambers and the safety rods. each This riecessitated instale 10 Final modifications of the sodium and the fuel loading systems were completed, and the building electric ond helium systems were made ready to accommodate the loading facilities when they are brought into the building. Final arrangements have been completed for atiaching the fuel-sampling system. [his sampling system will not be left connected when the power run is initiated. The engingering prints for the entire ARE instal- lation are now up to date, and the electrical prints, in particular, were used successfully in checking out the many heater circuits involved in the experi- ment, CHARACTERISTICS OF THE FUEL AND THE SODIUM SYSTEMS DURING WATER TESTS The characteristics of both the fuel and the sodium systems while circulating water have been determined. Upon removal of the fuel heat ex- changer bypass loop, a glass rotameter was tempo- rarily installed in the fuel circuit between the reactor outlet and the heat exchangers. A direct calibration of the high-temperature fue! rotameter against the glass rotameter was obtained with water as the fluid. Conversion of water flow to fuel flow! is made by 1/2 PF - P/2 92 pfz 9y P — Pf] ’ pf'l where g, = watsr flow (gpm) , g, = fuel flow (gpm) , p;, = water density (g/cm?) , 1 Pr, = fue! density (g/cm3) , pp = float density (g/em®) . During the period of water operation, data on pump speed vs flow rate were cbtained (Fig. 1.1). Also, the pressure head from the pump suction to the reactor inlet was measured (Fig. 1.2). The system characteristics, as shown in these two curves, should be the same during operation with fluorides. ]A. L.. Southern, Discussion of the Rotameter Used in the ARE Fuel Circuit, ARE Filas. ORNL—-LR--OWG 3087 2000 - FUMP SPEED trom) O < | | \ | 0 40 20 30 40 50 80 70 20 S0 FLOW RATE (gom) Fig. 1.1. Data on Pump Speed vs Flow Rate Obtained During Water Test of ARE Fuel System. Sl ORNL~LR--BWG 3088 100 s T T g s [ ’., ° L B0 e e — e e ;/,’,,,,, - ¥ /' o 1 * L<1IJ a0 L "/ . b i v Ll 1 o » Ty &0 e e o g L Ld '/‘ & 1 o« 20 e - fi-;.i"’("""” """""""" | [ eeeeeemeenedooes I R U O A S B O 10 20 30 40 50 50 [49] 80 FLOW RATE (gpm) Fig. 1.2, Data on Pressure Head from the Pump Suction to the Reactor Inlet vs the Flow Rate During Water Test of ARE Fuel System. By using a calibrated orifice installed tempo- rarily in the sodium bypass line around the reactor, it was possible to measure the flow rate while operating the sodium system with water. The pump speed was recorded and alsc the pressure head from the pump suction to the reactor outlet. The system characteristics with water, shown in Figs. 1.3 and 1.4, should be valid while operating with sodium as the fluid. During operation with sodium, the flow rate in each of the two loops will be measured with an electromagnetic flowmeter. The sum of the two flow rates should equal the total flow as obtained from the pump speed. PERIOD ENDING SEPTEMBER 10, 1954 g =ED {rpm) PUMP Sk 0 20 40 &0 2O tC0 120 140 160 180 FLOW RATE (gpm) Fig. 1.3. Data on Pump Speed vs Flow Rate Obtained During Water Test of ARE Sodium System, CRNL -LLR -DWG 309 e T S L 4 8 r T T T T T T T T T T T T T "T = - £ 1 e 2 | l | T | w 24 S s £ | . T 7 i 8 ?,,,,, ‘.} e ”,,,,,,,,;“/t FUU i . — ,,w‘ LT J 0L ‘ e L_ ......................... N o 20 40) 60 80 100 120 140 160 1380 FLOW RATE {gpm) Fig., 1.4. Data on Pressure Head from the Pbmp Suction to the Reactor Outlet vs the Flow Rate During Water Test of the ARE Sodium System, PREPARATIONS FOR LOADING THE ARE G. J. Nessle Mafericls Chemistry Division All the preliminary preparations for loading the barren and enriched fluoride mixtures into the ARE have been completed. The necessary control panels have been assembled, and furnaces, heating units, transfer lines, and various complementary equipment are now ready for installation. The exact location of the various operations has been determined.. However, because of the large amount of construction work which has been carried out in il ANP QUARTERLY PROGRESS REPORT the ARE building around these locations, none of the parts necessary for the loading and sampling operations have yet been installed. Testing of the equipment ofter installation and before the filling operation should require approximately ore week. It is anticipated that all personnel to be invelved with the loading and sampling operation will be made thoroughly familiar with the equipment and its operation during this testing period. FUEL-CONCENTRATE INJECTION NOZ/LE W, G. Cobb W. R. Huntley Aircraft Reactor Engineering Division Tests were made on a resistance-heated fuel- concentrate nozzle to dectermine the effectiveness of the heating in the region where the nozzle will enter the relatively cold pump tank cover plate, These tests showed that a temperature of 1300°F could be maintained in the injsction tube by using about 140 amp from o high-current transformer. This nozzle test was considered satisfactory to meet the operating reguirements for the ARE enrichment system, injection ARE UNLOADING EXPERIMENT J. Y. Estabrook Aircroft Reactor Engineering Division The mockup of the ARE unloading apparatus was tested twice. In the first test the storage tank was filled with 750 b of NaZiF,, which was successfully unloaded in 90-1b increments into aluminum cans. In two cases, however, the bottoms of the aluminum cons melted and the fluoride mixture {ecked out. Modifications made in the apparatus before the second test included improvements in the pouring indicators, and the pressure In addition, small omounts spout, the level medasuring equipment. of the fluoride mixture were frozen in the bottoms of the aluminum cans to minimize the possibility of melting these containers. For the second test the apporatus was filled with 500 1b of NaF.-ZrF, (50-50 mole %), and this material was fransferred at 1200 to 1250°F to the aluminum cans in 70~ to %0:1b increments without difficulty, Attempis to transfer at 1100 to 1150°F, however, led to occasional freezing at the pouring spout and indicated that unless much more reliable temperoture control could be effected, operation at this low temperature could hardly be considered to be feasible. 12 ARE FUEL RECOVERY AND REFPROCESSING D. E. Ferguson G. L. Cathers M. R. Bennett Chemical Technology Division A fluoride-volatility method of processing ARE fuel, hased on elemental fluorination of the molten Naf-ZrF ,-UF,, has been investigated because of its attractiveness from the stondpoint of cost, inventory of fissionable material, woste disposal, In preliminary tests2e3 and safety of operation. than 99.95% of the uranium tetrafluoride contained in the fuel was converted to the volatile hexafluoride with o gross beta decontamination factor of 100 to 300. Resublimation of this uranium hexafluoride resulted in an over-all decontamination factor of 4000 to 5000, excess fluorine UF ( is soluble in molten NaF-ZrF , the resublimed product con be dissolved, probably more Since in the absence of . . 4 according fo the reaction” UF ¢ + NaF > NaF.UF ~—> NaFUF , + F, The fuel from an aircraft reactor would be re- processed in three steps: recovery of the uranium from the used fuel by fluorination, separation of the uranium hexafluoride from excess fluorine by trapping the uranium salt in a cold trap and re- subliming the UF [ from it, and fabrication of a new reactor fused salt by absorbing the partially decon- taminated UF ( in molten NaFF-ZrF . Since uranium- zirconium and other alloys can be dissolved at reasonoble rates by hydrofluorination in a fused salt of the MaF-ZrF ,-KF type,> this also provides a msthod of processing fuels. A decertaminaiion foctor of 4000 to 5000 is sufficient if the fue! is to be recycled directly heterogeneous reactor to a fused salt reactor, but, if the uranium is to be returned to o diffusion plant or to a heter- ogeneous reacior, more decontamination is needed. In further experimenial werk, o decontamination factor of approximately 20,000 was obtained by scrubbing the fission-product fluorides from the 2G. I. Cathers, Recovery and Decontamination of Uranium from Fused Fluoride Fuels by Fluorination, ORNL-1709 (June 9, 1954), 3¢, N. Browder, G. l. Cathers, D. E. Ferguson, and E. O. Numni, ANP Quar, Prog. Rep. Mar. 10, 1954, ORNL-1692, p 42, 4H. Martin, A. Albers, and H, V. Dust, Z anorg, w allgem, Chem. 265, 128 (1951), SR. . Leuze and C. E. Schilling, personal com- municaticn. uranium hexafluoride with a molten NaF-ZrF | salt bath (at 650°C) in which the UF, is insoluble in the presence of excess fluorine. The uranium loss in the scrubbing step was less than 0.001%. Frac- tional distillation of the UF from this step, which would be performed in order to obtain complete decontamination, could be carried out in only lightly shieided equipment. Two experiments on scrubbing UF , with molten NaF-ZrF, were carried out. In one, an acetone-dry ice cold trap was used between the fluorination and scrub steps; the uranium hexafluoride was cone densed {and thus separated from excess fluorine) in the cold trap and was then resublimed from it. The uranium was recovered from the scrub salt by TABLE L1 PERIOD ENDING SEPTEMBER 10, 1954 refluorination, In the second experiment no cold trap was used, but the decontamination obtained was the same (Table 1.1). it is thought that resublimation of the salt-scrubbed product would achieve additional decontamination. Synethetic ARE fuel was prepared for the two runs by dissolving a 224-day-irradiated, 30-day- cooled, 6-g miniature vranium sfug in NaF-ZrF (56-44 mole %) at 650°C by passing HF through the molten bath. The final NaF-ZrF (-UF , mixture had a compesition of 54.0-42.4-3.6 mole % and an activity of 5 x 10° beta counts per minute per milligram of uranium. The dissolution and subse- quent fluorination were carried out in the same nickel reactor, In each run the scrub salt was DECONTAMINATION OBTAINED BY MOLTEN SALT SCRUBBING OF UF, IN THE FLUORIDE VOLATILITY PROCESS Synthetic ARE fuel; beta activity of 3 X ]06 counts per minute per milli- gram of uranium; fivorinated at 650°C with elemental fluorine gas at flow rate of 66 ml/min Run A; Fluorination period of 3 hy; UF6 plus excess fluorine pussed di- rectly to scrub salt from the flusrination step Run B: Fluorination period of 5hr; UF& condensed in cold trap andresub- limed from it prior to scrubking and recovered from scrub by re- flucrinagtion Beta Decontamination Factors Activity o RwmA . RumB Scrub Qver-all Scrub Cverall [ Gross 2 % 104 2 % 10% Ru 6 750 1.4 430 Zr 3 >7 x 10t 9 1 x 10° Nb 4 700 8 2 % 10 TRE 90 8 % 108 3 2 x 10¢ Cs 130 >3 x 10 S¢ 20 >3 x 109 Bq >8 x 10° Uranium Losses (% of Initial Chorge) In fluorinatien salt 1.6 3 x 10"3 In resublimation trap Q.74 In scrub salt 0,23 i3 *Calculated onthe gross Ru, Zr, Mb, and TRE activity rather then on total activity, which included much due to 5.13]- This does not materially offect comparison with previous gress beta decontaminatien data, 13 ANP QUARTERLY PROGRESS REPORT 67 g of an NaF-ZrF, mixture of the same compo- sition as that used in preparing the initial ARE fuel. Both fluorination and scrub operations were carried out at 650°C, The scrub decontamination factors Table 1.1 were calculated on the basis of the activity remaining in the scrub salt, The gross beta decontamination factor of 20,000 obtained in both runs is about fivefold better than the 4000 to 5000 obtained previously. The decontamination factors for individual activities uncertain because of the high amount of the product from the short-cooled slugs. The conversion of UF, to UF, in Run A was relatively poor, with 1.6% remaining in the fluori- nation salt. This was due to a fluorination period of only 3 hr. A longer fluorination period of 5 hr lowered the loss to 0.003%. The same time was used in refluorinating the scrub salt of Run B to recover 98.5 wt % of the initial uranium, Since less than 10~ 3% of the uranium passed through the scrub bath in the absence of fluorine, quantitative reconversion of UF , to UF , by the reaction UF6 + NaF —> NaF .UF ( ——> NaF.UFA +F, given in are somewhat |]3] in is indicated. ARE PUMPS Fabrication and Testing W. G. Cobb A, G, Grindell Aircraft Reactor Engineering Division Five of the six rotary elements fabricated for ARE pumps have passed all acceptance tests. The sixth rotary element is being used in the hot shokedown test stand to test welded-fabricated Inconel impellers and for other miscellaneous testing. Unsuccessful attempts were made by ORNL and by an outside vendor to lap the bronze bearing wear surfaces of the upper oil seals for the shaft to an acceptable flatness (three wave bands of helium light) with a water soluble scouring agent, Bon-Ami. However, acceptable flatness was attained with conventional lapping compounds. Four ARE pumps with welded-fabricated impellers have been installed in the ARE. used were four of the six welded-fabricated im- pellers that hove passed all acceptance tests. One of the impellers is being used in the pump in the reactor system component test loop at K-25 and one is being held as a spare. The impellers 14 Seal-Gas Pressure-Balancing System W. G. Cobb Aircraft Reactor Engineering Division A seal-gaos pressure-balancing system was de- veloped and has been installed in the ARE pumps. This system automatically maintains agaspressure in the lubricant chamber around the lower me- chanical seal that is slightly higher (]/2 to 2 psi) than the pressure in the pump tank volume in order to impede the outward leakage of vapors into the lubricating-oil system. The equipment involved consists of a differential supply valve, a differ- ential vent valve, and the requisite pipe con- nections. In addition to providing a master-slave relationship for maintaining the pressure differ- ential across the seal, the system has two charde- teristics which are advantageous to ARE pump operation. The first advantage is that it will provide automatic bleed-off of excessive oil system pressures created by the gas which will be evolved from irradiation of the lubricating oil. The second advantage is that it will maintain a greater (1’2- to 2-psi) pressure across the lower seal even when subatmospheric pressures are introduced to the pump tank, for example, when the ARE fuel system is being vacuum filled. Radiation Damage to Pump Drives W. G. Cobb A. G. Grindell Aircraft Reactor Engineering Division Some alternatives to the usual V-belt pump- power transmission system were investigated in an attempt to find a system that would not be affected by radiation. The alternatives studied in- cluded direct in-line drives, angle-gear-box drives, silent-chain drives, and improved V-belt drives. The use of direct drives would have required modification of the concrete shielding above the pump pits, and the use of the angle-gear-box drive would have required development of suitable lubri- cant seals. The silent-chain drive tested did not have the required mechanical reliability.® Two improved V-belts, a Browning B-105 Super- grip belt (rayon cords in a neoprene matrix) and a Browning B-103 Grip belt (cotton cords in a rubber matrix), were irradiated, under operating tension, through a short length of the belt by a Co%? source to total doses of 10% and 107 rep, respectively. 6A. G. Grindell, Morse Silent Chain Drives on ARE Pump Hot Shakedown Test Stand, ORNL CF-54-7-166 (Juty 16, 1954), After irradiation each belt was used in the cold shakedown test stand to drive a pump rotary ele- ment at 1600 rpm with no pump load. After 90 hr of operation, the B-105 belt showed no visible damage and was returned to ARE personnel. After 75 hr of operation, the B-103 belt showed no visible damage and was transferred to the pump in the reactor component test loop pump to transmit 3 hp at 800 rpm. This belt, which has now been in service for more than 900 hr, has increased in length approximately 1 in. Failure of the belt does not appear to be imminent, and pump operation has not been affected. An idler pulley is used to maintain tension, On the basis of these tests, it was decided to use the cotton-corded, rubber-matrix V-belts on the fuel and the sodium pumps in the ARE. As an additional sofeguard, the belt on the fuel pump drive will be shielded with lead. Radiation Damage to Shaft Seal W, G. Cobb J. M, Trummel Aircraft Reactor Engineering Division W. W. Parkinson Solid State Division The rotary subassembly of the ARE pump con- tains six Buna-N O-ring static seals which will receive radiation during reactor operation. To get information on the expected life of these seals, an O-ring sealing arrangement similar to two of the pump seals was constructed and was placed in the LITR flux. The test assembly included six radial O-ring seals and three gasket O-ring seals. One side of the seals was supplied with oil at 75-psig pressure, and the other side was vented to the reactor off-gas system through bubblers and a catch tank so that it was possible to detect any appreciable leakage by the seals. Since the seals PERIOD ENDING SEPTEMBER 10, 1954 were vented in four groups, it wos possible to get some information on variations in seal life. Differ- ences in the following might produce such vari- ations: (1) flux density over test assembly, (2) twisting and rolling of the rings during assembly for testing, (3) the O-rings, and (4) the machined parts. The gomma radiation intensity at the test lo- cation in the LITR was estimated by extrapolation from data taken on the Bulk Shielding Facility.” The neutron flux was negligible. The resulis are reported in Table 1.2, which gives time of irradi- ation to first observed leakage and the total esti- mated dose to first leakage. Examination of the O-rings aofter irradiation revealed substantial hardening. No appreciable elastic recovery of the rings occurred upon removal from the container. The O-rings to be used in the ARE will not be in sohigh a flux as that used for these experiments, and they will be under much lower pressures, Since it is not essential in the ARE application that the O-rings retain their resiliency, the resuits of these experiments do not have immediate signifi- cance. It is evident from these tests, however, that O-rings cannot be used successfully in future high-power reactor applications. Motor Test W. G. Cobb A. G. Grindell Aircraft Reactor Engineering Division A test of an ARE-type d-c motor operating in dry helium at 130 to 140°F was: terminated after 4000 hr of uneventful operation. Examination of the silver-impregnated graphite brushes used during the test indicated that total wear was limited to "H. E. Hungerford, The Skyshine Experiments at the ?gusli) Shielding Reactor, ORNL-1611, p 24 (July 2, TABLE 1.2, RESULTS OF RADIATION DAMAGE TESTS OF RUBBER O-RING SEALS Time to Leakage Dose to Leakage Two radial ring sedals on bottom piston Two radial ring seals on middle piston One radial ring on top piston Three gasket seal rings Uj) {rep) 128 5 x 168 167 6.7 % 108 259 1 x 107 259 1 % 107 15 ANP QUARTERLY PROGRESS REPORT approximately ]/“5 in. It is concluded that no problem will be encountered in the use of similarly equipped motors in the ARE, REACTOR SYSTEM COMPONENT TEST LOOP Cperation of Loop W. G, Cobb A. G. Grindell Aircraft Reactor Engineering Division D. B. Trauger G. A. Kuipers Technical Division, K-25 Operation of the reactor system component test loop® at K-25 was temporarily halted after 1047 hr of operation to remove one ARE core hairpin tube for metallographic examination, Shutdown, tube removal, tube replacement, and re-startup were accomplished under a complete helium blanket to prevent air from contaminating the fluoride mixture or the inside of the Inconel system. An irradiated V-belt that was installed, as described above, after 875 hr of system operation has operated for more than 900 hr of the approximately 1800 hr of system operation. Plugging of gas line con- nections to the pump tank by ZrF ,-vapor condensate has occurred periodically. In some instances the plugs have been removed by heating the gas line involved and blowing the vaporized ZrF, back into the pump tank. No completely satisfactory solution to this problem has been found to date; the wvse of o lorge-capacity vapor trap is being studied as a possible solution. 8H. W. Savage et al., ANP Quar. Prog, Rep. [fune 10, 1954, ORNL-1729, p 14, 16 Exominotion of Hairpin Tube G. M. Adamson R. S. Crouse Metallurgy Division Metallographic examination of the hairpin tube showed normal subsurface void attack. The maxi- mum penetration was to a depth of 5 mils. There- fore it appears that corrosion from impurities will not be a serious problem in the ARE. Since this loop was not precleaned, the corrosive conditions should have been worse than those in the ARE. No information on the amount of corrosion that may be expected from mass transfer could be obtained because the loop is being operated isc- thermally, REACTOR PHYSICS ¥W. K. Ergen M. E. LaVerne R. R. Coveyou C. B. Mills Aircraft Reactor Engineering Division R. Bate C. S. Burtnette United States Air Force The Eyewash code, developed by the ORNL Mathematics Panel for use on the UNIVAC, yielded the following results for the ARE: critical mass, 30 Ib of U235; U235 required to yield the 4% excess reactivity needed for operation, 35 Ib; k& . for 40 1b of U%3%, 1.07; critical mass without controls, safety, or instrument apparatus, 17 Ib (this empha- sizes the high cost of these materials); AR/(AM/M)= 0.232; temperature coefficient applicable to slow temperature changes and due to expansion of the whole reactor and to thermal base effects, ~4.3 x 1073/°F; reactivity value of U233 in tube bends at reactor ends, 0.006'in & _,.. PERIOD ENDING SEPTEMBER 10, 1954 2. REFLECTOR-MODERATED REACTOR A. P. Fraas Aircraft Reactor Engineering Division Four years of work on the ANP Project ot ORNL. have led to the belief that the circulating-fluoride- fuel reactor with reflector moderation and a spheri- cal-shefl heat exchanger can be developed into an aircraft power plant of exceptionally high per- formance.! This view has been supported ‘in recent months by the results of studies by USAF contractors. - The question has now become how best to bridge the gap between the ARE and a prototype aircraft power plant. It was proposed about a year ago that a 50-Mw reactor be built so that the feasibility of con- structing and operating a circulating-fuel reflector- moderated reactor could be investigated and its performance characteristics, particularly with ref- erence to control, shielding, heat transfer, and fluid flow, could be determined. Preliminary estimates indicated that a power of at least 25 Mw would be required to disclose the more important control characteristics. Other studies indicated that an output of 50 Mw would be required for the lowest powered aircraft likely to be tactically useful (for radar picket. ships, patrol bombers, etc.), while an output of 150 to 200 Mw would be required (with chemical-fuel augmentation)} to power a strategic bomber. Preliminary reactor test unit designs and cost estimates indicated that the time and cost involved would be roughly propertional to the rated output of the reactor (largely because the size and cost of the heat exchangers, pumps, and heat- dump equipment vary directly with reactor power). After much analysis and discussion it was decided that 60 Mw represented a good compromise for the power rating -and that such a reactor, to be called the Circulating-Fuel Reactor Experiment (CFRE), should be built by ORNL, with the aid of the Pratt & Whitney Aircraft Division. This test unit is to embody the basic features and proportions that have made the circulating-fuel reactor at- tractive for aircraft use, but the details of the pumps, radiators, and cuxiliary equipment will not have to meet aircraft requirements of size, weight, etc. An operating life of 500 hr, of which a sub- stantial portion should be at or near 60 Mw, was TA. P. Fraas and A. W. Savelainen, ORNL Aircruft Nuclear Power Plant Designs, ORNL-1721 (in press). considered to be a desirable goal. The information gained from this project should serve to provide a sound basis for the design of the full-scale aircraft reactor. As has been the case with the ARE, the CFRE will require a major supporting effort on the part of various. groups in the ORNL organization to obtain much essential information not yet avail- able. The program on fuel chemistry, for example, will continue along the lines it has followed. The results of research on both the basic chemistry and the in-pile and out-of-pile corrosion tests are obviously vital to the project. Metallurgical re- search on promising alloys, welding and brazing techniques, strength properties at various tempera- tures and in various ambients, and related work will help greatly to increase the reliability, and possibly the operating temperature, of the system. Tests under way ot the Tower Shielding Facility coupled with further Lid Tank Shielding Facility tests will provide a more complete basis for the aircraft shield design. Multigroup calculations coupled with critical experiments should provide additional information on the static physics of the reactor. Much test work in experimental engi- neering will be required to validate key features of the design. DESIGN OF THE CFRE A. P. Fraas R. W. Bussard Aircraft Reactor Engineering Division The design of the CFRE is still preliminary, but some tentative data can be presented at ‘this As presently conceived, the CFRE will include a reactor, heat-exchanger, pressure-sheli, and shield assembly, as described in previcus reports.?*3 A quasi-unit shield structurally and functionally similar to one suitable for an aircraft will probably be used,? although the structure will stage. 2A. P. Fraoas, ANP Quar. Prog. Rep. Mar. 106, 1953, ORNL-1515,.p 61. 3R. W. Bussard and A. P. Fraas, ANP Quur. Prog. Rep. Dec. 10, 1953, ORNL-1649, p 31. 1g. p. Blizard and H. Goldstein {eds), Report of the %353) Summer Shielding Session, ORMNML-1575 (June 11, 54). ' 17 ANP QUARTERLY PROGRESS REPORT be simplified if simplification proves to be ex- pedient. l arge blowers coupled to banks of NaK- to-air radiators will be used to provide heat dumps that will be simpler, more reliable, and easier to control than turbojet engines. While conventional heat exchangers like those of the ARE might be used, experience in fobricating and testing tube- and-plate finned radiators® indicates that aircraft- type radiators should cost little more and that they will make o much neater, more compact instal- lation. Further, they will effect a manyfold re- duction in the NaK heoldup in the system and will give a thermal inertia and a NaK transit time es- sentially the same as those for a full-scale aircraft power plant. The fabricating and operating ex- perience obtained should also prove to be quite valuable, |t appears that d-c motors will be the most practical means for driving the various pumps in the system, but the blowers will probably be driven by a-c motors, with control being effected by the use of air bypass valves and/or shutters. Key data for the CFRE are presented in Tables 2.1 and 2.2, and a flow sheet for the system is given in Fig. 2.1. The major components are shown approximately to scale on the flow sheet, although their arrangement relative to each other is purely schematic, The more important plumbing and wiring connections expected to be made through the reactor chomber wall and to quxiliary equipment are listed in Tgbles 2.3 and 2.4, CFRE COMPONENT DEVELOPMENT PROJECTS The projected design effort for the CFRE has been tied closely to o comprehensive series of component development projects. A list of these projects as currently conceived is given in Table 2.5. 1t is fully expected that the need for ad- ditional projects will develop as the program evolves, but those listed should provide the most important information required. The first project was completed recently, and the results are presented below; work on some of the other projects is well under way. The test of beryllium in contact with sodium and Inconel under thermal stress was given high priority be- cause the results were needed in the determination of the amount of poison to be expected in the reflector. With the amount of poison in the re- SW. S. Farmer et al., Preliminary Oesign and Per- formanc)e of Sodium-to-Air Radiators, ORNL-1509 (Aug. 3, 1953). 18 flector known, the reactor critical mass, the core size, the fuel concentration, etc. could be de- termined. In the second test, which is well under way, the hydrodynamics of the pump-volute, ple- num-chamber, and core system are being studied. Work on the same basic problem is also being done by the Heat Transfer and Physical Properties Group and by the Pratt & Whitney Aircraft Di- vision. It is hoped that the different approaches of the three groups will lead to a thoroughly sound solution. The objective of the third project is to produce a plastic model of a pump and header-tank arrangement that will remove xenon from the fuel, Data on the solubility of xenon in the fuel and on the rate of production of xenon in the fuel are being obtained by the Fuel Chemistry and Radi- ation Damage groups; the data thus obtained will provide a measure of the efficiency of xenon re- moval by the pump, An in-pile test of a full-scale pump may be required eventually. Xenon removal by the pump is very important, from the control standpoint; in addition, if it can be effected as planned, the bulk of the other volatile fission products will also be removed. This will dispose of a major potential hazard in the event of a reactor failure and should simplify the reactor installation design problem. The information given in Table 2.5 for most of the other component development projects is largely self-explanatory. To expedite the work, some of the projects can be undertaken by other organizations; for example, another organization might undertake the entire job of designing, de- veloping, and fabricating the fill-and-drain system, the fuel addition equipment, the control rod drive mechanism, or the sample heat exchanger tube bundle. The last project listed, the Zero Power Unit (ZPU), deserves special mention. Experience with the ARE has demonstrated that many problems will arise in the fabrication of the CFRE and that there will be o number of doubtful items, especially welds, that could hardly be trusted in a high-power reactor. This seems particularly true of the re- actor core, heat exchanger, and pressure shell assembly, If it is accepted that the first attempt at fabricating the uwnit will leave much to be de- sired, planning to use it as a hot mockup seems to be more sensible than spending much time and money trying to rework and patch it. It can be endurance-tested ot temperature without developing PERIOD ENDING SEPTEMBER 10, 1954 TABLE 2.1. TENTATIVE SYSTEM DATA FOR THE tFRE FLUID CI‘RCUITS Cirevit General Fuel ‘Na NaK Fuel-to-NaK heat exchanger Temperature drop (or rise), OF Pressure drop, psi Flow rate, cfs : Velocifyifhmugh tube mairix; fps Reynolds number ; Over-oll heat transfer coefficient, Btu/hr-ft2- °F Fuel-to-NaK temperature difference, °F Moderator cooling system Temperature drop (or rise) in heat exchanger, CF Sodium-NakK temperature difference, °F Pressure drop in heat exchanger, psi F’ressureg drop in reflector, psi Pressure drop in island, psi Pressure drop in pressure shell, psi Temperature rise in reflector, °F Temperature rise in island, °F Temperoture rise in pressure shell, °F Power generated in reflecfor,I kw Power generated in island, kw Power generated in pressure shell, kw Flow rate through reflector, cfs Flow rate through island and pressure shell, cfs NaK cooling system Airflow through NaK radiators, cfm Radiator air pressure drop, in. H20 Blower power required, hp (total) Blower power required, hp (per fan} Total radiator inlet face ared, 1% Shield cooling system . Power generated in lead layer, kw Power generated in water layer, kw Pump dato, Head, ft Fliow, gpm Specific speed Suction épecific speed impeller speed, rpm Drive, hp Total fluid volumes, ft° 400 f 400 35 | 40 2.7 | 12.6 8.0 = 36.4 4,600 3,150 95 100 100 43 36 32 100 72 28 2,040 600 210 1.35 0.53 150,000 24 720 180 72 132 <4 50 250 280 600 430 2,600 3,750 : : 13,200 2,700 25 16 200 4.3 1.0 16 19 ANP QUARTERLY PROGRESS REPORT TABLE 2.2, REFLECTOR-MODERATED REACTOR POWER PLANT SYSTEM DATA 200- Mw Full-5cale 60-Mw CFRE Aircraft NaK in radiator cores, b Header drums 128 40 Radiator tubes 128 40 Outlet lines 800 240 Inlet lines 200 270 Intermediate heat exchanger 450 135 NaoK flowing in circuits, 1b 2300 680 NaK in header tanks, |b 500 150 Total NaK in complete power plant, b 2800 840 NaK flow rate through radiator circuits (for a 400°F AT), cfs 40 12.6 NaK transit time through system, sec 1.25 1.25 NaK transit time through intermediate heat exchanger, sec 0.23 0.23 NaoK transit time through radiator tubes, sec 0.067 0.067 Fuel in core circuit (flowing), #3 6 3.9 Fuel in header tank, 3 1 0.4 Fuel in core, f13 1.8 1.8 Fuel flow rote, cfs 12.0 2.7 Fuel circuit transit time, sec 0.5 0.7 Therma! capacities of power plant, Btu/°F Radiator cores 600 180 Linres and pumps 250 75 NaK (flowing) 600 180 Intermediate heat exchanger 250 75 Total for NaK systems 1700 510 Fuel 300 210 Total for power plant, Bru/°F 2000 720 nuclear power. Even if a leak should develop in one of the welds, there would be no serious con- tamination problem. Upon completion of the en- durance test, it should not be difficult to make a thorough inspection of the parts —~ something that could hardly be done after sustained operation at high nuclear power. At first thought, the ZPU might appear to delay the program, but closer inspection indicates that it may actually save a substantial amount of time, because efforts to get it into operation can proceed at full speed without the inevitable hazards, re- 20 inhibitions, and hesitations that would accompany high nuclear power operation. Work on the CFRE can proceed at o more deliberate pace, and the mistakes disclosed by work on the ZPU can be avoided. The degree to which the ZPU will be a mockup of the CFRE will depend upon the problems that develop in the course of the component test work. It is hoped that a 10-Mw gas-fired furnace can be used, in place of one of the two CFRE heat dumps, in making some heat transfer tests on the fluoride-to-NaK heat ex- changer and in obtaining a temperature gradient straints, o - ~ \F 4 ! TN X PREHEAT RLANES PREFIEAT RURNEN MODRRATOR BOOO cfm 9:0WER CRNL LR~ 2G 309 T TN : : b : ! : Vo d ] ! } : ) ! ' } - E * o PREFFEAT DOORS — -~ { WMAIN Nok-TO- AR § REDIATOR 1z Tl 75000 ofm 2 ok Fuwe | RN <7 { g NaK PUMP ] N i b, SEWER WATER Mo | \\ / | ' Y ! L L LUBRICATION - G Ul:fi\\((:f?b%l\ GiL , MalN NG - OREREAT ELURNER BURNER -~ GAS ~ PREHDAT FoElL-T0-Nak K S S HEAT EXCHANGER ] . ! ey : \ / N { i \, i T ) ] SEWER WATER MAIN Fe VENT 75000 cfm HLOWER FLIWER He VENT N\ SR AND DRAIN TANK - FUZL i ] \____,//“\ Pl AND DRAIN TANK - o //\‘ 4 /1 [ - u Fig. 2.1. CFRE Flow Sheet, PS6L 0L ¥39WIIdIS ONIGNT QOyId ANP QUARTERLY PROGRESS REPORT TABLE 2.3. WIRING, PIPING, AND TUBING CONNECTIONS THROUGH REACTOR CHAMBER WALL Thermocouples Duplicate thermocouples to oll pipes leaving shield 52 Duplicate thermocouples to main pipes entering shield 8 Single thermocouples to oil inlet and outlet lines to pumps 8 Na and fuel temperatures 12 Pressure shell remperatures 12 Gamma-shield shell temperatures 12 Shield-water temperatures 12 Fuel and Na heut-dump system temperatures 24 Pump drive motor temperatures 8 Control rod drive temperatures 4 Miscellaneous support structure and leak warning temperctures 12 Total 164 Miscellaneous Instrument Wires Muclear instruments 20 Pump speed 8 Power Wiring Pump drive motors (4) 12 Control rod actuatar (1) 3 L.ights and receptacles 2 Fuel addition machine 1 Total 18 Tubing Ma and fuel dump valve actuators™ (]{4 in. OD) 10 Expansion-tank pressure®* (]/4 in. OD) 2 Liquid-level gages* (]{1 in, OD) 2 Main NaK lines (5 in. QD) 4 Moderator-cooling NaK lines (2 in. OD) 4 Gamma-shield cooling water (1 in. OD) 2 Pump cooling and lubricating eil (]/5 ins OD) 4 Xenon vent ] Atmosphere vent 1 Helium supply 2 T otal 32 *These tubes can be replaced by wires. 22 PERIOD ENDING SEPTEMBER 10, 1954 TABLE 2.4. PRINCIPAL WIRING, PIPING, AND TUBING CONNECTIONS TO AUXILIARY EQUIPMENT Thermocouples Duplicate thermocouples to NaK monifold at outlets of all radiator cores 32 Dupllcrate thermocouples to main NaK pipes erflermg radiator cores 8 Single thermocouples to mlet and outlet lines to oil coolers 6 Rediater air inlet and outlef temperatures : f : 12 Shield_wuter inlet and oui‘lret temperatures . 4 NakK dump system temperatures ' ; 12 NaK pump drive motor temperatures , - ' 4 NaK pump temperatures Miscellaneous support structure and leak warning temperatures ' 12 Total 28 Miscellaneous lnstrument Wires Nucledr instruments _ : 20 NakK flow meters 6 NaK pump speed 4 Qil expansion tank level md:cotors 3 Shleld_wafer expansion tank level indicaters 3 Total 36 Power Wiring NaK pbmp drive motors (2) 6 Oil pump drive motors (3) : 6 Shield:water pump drive motor (1) 2 Main blower drive motors (4) 8 Moderator-cooling system blower drive motor (1) 2 Preheat burner drive motors {3) é Total 30 - Tubing NaK dump-valve actuators (] in. OD) ‘ , ; 10 Na and fuel dump-valve octuators (] in. OD) * : i0 Expansion-tank pressures (/ ine OD) : f 2 NaK liquid-level gages ( Cin. OD) 2 Main NaK lines (5 in. OD) 4 Moderator-cooling NaK lires (2 in. OD) 4 Gamma-shield cooling water (1 in. OD) 2 Pump cooling and lubricating oil (16 in. OD) 7 Xenon vent to buried charcoal bed _ 1 Helium supply to expansicn tanks : 5 Total 37 23 in the ZPU system. While this arrangement would hardly represent normal operation, it would simu- late the one-engine-out condition rather well, It is not yet clear just what will be done, but every effort will be made to get as much information as possible from the ZPU, consistent with the money and the time available. REACTOR PHYSICS W. K. Ergen M. E. LaVerne R. R. Coveyou C. B. Mills Aircraft Reactor Engineering Division R. Bate C. S. Burtnetie United States Air Force The three-group, three-region code for reactor statics calculations® on the ORACLE has been completed. The temperature effects for the CFRE were computed on the UNIVAC for three reactors: a 50-Mw reactor with Incone! coolant tubes in the reflector, a reactor without these tubes, and a reactor in which the !Inconel in the core shells was replaced by columbium clad with 0.010 in. of The temperature coefficient for rapid temperature changes turned out to be about -3.5 x 1073 /°F, and the temperature coefficient for slow effects is —2.6 x 1072 /°F for the heavily poisoned reactor with the Inconel coolant tubes, but it in- creases to —3.3 x 107°/°F and -3.8 x 1073 /°F, respectively, for the reactor without the lnconel coolant tubes and the reactor with the Inconel-clad columbium core shells. Inconel. REACTOR CALCULATIONS M. E. LaVYerne Aircraft Reactor Engineering Division C. S. Burtnette United States Air Force Results of the parametric reactor study done on the UNIVAC ot the AEC Computing Facility in New York have been published,”*® and the Curtiss- Wright Corp. has completed, under contract to the United States Air Force, a series of multigroup, multiregion calculations for the ORNL-ANP Proj- 5w, K. Ergen, ANP Quar. Prog. Rep. June 10, 1954, ORNML-1729, p 32. 7C. S. Burtnette, M. E. LaVerne, and C. B. Mills, Reflector-Moderated-Reactor Design Parameter Study, Part 1, Effects of Reactor Proportions, ORNI. CF-54- 7-5 (to be publishad). BM. E. LaVerne and C. S. Burinette, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 32. 24 ect. A report on the results of the latter calcu- lations is being prepared and will be issued by the Curtiss-Wright Corp. A redetermination of the critical mass for the (CA-19) was found to be necessary because of an error in carbon transport cross-section scaling, a differ- rhombicuboctahedron critical assembly ence in reftlector geometry from that assumed for the original calculations, and an actual Teflon density lower than that used in the calculations. In addition, a small correction {about 1% of critical mass) was made for the presence of the aluminum lattice supporting the reactor and for the aluminum control red sheaths. The eigenvalues of the difference-equation ap- proximation fo the age-diffusion differential equa- tion vary with lattice spacing. Calculations were therefore made with three different spacings in order to evaluate this effect. A foil of 0.004 in. thickness was used in the calculagtions. The resuvlts, corrected as mentioned in the preceding paragraph, are given in the following: Space Interval, Ar Calculated Critical Mass (em) (Ib) 0.457 22.75 0.914 21.47 1.828 20.91 The data are plotted in Fig. 2.2. A Lagrangian extrapolation to a zero lattice spacing gives a critical mass estimate of 24.70 Ib. The assembly became critical at a loading of 24.35 b of U233, It is recognized that the agreement between calculation and experiment may well be fortuitous, since the critical mass is not very sensitive to errors in detail. Further evaluation of agreement (or lack thereof) must await additional experi- mental results. Calculations are now being made for the next critical assemblies, that is, the three- region assemblies consisting of a beryllium island and reflector with a Teflon and uranium-foil fuel annulus. The fuel annulus will be surrounded by core shells; aluminum core shells will be used for one assembly and Inconel for the next. BERYLLIUM THERMAL STRESS TEST R. W. Bussard R, E. MacPherson Aircraft Reactor Engineering Division One of the key questions in the design of the CFRE, as mentioned above, has been that of the TABLE 2.5. SUMMARY OF CFRE DEVELOPMENT PROJECTS o Estimated ject roies Project Title Objective Number of References Status September 1, 1954 Number Man Hours 1 Beryllium Thermal Stress investigation of beryllium under severe therma! stress 2280 ORNL-1515, -1517, and -1721 First test completed; no cracks conditions . appeared in beryllium speci- men 2 Core Hydrodynamics Investigation of flow separation, velocity distri- 3880 ORNL-1515, -1692, -1701; YF-15-11; Memo, Tests under way with water bution, and perverse particle dwell time Fraas to file, B/4/54; Memo, Wislicenus to Briant, 6/19/53 3 Xenon Removal Pump (plastic Development of attitude-insensitive pump and header 3080 ORNL-1649, ORNL CF-54+5-1; Memo, Tests under way with water model) tank geometry to remove xenon and degas fuel Fraas to file, 8/4/54 4 Xenon Removal Pump (hot test Performance and cavitation tests with water, 5200 ORNL-1649; Memo, Froas to file, 8/4/54 Design completed; fabricotion unit) sodium, ond fuel; endurance test under way 5 Model-T Component Test Check of fabricability and endurance test of fuel 8920 Pump, header-tank, and core- {isothermal) pump, plenum chamber, and core shell assembly shell design held pending results of projects Z and 3 6 Model-T Component Test (100-kw Check for flow separation; estimate of full-scale 4400 ORNI_ CF-54-2-35 Same as above heat input to core} boundary layer thickness 7 Second Fuel-to-NaK Heat Exchanger Performance and endurance test at full-scale pres- 9280 ORNL-1215, -1330, -1509, 1729 (p 28); Design of about one half of {4000 kw, 1000-kw input) sures, pressure drops, and temperature differentials CF-54-1-155; Memo, Ahern to Fraas, components completed; heat 6/12/54 exchanger being fabricated 8 Incone! Therma! Stress Investigation of effects of severe thermal stress 2680 Design of test rig nearly cycling on cracking and distortion completed 9 Leak Choracteristics of Tube o\ ® EXPERIMENTAL VALUE (CA—19) E 24 + \VL ‘ : : | | | = N FOIL THICKNESS = 0.004 in. w) \\ | n 23 . ~ - . / | i} i < n = Ja2 | N I S l = a 20 ! | | ‘ } 19 ~E O 02 04 06 08 10 412 14 16 48 2.0 Fig. 2.2. Effect of Space Interval Size on Com- puted Critical Mass. effects of severe thermal stresses in beryllium at temperatures between 1000 and 1300°F.% 1% There has been serious concern as to whether cracking or distortion in the beryllium would prove to be a major problem. Therefore a test on the beryllium block shown in Fig. 2.3 was devised to investigate the effects of power (and hence thermal) cycling. The test loop used is shown in Fig. 2.4. Heat was generated in the beryllium block by simple electrical-resistance heating with a current of about 15,000 amp through the beryllium block. Sodium flowed from an electromagnetic pump to the block inlet header, upward through the lower %R. W. Bussard et al., The Moderator Cooling System for the Reflector-Moderated Reactor, ORNL-1517 (Jan. 22, 1954). 10g, A, Field, Temperature Gradient ond Thermal Stresses in Heat Generating Bodies, ORNL CF-54-5-196 (May 21, 1954). UNCLASSIFIED PHOTO 22390 BERYLLIUM Fig. 2.3. Beryllium Test Block. 27 ANP QUARTERLY PROGRESS REPORT RADIATOR ———-___ ELECTROMAGNETIC FLOWMETER SURGE TANK BYPASS FILTER LUG COOLANT LINES UNCLASSIFIED ORNL-LR-DWG 3093 250-kva TRANSFORMER | I| I L m ' ["~—AIR-COOLED COPPER BUS BAR —_ X N\ X CONTACT LUG L Na INLET HEADER CANNED BERYLLIUM BLOCK Fig. 2.4. Beryllium Thermal Stress Test Apparatus. tube bank to the corresponding holes in the be- ryllium block, through the upper tube back into the outlet header, out to the radiator section, and back to the pump. Banks of tubes were used to duct the sodium to and from the block to minimize the flow of electrical current through the sodium inlet and outlet header tanks. A bypass filter arrangement was provided, as well as bypass cooling flow to the sodium-filled lugs connected to the transformer bus bars. The high-power-density volume heat source coupled with transverse sodium flow through the drilled holes gave a high thermal gradient around the holes. During the course of the test the power density was cycled regularly by alternating the 28 operating conditions as shown in Table 2.6, which also compares the operating conditions for the thermal stress test with those for the CFRE, The test included an initial 100-hr period at a constant high power density followed by power cycling at the rate of one cycle per day. The changes from one power level to the other were, in all cases, performed at a fairly uniform rate in a l-min interval. The test was concluded after 1000 hr of operation, including 36 cycles. The beryllium sample is now being examined by the Metallurgy Division for dimensional stability and evidence of mass transfer, corrosion, or erosion. Visual inspection of the sample after the test indicated that no distortion, cracking, or erosion of the beryllium had taken place. PERIOD ENDING SEPTEMBER 10, 1954 TABLE 2.6, COMPARISON OF OPERATING CONDITIONS FOR BERYLLIUM THERMAL STRESS APPARATUS AND CFRE Beryllium Block CFRE Power level High Low 60 Mw Peak power density in berylliom, w/cm’ 40 14 70 Average power density in baryllium, w/cm® 0 14 1.5 Cooling passage diameter, in. 0.25 0.25 0.25 Cooling passage spacing (minimum), in. 1.5 1.5 1.0 Calculated thermal stress (no plastic flow), psi 108,000 38,000 42,000 Sodium inlet temperature to beryllium, °F 1,040 1,060 - 1,050 ’ Sodium outlet temperature from beryllium, °F 1,100 1,080 1,150 Average sedium temperature through beryllium, O F 1,070 1,070 1,100 Maximum thermocouple reading in beryllium, ©F . 1,200 1,110 Calculated maximum beryllia;m temperature, ° F . 1,280 1,150 | 1,230 Time at power level, hr/doy 8 16 Heating .currenf, amp 14,300 8,450 BERYLLIUM-INCONEL-50DIUM SYSTEMS The compatibility of sodium, beryllium, and In- conel has been studied by using whirligig, seesaw, thermal-convection-loop, and = forced-circulation- loop tests. The results obtained to date indicate that it will be possible to use beryllium unclad in the CFRE if the temperature of the sodium- beryllium-Inconel region (that is, the reflector and the island) is kept below 1200°F. The main effect observed thus for in the tests has been the type of mass transfer usually found when dissimilar metals are used in a system. In this case the mass transfer has been evidenced by the alloying of beryllium with the Inconel walls in regions where the beryllium and Inconel were close to- gether, with only stagnant sodium separating them. Beryllium-lhconel-Sodium Compatibility Tests E. E. Hoffman W. H. Cook C. R. Brooks C. F. Leitten Metallurgy Division Nine tests of the compatibility of beryllium, sodium, and inconel in dynamic systems have been completed (Table 2.7). In the whirligig tests o circular loop of Inconel tubing with a berylllum insert was pdrfiy filled with sodium which was circulated at o velocity of 10 fps. For four of these tests there was no temperature differential, but for one test there was a differential of 140°F. These tests were of 300-hr duration. Two thermal- convection loops of Inconel with beryllium inserts were tested for 1000 hr with sodium circulating at a relatively low velocity (approximately 2 to 10 fpm). In addition, seesaw tests were performed in which beryllium specimens were retained in the hot zones of osciilating Inconel tubes partly filled with sodium which cvrculoted at a velocity of 3 fpm. Macroscopic examination failed to show attack on the Inconel, but two specimens showed de- posits. Black layers were formed in the annular space between the beryllium insert and the en- closing Inconel sleeve in the whirligig and thermal- convection-loop tests. The layer was very ad- herent and made removal of the insert difficult; consequently, the weight change data obtained were not considered as being relmb!e. A picture of this layer from © whirligig ioop operated at 1300°F is shown in Fig. 2.5. Micro- scopic examination showed that the maximum depth of attack on the inside surface of the Inconel did not exceed 0.5 mil. Chemical analysis of the sodium bath reveaied less than 0.04% berylliom in the sodium. In all tests, beryllium was detected 29 ANP QUARTERLY PROGRESS REPORT TABLE 2.7. RESULTS OF BERYLLIUM-SODIUM-INCONEL COMPATIBILITY TESTS Concentration of Test T Time Temperature Be on Surface Q . est IYPE b CE) of Inconel Tube emarks (g/em?) Whirligig 300 1100 (isothermal) 53ta 7.9 Two dark deposits on Inconel analyzed high in heryllivm; macroscopically, beryllium showed fine pitted appearance 300 1200 (isothermal) 0.05 No visible deposits on Inconel; beryllium smooth but discolored 300 1300 (isothermal) 0.07 No visible deposits on lnconel; beryllium showed dark discontinuous deposit covering surface 300 1400 (isothermal) 39.6 to 127.0 No visible deposits on Inconel; beryllium surface pitted and partly covered with small deposits 270 Hot zone, 1200 loose, adherent gray deposit found in cold zone, Cold zone, 1060 analyzed high in sodium; beryllium smooth but discolored Thermal- 1000 Hot zone, 1150 Inconel appeared smooth and showed no visible convection Cold zone, 990 deposits; beryllium smooth but discolored loop 1000 Hot zone, 1300 Inconel appeared smooth and showed no visible Cold zone, 1130 deposits; beryllium specimen smooth and light gray in color Seesaw 300 Hot zone, 1400 Hot zone, 14.0 Beryllium specimen covered with black flaky Cold zone, 710 Middle, 9.9 deposit; surface of Inconel tube discolored Cold zone, 20.6 300 Hot zone, 1100 Hot zone, 0.07 Similar to above on the Inconel tube wall surface. of the Inconel surfaces indicated that the beryllium Cold zone, 740 X-ray analyses Middle, 0.02 Cold zone, 0.01 the inside surface, but attack up to 5 mils was observed on the ocutside surface which was in was present as the metal. In the thermal-convection loops, a deposit was formed between the contacting ends of the be- ryllium insert and the Incone! tube. X-ray anal- yses of these deposits and those found in the whirligig loops showed beryllium, Inconel, and lines very similar to sodium bicarbonate. The beryllium specimens from the seesaw tests were covered with a black, flaky deposit, which has been identified by x-ray anaclysis as beryllium oxide and beryllivim, Again, lines similar to sodium bicarbonate were observed. Microscopic examination of the beryllium speci- men from a whirligig test showed little attack on 30 contact with the relatively stagnant sodium in the 0.005-in. annular space (Fig. 2.6). Mass Transfer Tests in Thermal-Convection Loops G. M. Adamson Metallurgy Division A series of thermal-convection loops is being operated to determine the effect of dissimilar metal transfer in the beryllium-Inconel-sodium These studies are being carried ouf in mass system. Inconel loops with short beryllium inserts in the hot legs; sodium is the circulated fluid, The Figa 2-5- PERIOD ENDING SEPTEMBER 10, 1954 UNCLASSIFIED Y-12811 S INCH Inside Surface of Inconel Sleeve Enclosing Beryllium Insert Exposed in Whirligig Test to Sodium for 300 hr at 1300°F, Etched with glyceria regia. 1000X. TABLE 2.8. EFFECT OF CIRCULATING SODIUM AT YARIOUS TEMPERATURES FOR 500 he IN INCONEL THERMAL-CONVECTION LOOPS WiTH BERYLLIUM SPECIMENS IN THE HOT LEGS Metallographic Examination Temperature N °F) Beryllium Specimen inconel Hot Leg 1500 Outside surface of insert showed holes to a depth Deposit opposite the beryllium of 14 mils; inside surface slightly rough insert 1300 Intergranular penetrations to a depth of 2 mils on Mo attack or deposit outside surface; no attack on inside surface 1100 No layer or attack No attack or deposit 200 No layer or attack No attack or deposit results reported in Table 2.8 are from tests oper- ated for 500 hr. Other loops are now in operation that wifl be run for longer periods. k Photomicrographs of the outsides of the be- ryllium inserts are shown in Fig. 2.7. Black, flaky material was found in the cold legs of the loops operated ot 1100, 1300, and 1500°F and in the hot leg of the loop operated at 1500°F. The deposit material has not yet been identified by spectrographic or diffraction studies. The deposit found on the Inconel opposite the beryllium insert in the loop operated at 1500°F is shown in Fig. 2.8; it has been identified as a nickel-beryllium alloy. 31 ANP QUARTERLY PROGRESS REPORT NCLASSIFIED ¥-12994 Fig. 2.6. Beryllium Specimen After Whirligig Test for 300 br ot 1300°F. (a) Surface exposed to flowing sodium, and the Inconel sleeve. Mass Transfer in Forced-Circulation-Loop Tests L. A, Mann Aircraft Reactor Engineering Division F. A. Anderson'! University of Mississippi A test unit!? for investigating mass transfer of beryllium to Incone! in flowing sodium was oper- ated for 1006 hr under conditions nearly the same as those proposed for the reflector-moderated re- The approximate conditions of operation, with minor fluctuations, were as follows: sodium flow rate, 3 gpm; beryllium temperafure, 1200°F; temperature, 1245°F (sodium entering economizer); minimum flowing sodium temperature, 900°F (re-entering economizer from cooling sec- tion); Reynolds number ot minor diameter (0.25 in.) actor. Maxiimum nSummer Resesarch Participant. 120 A. Mann, ANP Quar. Prog. Rep. Jure 10, 1954, ORNL-1729, p 22. 32 (b} Surface exposed to relatively stagnant sodium in the annular space betwaen the specimen Etched with oxalic acid. 250X, of beryllium insert, 155,000; minimum Reynolds number (at coldest point in economizer annulus), 61,000. Upon termination of the test, the unit was cooled rapidly to room temperature to freeze the sodium. Samples of both the sodium and the Inconel! were then taken from each of 24 locations in the unit. The sodium samples were analyzed chemically for beryllium, und none was found. Metallurgical ex- amination of the beryllium piece and the Inconel samples indicated that the beryllium-lnconel- sodium system should give no trouble from cor- rosion or transfer under the conditions presently proposed for the reflector-moderated re- 1200°F, Reynolds Mmass actor (temperature not over number not over 150,000). The dbove-described test unit was essentially duplicated, and o second test was started with maximum and minimum temperatures in the flowing sodium of 1300°F {at the beryllium) and 1000°F, respectively. A power failure terminated this test B UNCLASSITIED T-583939 ) TRERD §. t-5%az it s UNOLASSIFIED . Y-goat Fig. 2.7. Surfaces of Berylliuminserts in Inconel Thermal-Convection Loops After Exposure to Flowing Sodium for 500 hr at (z) 1100°F, (%) 1300°F, (c) 1500°F. Unetched. 250X. Reduced 36%. ' PERIOD ENDING SEPTEMBER 10, 1954 after 268 hr of operation. Sodium and Inconel samples and the beryllium insert were removed for analysis and examination, but results have not yet been received. A third unit has been built and is intended to operafe for 1000 hr. M onduassierenl - r.s0s2 | WS Fig. 2.8. Layer on Inconel Opposite Beryllium Insert in Thermal-Convection Loop That Circu- lated Sodium at 1500°F for 500 hr. Unetched. 250X. Reduced 36%. : 33 ANP QUARTERLY PROGRESS REPORT 3. EXPERIMENTAL REACTOR ENGINEERING H. W. Savage Aircraft Reactor Engineering Division The emphasis in the engineering work has shified to research and component development for the CFRE and for in-pile loops, since most of the preoperational ARE tests have been completed. Twe types of puinps have been developed for in- pile loop use, The vertical-shaft centrifugal sump pump, which would be installed outside the reoctor shield and thus would require auxiliary shielding, is now being fabricated in sufficient quantity to meet the demands of the in-pile loop program and the forced-circulation corrosion testing program, The small (4-in.-OD)}, air-driven, horizontal-shaft sump pump being developed for insertion in a reactor beam hole was tested with NaF-ZiF, at 1350°F. Some difficulty was encountered with initial priming, but operation was otherwise satis- factory. A new, small pump that has the required small holdup volume is being designed. This pump, which will use o turbine-type impeller, has the advantage that both the inlet and discharge can be at the bottom. Hydraulic drive motors of suitably small dimensions have been found to be satisfoctory drives for these pumps. Additional work has been done on the development of forced-circulation corrosion-testing loops for obtaining information on the corrosion of Inconel in high-velocity turbulent fluoride mixtures with large temperature differentials in the system. Two series of loops are being constructed to meet the following requirements: a Reynolds number of 10,000 with temperature gradients of 100, 200, and 300°F and o temperature grodient of 200°F with Reynolds numbers of 800, 3,000, and 15,000. The maximum fluid temperature is to be 1500°F. A forced-circulation loop is also being developed for testing combinations of structural metals in contact with high-velocity turbulent liquid metals under high temperature differentials, Several burners for use with the proposed gas-furnace heat source for high-temperature reactor mockup tests. Also, a study is under way of the cavitation phe- nomenon associated with operating liquid metal systems at elevated temperatures, high flow rates, exploratory tests were made of gas 'D, F. Salmon, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 19. 34 and high pump speeds. A correlation of fluid-flow- noise intensity with pressure data was noted. The number of stations available for convection loop testing has been increased from 18 to 31 so that many more long-term tests (2000 hr or longer) and intensive tests of special materials can be The basic design of the convection loops has been simplified, and various means of heating the loops and of making operation of them more gutomdatic are being studied. made. IN-PILE LOOP COMPONENT DEVELOPMENT ¥W. B, McDonald Aircratt Reactor Engineering Division An in-pile loop for insertion in the MTR is to be designed, constructed, and operoted as a joint ORNL and Pratt & Whitney Aircraft Division project. The loop is being developed for circu. lating proposed fuel mixtures so that the extent of radiation damage to materials of construction and the effect of radiation on the fuel can be de- termined, Preliminary ORNL work on this project is concerned with the design and development of components of a loop to operate in o horizontal beam hole. Further developmental work was also done on the vertical-shaft centrifugal pump for use with both in-pile and out-of-pile forced-circu- lation loops. Horizontal-Shaft Sump Pump D. F. Salmon J. A, Conlin Aircraft Reactor Engineering Division The air-driven horizontal-shaft sump pump de- scribed previously! was operated with NaF-ZrF , at 1350°F, |t produced a 70-psi head at 6000 rpm and a flow rate of 1.5 gpm. Initial difficulty in priming the pump was solved by momentarily flooding the impel!ler labyrinth. This pump, which was built as a pilot model for checking design principles, hot pump performance, and reliability, has too large a volume holdup for in-pile use and will now be used as a laboratory pump. A new pump has been desighed that has the required small pump holdup volume. For priming, this design incorporates a boffle in the sump tank that will permit remote, momentary flooding of the impeller eye and the labyrinth behind the impeller. A turbine-type impeller is used instead of the conventional centrifugal impeller. it has been experimentally determined, with water, that the turbine type of impeller will prime itself if the inlet and outlet are placed at the bottom of the pump and the fluid is maintained at a level about half way between the bottom of the impeller and the impeller center. The horizontal sump pump principle is used, with the sump having a minimum volume that acts merely as an expansion chamber and a reservoir to replace and catch the fluid leaking past the labyrinth seal. Excessive impeller end clearances are necessary for this type of pump because of the severe temperature gradients present. However, estimated available pump heads (25 ft at 4500 rpm and 1.5 gpm} are greater than the estimated requirements. Hydraulic efficiency of the pump will be low but performance will be reliable. Further work has been done on centrifugally sealed and frozen-fluoride-sealed pumps, and it has been established that the centrifugally sealed pump requires too greata volume of fluid for sealing for it to be considered for in-pile use and that the leakage rate of the frozen-fluoride-sealed pump is excessive. However, the centrifugally sealed pump, which will be useful for other applications, has been modified to include a system that makes the pump self-priming and deaerating. By con- trolling the seal level and using a proper venting __PRESSURE EQUALIZATION LINE L RARGE ~ PUMP HOUSING VENT —- CISCHARGE J ‘“] TN PROBE. . ) 4 s HOLE - -aa_ \ (I e { B2 [0 —T J\{ - - SHAFT \ PERIOD ENDING SEPTEMBER 10, 1954 sequence, the pump can be easily stopped and restarted. A pilot mode! of the pump, similar to the pump previously described,? has been designed that incorporates these features (Fig. 3.1} and an integral sump tank connected by a passage to the seal cavity. In operation, the system would be filled tc the full level. The seal cavity and sump tank would then be pressurized through the gas inlet tube to the predetermined pressure necessary to obtain a positive pump inlet pressure during operation, and the pump would be started. The centrifugal force would cause the liquid in the seal cavity to form a rotating annulus. The annulus of fluid would develop a small pressure and thus pump fluid into the discharge tube, through the loop, and back to the pump. The small bieed hole from the pump discharge to the sump tank would permit deaeration by bypussing aerated fluid from the pump into the sump. The bypassed fluid would be replaced by clear fluid from the seal. When the pump was primed and deaerated, a continuous flow of fluid would pass from the bleed hole to the sump to remove any accumulation of gas. Throughout this operation, fluid entering the loop from the seal to either fill the loop or replace the bleed flow would be replaced by fluid from the sump, and sufficient fluid would thus be in the seal for sealing. There- fore there is a lower limit on the run level in the 2J. A, Conlin, ANP Quar, Prog. Rep., June 10, 1954, ORNL-1729, p 18. UNCLASSIFIED QRMIL~LR-DWG 3094 LGAS INLET Ol OUTLET - ™~ — A % SOl INLET OlL. DRAIN Fig. 3.1. Centrifugally Sealed Sump Pump, 35 ANP QUARTERLY PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 3095 lj?‘s'm e SIR Y 0% _.—ROTARY SEAL UNION _|-SPARK PLUG PROBE | | £ \—‘\N M ’ o i COOLING OIL TO ANNULUS | | PRESSURIZING GAS LINE ‘ e f— ! i i LIQUID LEVEL - - VOLUME -REDUCING . ATTACHMENT (OPTIONAL) — fm INLET | ANTIFLOWBACK SEAL \ ! IMPELLER VANE = I . | Y3,-in HOLE | DRILLED IN SHAFT [ DISCHARGE Fig. 3.2. VYertical-Shaft Centrifugal Sump Pump (Model LFB), 36 sump in that it must be sufficiently high at all times to provide the necessary flow into the seal. To stop the unit, the sump tank would be vented at the same time the pump was stopped, but the gas pressure would be maintained unti! draining had been completed. This would cause a flow of gas along the shaft into the seal cavity and would blow out any fluid that might enter along the shoft. I+ would also maintain a seal cavity pressure somewhat greater than the sump tank pressure and would force the fluid out of the seal into the sump. If this procedure were not followed, the sealing annulus of fluid would, upon collapsing when the pump was stopped, momentarily fill the seal cavity above the shaft level and cause fluid to enter along the shaft, Vertical-Shaft Centrifugal Sump Pump D. R. Ward Aircraft Reactor Engineering Division Developmental work has been continued on the vertical-shaft cenirifugal sump pump (Model LFB) for use with forced-circulation loops, both in-pile and out. This is a vertical-shaft cenirifugal sump pump with a side inlet and a bottom discharge (Fig. 3.2). A Graphitar face seal prevents the pressurizing gas from escaping along the shaft. Circulating spindle oil serves to lubricate this face seal and to carry away heat from the seal region. Typical performance curves for this pump with water are shown in Fig. 3.3. In o recent test the pump was run for 864 hr at 6000 rpm while pumping 1400°F sodium at about 3 gpm. Typical oil leakage past the shaft face seal was under 0.1 cm3/hr. Developmental work during the past eight months has resulted in the following improvements in this pump: the elimination of the tendency for the pump to introduce gas into the liquid stream at flow rates up to 6 gpm; an improved anti-flowback seal to prevent recirculation of the liquid within the pump; an improved means for cooling the face-seal region of the pump; and a reduction of pump volume, which is desirable when the pump is used for in-pile loop tests. The main difficulty, that of gas entrainment, was primarily caused by vortexing and the passage of gas downward along the rotating shaft. The ad- PERIOD ENDING SEPTEMBER 10, 1954 UNCILLASSIFIED ORNL-LR-OWG 3096 48 [ ’_'--...,__‘_. i \*\ ,,,,,,, *\ 6000 rpm T —— ~g — 32 b & \4 —_ S —— o - e e -~ 5000 7 .\J 5 1 rpm 2 e < pa - - e T s e —_ 2 Ay ! \.\ 4000 rpm o f—1 e .;'!.fiv-.....‘;_;i.; ---------- — I ! \’\?N 3000 rpm 8 ..................... — “\+;fi.__ ,,,,,,,,,,,,, 0 0 2 4 8 8 FLow {gpm) Fig. 3.3. Performance Curves for Vertical- Shaft Centrifugal Sump Pump (Model LFB). dition of vertical baffles adjacent to the shaft merely broke up the large vortex into several smaller ones, which were just os objectionable. The problem is solved in the present arrangement by permitting liquid from the high-pressure region of the pump to be bypassed up through a hole in the shaft to an annular space surrounding the shaft. This liquid, in turn, forms a seal and prevents the gas from following its former path. Acceptance tests for these pumps include oper- ating each pump with water to check the pumping characteristics and freedom from gas entrainment and pumping sodium at T1400°F for 72 hr at 6000 rpm with acceptable oil leakage past the face seal. This hot test also serves to clean oxide film from the wetted pump parts, as well as to reveal any mechanical faults that might be present. Five of these pumps are being built, and four similar pumps {Model LFA} are now in use. 37 ANP QUARTERLY PROGRESS REPORT Hydrauvlic Motor Pump Drives 4. A. Conlin Aircraft Reactor Engineering Division Two types of hydraulic motors have been found to be suitable for use os in-pile pump drives. The first, a gear-type motor rated at 1.75 shp at 3500 rpm, was operated at 6000 rpm for 500 hr under a light load (about 0.08 shp) that was obtained by Upon completion of the test the motor was disassembled and was found to be in good condition. The rubber lip shaft seal had worn a slight groove in the shaft, and o carbon-like deposit which had ac- cumulated on the shaft just outside the seal caused the seal to stick slightly after it had been idle for a prolonged period. However, throughout the test, sea! leakage was nregligible. The only other indi- cation of weoar was a slight scoring on the gear sides adjocent to the shaft end of the motor. The second motor, an axial-piston type rated at 3.8 shp at 6060 rpm continuous duty, was run for 100 hr at 6000 rpm. Operation was completely satisfactory, and, since this motor is rated at 6060 rpm, further testing was judged to be un- necessary. The axialepiston motor was chosen for in-pile service, since it is considered to he the more reliable unit. Throughout these tests, hydraulic oil with a viscosity of 155 SSU (seconds Saybolt universal) at 100°F and a viscosity index of 100 It should be noted that during initial testing of the gear-type motor, oil with a viscosity of 290 to 325 SSU at 100°F was used, and, as a result, excessive overheating of the motor occurred to the extent that the shaft turned blue. Despite this, the motor completed the test satisfactorily. coupling the motor to @ small air blower. was used, FORCED-CIRCULATION CORROSION LOOPS Incone! Loops W, C. Tunnell Aircroft Reactor Engineering Division W. K. Stair, Consultant J. F. Bailey, Consultant University of Tennessee The previously described, small, Inconel, forced- circulation loop designed to have a temperature differential of 135°F has been operated for a total of 141 hr with NoF.ZrF UF (50-46-4 mole %) at ain apparent mass flow rate of 400 to 450 Btu/hr with a temperature differential up to 300°F.% The 38 fluid velocity in the loop normally was maintained at about 1.7 fps at a temperature gradient of 165°F. Termination of the test resulted from a short in the motor leads and subsequent rupture of the heated section of the loop. The test facility, shown in Fig. 3.4, is being rebuilt and is approxi- mately 80% complete. The final design conditions are presented in Fig. 3.5. Reports which give a detailed presentation of the design and an analysis of the initial operation are being prepared.?s3 In the design of the loop for electrical-resistance heating, a resistivity value® of 98 gohm-cm was used for Inconel in computing the necessary length of the resistonce-heated tube. However, data obtained during the loop operation indicated that the actual resistivity of the Inconel was about 70 to 80 pohm-cm. To resolve the discrepancy, a duplicate of the heuating section of the loop was filled with NaF-ZrF .UF, (50-46-4 mole %) and alternately heated and potential ond current measurements were made. Sixty-two measurements were made in the tempera- ture range of 800 to 1600°F, and the average resistivity of the Inconel was found to be 75 pohm-cm, cooled while accurate A program is under way for the design, con- struction, and operation of two series of forcad- circulation loops for studying the effect of temper- ature gradient and fluid velocity on the phenomenon of mass transfer in Inconel systems containing fluoride mixtures at elevated temperatures.” The requirements of the two series of loops are a Reynolds number of 10,000 with temperature gradi- ents of 100, 200, and 300°F and o temperature gradient of 200°F ond Reynolds numbers of 800, 3,000, and 15,000. The maximum fluid tempercture is specified as 1500°F, and the Reynolds numbers are to be evaluated at that temperature. The loops are to be consiructed of Inconel tubing. The maximum tube-wall temperature is to be 1700°F, and the surface-to-volume rotio is to be held essentially the same for each loop. 3W, C. Tunnell, W. K. Stair, and J. F. Bailey, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 21, ‘w. K. Stair, The Design of a Small Forced Circu- lation Cosrrosion Loop (to be published). SW. C. Tunnell, Operation Report of the First Small ATV Study Loop (to be publishedf. 6lnterrw.fionol Nickel Co., Properties of Some Metals and Alloys. 7W. D. Manly, High Flow Velocity and figh Temper- ature Gradient Loops, ORNL CF-54-3-193 {Mor. 18, 1954). PERIOD ENDING SEPTEMBER 10, 1954 UNCLASSIFIED FHCOTO 22350 é i Fig. 3.4. Inconel Forced-Circulation Corrosion Loop. 39 ANP QUARTERLY PROGRESS REPORT TEMPERATURE DIFFERENTIAL, 159.2°F REYNOLDS NO. AT 1300°F, 3730 REYNOLDS NO. AT 1459°F, 4540 FLOW, O.B1gpm VELOCITY, 5.29 fps MASS FLOW RATE, 0.361 Ib/sec LENGTH OF HEATING SECTION, 5.43 ft LENGTH OF COOLING SECTION, 5.86 ft ORNL-~LR -DWG 3097 1300°F POWER POWER, 17.6 kw 1300°F TERMINAL CURRENT, 400 amp \ COOLING VOLTAGE, 44 SECTION—n T Fig- 3.5. Temperature Differential of Approximately 160°F, A summary of the design dimensions and expected operating conditions for each of the six loops is presented in Table 3.1, which also indicates the current status of each loop. Three methods of heating were studied: electrical-resistance, gas- fired furnace, and liquid bath. The liquid bath was discounted because it would require pres- surization of the liquid metal an excessively long heat exchanger,and a large investment. A gas- fired furnace was considered, but lack of operating experience and the high heat flux necessary for the loops indicated the use of electrical heating, with which considerable experience has been ob- tained. A gas-fired furnace is being studied as a possible heat source for a large number of cor- rosion loops. Preliminary calculations of the pressure drop through the loops indicate that the 40 HEATING SECTION 0.30-in. OD, 0.025-in. WALL INCONEL TUBING 1459.2°F Configuration and Design Conditions for Inconel Forced-Circulation Loop to Operate at a Model LFA pump® can be used, and thus it will not be necessary to develop a pump. For purposes of preliminary design the following assumptions were made: (1) The specific resis- tivity of molten NaF-ZrF4-UF4 (50-46-4 mole %) is so high that the resistance of the electrically heated Inconel tube may be taken as the parallel circuit resistance.” (2} Heat transfer to the fluoride mixture may be calculated by relations established for water.'® (3) A thermal loss equal 8W. B. McDonald et al., ANP Quar. Prog. Rep. Mar, 16, 1954, ORNL-1692, p 15. 9 ; . A, L. Southern, personal communication. ]OD. F. Salmeon, Turbulent Heat Transfer from a Moliten Fluoride Salt Mixture to NaK in a Double Tube Heat Exchanger, ORNL-1716 {to be issued). PERIOD ENDING SEPTEMBER 10, 1954 TABLE 3.1. PRINCIPAL FEATURES AND STATUS OF FORCED-CIRCULATION CORROSION LOOPS Loop Number 3 4 5 6 Reynolds number at 1500°F 10,000 10,000 10,000 200 3000 15,000 Temperature differential, °F 100 200 300 200 200 200 Power input, kw 27.7 54.2 80.0 6.8 25.5 §6.% Heating section length, ft* 8.08 8.08 8.08 6 22.3 8.08 Cooling section length, ft* 15.64 15.64 15.64 1.4 4 15.64 Heat exchanger tength, ft** 18.75 18.75 18.75 None None 18.75 Pressure drop, psi 35.12 36.54 47.5 G.06 0.57 115.5 Status Detail Detail Construction Detail Preliminory Construction design design and as- design design 20% com- complete complete sembly 90% complete; plete 80% com- complete detail de- plete sign 20% complete *0.5-in-0D, 0.020-in.~wall tubing. **1.0-in.-0D, 0.049-in~~wall tubing used for outer tube of concentric-tube heat exchanger. to 10% of the total heat transfer exists at the heater section and at the heat exchanger. Dissimilar |,__. > g 16 e Lo e . . ] << BARE MINUS CADMIUM- GOVERED ALUMINUM FOIL AGCTIVATION - 1o CADMIUM - COVERED ALUMINUM Q1L ACTIvATION - 2 Bl o AN TRV ~ Nl L 108 = é & L i | 08 & " - - & _ 7 : % 04 2 . - - g CADMIUM FRACTION-. e oz 2 | , 3 : T Q o 0 0 1 2 3 5 6 7 a RADIAL DISTANCE FROM CENTER OF GURVE {in.} Fig. 4.3. Power Distribution of Mid-Plane of Reactor. The paired points on the bare aluminum foil activation curve at 6.4, 6.9, and 7.1 in. were obtained from foils placed on opposite sides of the same ura- nium foil. metals, such as Inconel, nickel, and cadmium. Upon completion of the experiments on this as- sembly, it is planned to construct a larger reactor of the some shape that will consist of three regions, with the beryllium island and reflector separated by the fuel annulus. The diameters of the beryilium islond and the fuel annulus will be about 10.4 and 20 in,, respectively. Initially, the assembly will have no structural materials, such as Inconel core shells, and it will provide a further check of the calculational methods, SUPERCRITICAL-WATER REACTOR J. S, Grudele J. W, Noaks Pratt and Whitney Aircraft Division The Supercritical-Water Reactor (SCWR) critical assembly was described previously* as consisting of an aqueous solution of enriched UC,F, con- 4. L. Zimmérmcn, J. 5. Crudele, and J. W, Nocks, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 45. tained in stainless steel tubes distributed in on organic liquid in a pattern designed 1o give a uni- form radial thermal-neutron flux. The liquid, Furfural (C5H402), which also serves as a neutron reflector on the lateral surface of the cylindrical tube bundle, has a hydrogen density opproximately the same as that of water under the temperature and pressure conditions designed for the SCWR. It simulates the nuclear properties of supercritical water as well as can be determined without further knowledge of the effect of binding energies on diffusion and slowing down lengths, Stainless steel was inserted in the fuel tubes to represent the reactor structure. The loading of the critical assembly was calcu- lated by Pratt & Whitney Aircraft Division person- nel,® by use of a four-group diffusion theory method, to be 15.34 kg of U?3% and 261 kg of stainless steel in o 76.2-cm equilateral right cylinder with a 3Brivate cammunication from Geoarge Chase, Fox Project, Prott & Whitney Aircraft Division. 47 ANP QUARTERLY PROGRESS REPORT side reflector 6.5 cm thick. The calculated radial fuel distribution is plotted in Fig. 4.4 as the smooth curve and is compared to the stepwise distribution achieved when aoll tubes are identically loaded, In the experimental study, a quantity of UO,F, aquecus solution, containing uranium enriched to 93.14% in U?3%, was distributed among the tubes at a U23% concentration of 0.505 g/cm®. The number of tubes required for criticality was meas- sured as a function of the Furfural height as in- creasing quontities of stainless steel were inserted ond as the solution was diluted. In this manner a stepwise opproach was made to a loading of uniform 0.30 \ concentration which would be critical at the de- signed linear dimensions of both fuel solution and Furfural and which would contain the prescribed mass of stainless steel. These conditions have not been achieved because the last dilution was overestimated and a large part of the solution is at a concentration somewhat lower than required for criticality, a deficiency compensated for by locating about 50 tubes of higher concentration fuel in one peripheral section. in Table 4.2. Some neutron-flux measurements have been made by using effectively thin (3 x 107% in.) indium foils The configuration is described S ORNL-LR-DWG 3019 0.28 |— 0.26 ACTOR 0.24 RE - 0.22 0.20 FRACTION OF TOTAL FUEL X 102/UN{T AREA O 042 e S 0AQ fmmrmmrserer e B 00B b b o 006 |- i 0.04 . 0 5 10 15 20 25 RADIUS (cm) Fig. 4.4. Rodial Fuel Distribution in Supercritical-Water Reactor. 48 with and without 0.02-in.-thick cadmium covers, The data obtained along a radius 36.7 cm from the bottom of the core and somewhat removed from the high-solution-density perturbation ore shown in Fig. 4.5. Similar data from a longitudinal traverse located 2.4 c¢m from the cylinder axis are given in Fig. 4.6. Although these results are preliminary, it is believed that the apparent nonuniformity of the radial thermal flux is greater than the experi- A few measures show the U233 1o also decrease from mental uncertainty, radial importance of the center, In one experiment to evaluate the Furfural re- flector, annular sheets of aluminum were inserted adjacent to the wall of the reactor tank, thereby reducing the reflector thickness from 10.0 em to 3.6 cm. The effect of the aluminum, indicated by the critical height of the Furfural, was a slight increase in the reactivity. PERIOD ENDING SEPTEMBER 10, 1954 TABLE 4.2, CONFIGURATION OF SCWR CRITICAL EXPERIMENT Experimental Design Number of tubes™ 215.0 215.0 Height of UO,F, salution, cm 71.2 76:2 Height of Furfural, em 69.5 76.2 Mass of UP33 | g** 11.48 15.34 Thickness of reflector, em 10.0 6.5 Mass of stainless steel, kg 220.4 261.0 *Expressed as equivalent number of tubes 1 in. in diometer; the array contains 193 which are 1 in. in diameter, 8 which are 3/4 in., and 25 which are 1/2, in, **Of this loading, 86.8% was in a solution having o U%33 concentration of 0.196 g/cma, and 13.2% was in one with a U3 concentration of 0.505 g/cm3. - ORNL-LR-OWG 3020 TY RELATIVE ACTIVA S [ CADMIUM FRACTICN 5 DISTANCE FROM REACTOR CENTER (cm) Fig. 4.5. Radial Neutron Distribution in Supercritical-Water Reactor Critical Assembly, 49 RELATIVE ACTIVITY 50 28 26 24 22 20 iy mn O ORNL-LR—-DWG 3021 [T e e _[ 77777 : ! I - [ ! [ . ; & i ! ‘ | ,,4:r7, _ i e e bl I — e 1 - . ,,,r ________ | ! | | | I | ‘ ‘ ‘ | | 4 i _ ) o o o —1. - . | INDIUM ACTIVATION | (R I L L ‘ | . —— ‘e ‘ e - L - —} 40 ] | | Lo i L L | l | \ 0.9 e - e e L - CADMIUM - COVERED | x| ‘ | F‘*’J— — TTINDIUM ACTIVATION.. © ! T g ‘ 0.8 1] ' i L x | s v I 0.7 e e R L 1 o | O ! i T 1 13 0 I 0 5 SO - R -4 06 = = s | | 5 | g | ‘ g @ | [ x i — e — — 0.5 & | ; . | \ s * B i i \ ga 1| ; ’ e ‘ ’r T T E | | i i | T e A ! < o e i T | —Jo3"” [ : ! =" CADMIUM FRACTION ) ! g ‘ ! ’ . : | i | 0 b S B | ' | | [ INDIUM ACTIVATION MINUS S —1 92 | i ! ! CADMIUM - COVERED INDIUM ACTIVATION ™ | . N . [ ey T T . | oa 1— i ; [ ! | | | ! i w‘ . - 1 I b L 0 0 5 10 15 20 25 30 35 40 45 50 55 80 85 70 75 1o DISTANCE FROM REACTOR TANK FLOOR (cm) Fig. 4.6. Longitudinal Meutron Distribution in Supercritical-Water Reactor Critical Assembly, Part 11 MATERIALS RESEARCH 5. CHEMISTRY OF MOLTEN MATERIALS W. R. Grimes Materials Chemistry Division Studies of the NaF-ZrF -UF , system were initi- ated after the solid phase studies in the NaF-ZrF, system were completed, ond to date no ternary com- pounds have been discovered in the NaF-ZrF -UF, system, A tenfative equilibrium diagrom has been prepared. The apparatus for visual observation of melfing temperatures was used to observe eleven NaF-UF, mixtures in the composition range 20 to 50 mole % UF ,, and a partial phase diagram of the system was prepared. A new apparatus has been constructed that will permit some manipulations to be carried out with the fused salts in an inert atmosphere. Recent afttempts to correlate the anticipated re- duction of UF, in the preparation of UF,-bearing melts with wet chemical analysis for UF, and UF and results of petrographic examination show some surprising anomalies. When UF , dissolved in LiF, in NaF-ZrF , mixtures, or in NaF LiF mixtures is treated under flowing hydrogen at BO0°C with ex- 90% or more of the UF, is reduced to UF,. However, when this techmque is applied to UF, in NaF-KF-LiF mixtures, the re- duction is only 50% complete at 800°C and, perhaps, 75% complete at 600°C. Petrographic exominations of the specimens reveal no complex compounds of tetravalent uranium; it is possible that the UF is “*hidden’’in solid solutions or in complex UF ;-UF, compounds in which it is not at present recogniz- able, A method for the large-scale purification of rubidium fluoride has been developed. All the rubidium fluoride obtained from commetcial sup- pliers to date has contained considerably more than the specified quantity of cesium compounds, and therefore almost all of it has been returned to the vendors for further processing. The purification method developed can be used at a reasonable cost cess uranium metal, on a large scale if material sufficiently free from cesium cannot be obtained from commercial sources, Values for the equilibrium constant for the re- action Cro 4 2UF, o= 2UF,; + GF, in molten NaZrF hove been re-examined by the use of various ratios of UF:;/UF4 with chromium metal in the charge., From the data obtained when this initial ratio is less than 3, average values of 5 % 10-4 at 600°C and 6 x 10~% at 800°C can be computed. At ratios larger than 3 the values in- crease regularly; this rise is probably due to the extreme difficulty in filtration and analysis of samples containing 1 to 20 ppm of Cr**, The values at low UF /UF ratios agree quite weli with the correspondmg vciues of 42 x 104 600°C and 4.1 x 1074 at 800°C obtained prevuously when UF, and chromium metal were the charge material. Fundamental studies were made of the reduction of NiF, and FeF, by H, in NaF-ZrF, systems as a means of determining possible improvements in purification techniques. Also, methods for pre- poring simple fluorides were studied. Additional measurements of decomposition potentials of KCI and of various chlorides in molten KCI at 850°C were made. Preliminary measurements of the solubility of xenon in KNO,-NaNO, eutectic (66 mole % NaNO ) show values of 8.5 x ]0 =8 and 10-7 mole/cm? at 280 and 360°C, respectively, af 1 atm xenon pres- The all-glass apparatus used has been re- structural metal sure, placed with o nickel and glass combination, and measurements of xenon solubility in molten fluo- rides are being made. Lorge quantities of purified ZrF -base fluorides are being prepared for engineering tests at ORNL and elsewhere, and the demand for purified fluo- rides of other types for possible reactor fuel appli- cation is increasing rapidly. Consequently, an increasing fraction of the effort is devoted to pro- duction or to research in direct support of produc- tion functions. Since the consumption of NaF-ZrF, and NaF- ZrF ,-UF, mixtures is expected to reach 10,000 Ib during the fiscal year, it is important to decrease the cost of production of this material. The HF-H, processing currently used is adequate from ali points of view, but processing times of nearly 100 hr per batch are now required with the rela- tively poor raw materials available. The sub- stitution of hafnium-free ZrF, and commercially available NaF .ZrF, for the impure ZrF, now used should afford « considerable saving. The 53 ANP QUARTERLY PROGRESS REPORT proposed use of zirconium metal as a scavenger (to replace most of the hydrogen processing) has shown promise on a small scale, and rapid electrolytic deposition of iron and nickel from the ZrF -base mixtures appears to be feasible. It is anticipated thai the price of purified NaZrF . and NaF-ZrF -UF mixtures may be halved in the next few months. SOLID PHASE STUDIES IN THE NanZrF4-UF4 SYSTEM C. J. Barton R. E. Moore R. E. Thoma Materials Chemistry Division G. D. White, Metallurgy Division H. Insley, Consultont Studies of the NaF-ZrF ,-UF, system were initi- ated after the solid phase studies in the NaF-ZrF binay system were completed. The solid phases present in both the slowly cooled and the quenched were studied by x-ray diffraction and The study of the NaF-UF, system was suspended, except for some visual-observation experiments, and will be resumed when time permits., Visual-observation sxperiments (cf. section below on ‘‘Visual Observation of Melting Temperatures in the NaF-UF ; System"’), together with petrographic studies of slowly cooled NoF—UF4 compositions, demonstrated that the compound previously desig- nated 1+?:3 as NaUF . is a congruently melting com- pound Nagll F, | that is anologous to the NagZrgF, somples petrographic analysis techniques. 41 Guenching experiments with fernary compositions on the join Na U F,-Naglr F show that this join comprises a completely (or nearly completely) miscible series, with liquidus compound.* temperatures descending from 722°C at the uranium compound to 520°C at the zirconium compound. Earlier thermal analysis daota and filiration ex- periments>*® demonsirated that there is another series of complete solid solutions along the join T4 R, Grimes ¢1 al., ANP Quan Prog. Rep. Man. 10, 1951, ANP-60, p 129. 2\, H. Zachaoriasen, J. Am. Chem, Soc. 70, 2147 (1948). 3¢. A Kraus, Phase Diagrams of Some Complex Salts of Uraniwm with Halides of the Alkali and Alkaline Earth Metals, M-251 (July 1, 1943). 4R, E. Thoma, et al., ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, Fig. 4.1, p 41. 5S¢ J. Barton, 5. A. Boyer, and R. J. Sheil, ANP Quar. Prog. Rep. Dec. 10, 1953, ORNL.-1649, pp 50 and 55. 6C. J. Barton and R. J. Sheil, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL.-1692, p 54, 54 Na,UF _-NayZrF ,, with the high liquidus tempera- ture {BgOOC) at the zirconium compound and the low liquidus temperature (629°C) at the uronium com- pound. In order to study equilibrium relationships in the region between these two solid solution joins, a series of nine compositions equally spaced along the Na,ZrF -Na UF . join was prepared, Phase analyses on completely crystallized preparations of these compositions indicate that Na,ZrF, and Na,UF, do not exist in equilibrium with each other and with liquid. Instead, the join between Na, ZrF, and Na,UF crosses the two three-phase triangles having their common base on the NagZrF -Na U F line and their opexes at Na,ZrF, and Na,UF,, Quenching experiments are being made for estab- lishing liquidus relationships and boundary curves in this area, Four ternary mixtures were prepared in order to obtain a preliminary indication of equilibrium rele- tionships in other parts of the system. The com- positions of these mixtures are given in Table 5.1, Identification, by x-ray diffraction and petrogrophic techniques, of the phases in completely crystal- lized preparations of the mixtures indicated that the compound Na.Zr, F .o occupies a small primary phase area contiguous to the primary phase fields of the NagZr F,,-NajU.F, | and the ZrF UF, solid solution series. It also appears that the primary phase field of the ZrF -UF , solid sclution series is configuous to the primary phase field of the NagZr,F,-NagUgF,, solid solution series along the greater part of the boundary curve that enters the ternary system at the Nagl,F -UF, eutectic (42 mole % NaF-58 mole % UF ,, melting point 680°C), follows approximately the 50 mole % NaF join for most of its course, and leaves the ternary system at the NagZrgF ,-Na Zr F o eutec- tic (approximately 50 mole % NaF--50 mole % ZrfF ,, melting point 510°C). The mixtures listed in Table 5.1 will be used for quenching experiments TABLE 5.1. COMPOSITIONS OF FOUR MIXTURES IN THE NqF-ZrF4rUF4 SYSTEM Sample _ Composition {mole %) Designation NgF ZrF4 UFA 12 41 55 4 T3 42 48 10 T4 48 39 13 in the near future to establish solidus and liguidus relationships. No fernary compounds have ‘been discovered, to date, in the NaF-ZrF -UF, system. A tentative equilibrium diagram, which shows the general relo- tionships believed to exist in this complex system, is presented in Fig. 5.1, 7R. J. Sheil and C. J. Barton, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 42. 162 PERIOD ENDING SEPTEMBER 10, 1954 VISUAL OBSERVATION OF MELTING TEMPERATURES IN THE NuF-UF4 SYSTEM M. S. Grim Materials Chemistry Division The apparatus for visual observation of melting temperatures, described in the previous quarterly report,’ was used to observe eleven NdF“UF4 mixtures in the composition range 20 to 50 mole % UF,. Liquidus temperatures obtained with this ORNL ~ LR~ DWG 2821 1000 LF, 1035°C 975 YALUES ARE TEMPERATURES IN °C 950 825 200 8?5 850 e NugUgFyy 800 L.— 825 -85 —875 - 900 NgF 999°C Z(F, 910°C Fig. 5.1. Tentative Equilibrium Diagrom for the NaF-ZrF ,-UF, System. 55 ANP QUARTERLY PROGRESS REPORT apparatus, both from visual observation of crystal- lizotion and from thermal effects recorded during the course of the visual observations, were generally a little higher than those indicated by the published diagram for this system.! The data are shown on the partial phase diagram for the NaF-UF , system in Fig. 5.2, A solid circle indicates the tempera- ture at which the composition oppeared to be com- pletely solidified. This point could not be de- termined with any degree of certuinty with some compositions, and the disappearance of the liquid phase was not always accompanied by a noticeable thermal effect. The liquidus temperatures shown in Fig. 5.2 seem fo indicate that NajUF, melts con- gruently at 629°C, that is, ot a temperature only a few degrees higher than the sutectic temperatures on both sides of the compound, The data alse support the results of petrographic studies which indicate that the compound in the 50 mole % UF, region has the approximate composition Na U F,, rather than NaUF . A new apparatus has been constructed that will permit some manipulations to be carried out with the fused salts in an inert utmosphere. |t consists essentially of a small dry box, with 5/B—in.-'i“hick ORNL*! R-0OWG 2922 700 - MPERATURE {°C) » w o Ll | e | \ | 600 : : : : 1‘ w ! 5 \ D o I o - LL’\ = uj,r 2 @ . om = | = 3 ‘ . =z ! B e e cemmme D e o e e 15 20 25 30 35 40 45 50 UFg (mole %} Fig. 5.2. Partial Pt ase Diagram of the NaF.UF System. 56 Lucite side walls and top, attached to a stainless steel bottom plate which has a 6-in. length of 2-in.-dia stainless steel pipe welded ontfo it to serve as a furnace core, The pipe is surrounded by resistance coils and insulating tape. The box is equipped with face plates for covering the glove holes so that the box can be partly evacuated before it is filled with inert gas, PHASE RELATIONSHIPS IN UF3-BEARING SYSTEMS C. 1. Barton R. J. Sheil l.. M, Bratcher A, B, Wilkerson Materials Chemistry Division G. D. White, Metallurgy Division T. N. McVay, Consultant It was assumed in the previous studies® of solu- bility relationships for UF, in various fluoride mixtures that in the presence of excess metallic uranium or zirconium the uranium present in the fused salt solution was trivalent. This assumption seemed justified, since, for the reaction 3 Lilo 7UF, + 4U°— UF, Brewer's tabulation® of standard free energies of formation yields AF® = =16 keal . In addition, petragraphic exomination of the melts obtained revealed the presence of red materiols tentatively identified as complex compounds of UF,, with green compounds of UF, showing only in traces, if ot all. Since these observations were those anticipated and since the petrographic ex- aminations had previously been shown to be very sensitive in detecting compounds of trivalent ure- nium in systems which had been slightly reduced, there seemed to be little reason fo believe that the results of such examination might be unreliable, During the past quarter, however, considerable evi- dence has been accumulated by the most accurate wet-chemical methods yet devised for determining U3Y and U*" in fluoride mixtures which shows that the reduction of UF, by excess uranivin metal is not complete at B00°C. While the reduction in BG. M. Watson and C. M. Blood, ANP Quar. Ffrog. Rep. June 10, 1954, ORNL-1729 p 51. ?1.. Brewer et al., Thermodynamic Properties and Equi- libria at High Temperatures of Uranium Halides, Oxides, Nitrides, and Carbides, MDDC-1543 (Sept. 20, 1945, rev. Apr. 1, 1947). NaF-ZrF, mixtures and in LiF appears to be 90%, or more, complete, in such mixtures as NaF-KF-LiF only about 50% of the UF, is reduced. Toward the end of the quarter it became evident that severd of the materials such as “3KF-2UF ;" previously believed to contain only U®¥ regularly contained large and varying quantities of U**. Furthermore, the crystals containing variable quantities of U4* are distinguishable as such only with great diffi- culty, if at all, by petrographic and x-ray diffraction examination, Consequently, the previously reported data on the solubility of UF; in various systems must be reinterpreted, The reasons for the incomplete reduction of the tetravalent uranium are not yet completely under- stood, and it is not possible at present to define the extent to which the reduction will proceed at various temperatures aond in the various solvents, Accordingly, the significance of much of the ma- terial presented below is not completely known, UF, in ZrF -Bearing Systems The solubility of UF, in NaF-ZrF ; mixtures was previously shown® to increase with increasing temperature and with increasing Zrf, concentration of the solvent over the range 47 to 57 mole % ZrF ,, Since the publication of that information it has been shown that the reduction of UF, in ZrF -bearing melts by excess uranium metal is slightly less than 90% complete at 800°C, The temperature de- pendence of the reduction is not yet known for this system, but it is likely that the UF, is more completely reduced at lower temperatures, An examination of the NaF-ZrF -UF; system has been ottempted by thermal analysis, with petro- graphic examination of the resulting solid phases, In these studies, UF:‘4 and excess uranium metal are added to the desired NaF-ZrF ; mixiure before the sample is heated, and the somple is stirred constantly while in the molten state. the reaction Therefore YUF, + hUe+— UF, might be expected to reach its equilibrium value at any temperature above the melting point. Conse- quently, in contrast to experiments in which the metallic uranium is removed by filtration af high temperatures, the solid products in slowly cooled melts might be expected to be nearly completely reduced. These studies indicate that UF, is the primary phase in systems in which the NaF-to-ZrF , ratic is PERIOD ENDING SEPTEMBER 10, 1954 less than 5 to 6 and that Na,U,F, appears as the primary phase in systems in which the NaF-to-Z¢F ratio is about 15, It also appears that Na,Zr F ., crystals may contain small quantities of UF, in solid solution. The solubility data obtained from these ond previous studies give little reason to expect that UF, can be dissolved in NaF-ZrF, mixtures in sufficient amounts to provide fuel for reflector-moderated reactors, UF, in NaF-KF.LiF Mixtures The solubility of UF, in the NaF-KF-LiF eutectic was stated previously'? to be equivalent to at least 15 wt % ot temperatures as low as 525°C, Subsequent careful examination of this system has revealed that when UF, and an excess of uranium metal are added to the purified NaF-KF-LiF mixture the dissolved uranium species aggregate at least 22% total uranium in the mixture. However, it is obvious that only 40 to 45% of the soluble uranium is present as UF; at 800°C, while 55 to 60% may be trivalent at 600°C, Further study of the system will be necessary before these values can be determined more accurately, Thermal analysis date have been obtained for several mixtures which were prepared from UF , and the NaF-KF-LiF eutectic and then treated with an excess of uranium metal. The data obtained, as shown in Table 5.2, are in agreement with the data obtained from filtration studies which showed high uranium concentrations at low temperatures. Petro- graphic examination of these materialsreveoled that at low uranium concentrations a red phase (refractive index, about 1.44), which is probably I'(::,)UF:‘S and which may contain UF ,, is predominant. At high 105, M. Watson and C. M. Blood, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 53. TABLE 5.2. THERMAL ANALYSIS DATA FOR UF3-BEARING NaF-KF-LiF MIXTURES Theoretical Composition (mole %)* U° Used Thermal Effects (% of theory) Q) NaF KF LiF U F3 10.8 39.5 43.7 6.0 200 475,455 16.6 38.6 428 8.0 110 510,460,455 10.¥ 37.0 40.9 120 110 520,455,445 9.6 353 39.1 140 130 565,475 8.8 322 357 23.3 200 570,490,470 *Bgsed on complete reaction of UF, with ue, 57 ANP QUARTERLY PROGRESS REPORT uranium concentrations an olive-drab cubic phase (refractive index, 1.46) and a blue biaxial phase (refractive index, 1.544) also appear. There is evidence that the blue material is a solid solution of Na,U,Fqs-K,U,Fg; whether it dissolves UF, is not known. UF3 in NaF-RhF.LiF Mixtures The previously reported!! data on the solubility of UF, in the NaF-RbF-LiF eutectic composition were based on the assumption that all the uranium in the filtered specimens was UF ;. For more.recent studies, the solubilities were determined by using 200-g samples of melt, mechanical agitation during the equilibration period, and 20% excess uranium for the reduction step. Samples of the equilibrium mixture were withdrawn with a filter stick contain- ing sintered nicke!l as the filter medium. The samples were then analyzed for trivalent uranium and for total uranium. The results of the analyses are presented in Table 5.3. VIR, J. Sheil and C. J. Barton, ANP Quar, Prog. Rep. June 10, 1954, ORNL-1729, p 53, The significance of the analyses for UF, is not entirely clear, Petrographic examination of the solidified filtrates indicated o reddish-brown ma- terial, presumed to be RbyUF ., as the only colored phase, While it is possible that this material could accommodate UF, in solid solution, it deoes not appear likely that as much as 75% of the total uranium present could be hidden in this fashion, The good agreement, for some of the mixtures studied, between the theorstical uranium content and the measured total uranium content indicates that the reaction of UF, with uranium metal pro- ceeded nearly to completion. Additional data on this system will be needed before an interpretation of the results can be presented. UF3 in the Individual Alkali-Metal Fluorides Filtered specimens from preparations in which UF, dissolved in LiF was treated at 825 to 850°C with excess uranium metal have shown that more than 90% of the dissolved uranium is trivalent and that it crystallizes as UF;. Accordingly, it is believed that the behavior observed by thermal analysis of this system is that of LiF with pure, TABLE 5.3. SOLUBILITY OF UF, IN NoF-RbF-LiF MIXTURES AT VARIOUS TEMFERATURES Theoretical Composition Theoretical Filtration Filtrate Composition (mole %) U3* Content Temperature (wt %) Naf RbF LiF UF, (wt %) e udt Total U 14.1 37.6 42.3 6.0 19.3 793 2.92 15.2 700 1.41 16.9 645 4.54 16.7 600 1.17 14.7 550 10.7 495 5.26 9.4 47.0 37.6 6.0 17.7 750 5.51 17.0 675 9.18 16.9 650 3.22 18.0 625 9.95 16.4 600 No sample* 5.6 48.9 39.5 6.0 17.5 800 5.07 16.8 750 4,02 17.0 700 4.07 17.3 650 4.76 16.4 625 4,86 13.1 600 6.35 * Filtration was attempted, but the filter apparently clogged at this temperature ond the sample could not be ob- tained. 58 or nearly pure, UF,. Recent studies of LiF-UF, mixtures have confirmed the belief!2 that no com- pounds form and that these materials comprise a simple eutectic system. A partial phase diagram for this simple system is shown as Fig. 5.3. The NaF-UF, system has been studied more thoroughly than any other binary alkali-metal fluoride—UF, system. The best thermal analysis data obtained for this system (data obtained with those mixtures which showed the least evidence of contamination of the melt by tetravalent uranium) are shown in Fig., 5.4, The earlier data, 13:14 ob- tained by mixing NaF with previously synthesized UF; in sealed capsules, are shown by open circles. The more recent data, shown by solid circles, were 12, M. Bratcher et al., ANP Quar, Prog. Rep. June 10, 1954, ORNL-1729, p 43. 13y, s, Coleman and W. C. Whitley, ANP Quar. Prog. Rep. Sept. 10, 1952, ORNL-1375, p 79. My c. Whitley, V. 5 Coleman, and C. J. Barton, ANP Quar. Prog. Rep. Dec. 10, 1952, ORNL-1439, p 109. ORNL—-LR—DW( 2923 1000 Q50 350 BOO [ TEMPERATURE (°C) — o C ?OO ..... e e L - B5O e bt T o oo J) AN R - —Lv ----- o 1o J5 ESCUUNIUN NONIIURUUE SO0 SN SN ] Q 500 L | - ] 0 10 20 30 40 50 80 TC UFy (mole %) Fig. 5.3. Tentative Partial Phase Diagram of the LiF-UF, System, PERIOD ENDING SEPTEMBER 10, 1954 obtained by reducing NaF-UF, mixtures with ex- cess uranium in open crucubles protected by an atmosphere of helium. Some data, also obtoined during the past quarter, were obtained by a method which represents a compromise between the two previously mentioned methods, Mixtures were pre- pared in open stirrer-equipped crucibles by combin- ing dry NaF with freshly prepared!? UF,. These data are indicated in Fig. 5.4 by the solid triangles. The agreement between the data obtained by the various methods is considered fo be gratilying, in view of the difficulties associated with the ease of oxidation of U*" to U4*, The identification of the compound is based chiefly upon the results of petrographic examination of slowly cooled melts, The possibility that the compound might be Na ,UF, rather than Na,U,F, and the pOSSIbI]Ify thof appreciable quantities of UF, may be “hidden’’ in these materials have not been completely ruled out by these studies. Very few compositions contain- ing more than 50 mole % UF, have been prepared, and the data in this part of the system are probably of little value, ORNL~( RADW(" 2924 M!XTURFS PREPARED N %m_co CAFTSUL.ES * MXTURES PREPARED IN OPEN CRUCIBLES : IN A HELIUM ATMOSFHERE T MIXTURES PREPARED N OPEN STIRRER - EQUIPPED CRUCIBLES .. Q OO - I TG0 [ TEMPERATURE °C} BO0 o= o0 - - I T UF 3 {mole ¥) Fig. 5.4. Tentative Partial Phase Diagram of the NuF-UFs System. Phase equilibrium studies of the KF-UF, and RbF-UFB systems are not so far advanced. |t appears that the most probable compositions for the compounds observed are K ,UF,, K,U,F,, ‘and Rb,UF, ond that the KF-K UF, eutechc melts at about 720 C at upproxumately 15 mole % UF,, It lsPrepmed'by W. C. Whitley, July 1954, 59 ANP QUARTERLY PROGRESS REPORT seems that in both systems considercble UF, is present along with the UF, in a manner such that it can escape detection by petrographic examination, UF, in Binary Alkali-Metal Fluoride Systems A number of thermal analyses followed by petro- graphic examination of the solid products have been performed with UF, in each of the NaF-LiF, NaF-KF, and LiF-KF eutectic mixiures. These specimens were prepared, in each case, by reducing the desired UF ,~alkali-metal fluoride mixture with 10% excess metallic uronium in nickel crucibles blanketed with inert gas and fitted with stirrers of nickel. In the NaF-LiF system the lowest liquidus tem- perature observed was at about 630°C at 10 mole % UF; ond 36 mole % NaF, In all cases studied (up to 30 mole % UF ;) only the free alkali-metal fluo- rides ond the NaF-UF ; complex compound (Najl,F o) were observed. In the NaF-KF-UF; system the red compound K3UF, appears as the only colored species in mixtures containing up to 10 mole % UF,; the optical properties are slightly different from those of the compound crystallized fromNaF-free mixtures and may indicate slight solid solution of NaF in the crystals. Mixtures containing 20 to 40 mole % UF; show increasing quantities of a blue phase with optical properties which suggest that it is a solid solution of NajU,Fg and K U, 1. Liquidus temperatures in this system seem to be higher than 700°C in all cases. Solutions of UF, in the LiF-KF binary system show liquidus temperatures in the range 485 to 600°C. Petrographic examination of thesematerials shows free LiF in all specimens, K,UF, in those containing less than 10 mole % UF,, and K U, Fy in those containing much above 10 mole % UF,, Chemical analyses were not performed on any of these materials, but it is likely thot considerable UF , was present in these specimens, especially at the higher temperotures. Additional studies of these potentially valuable systems will be made, PURIFICATION CF RUBIDIUM FLUORIDE C. J. Barton, Materiols Chemisiry Division D, L.. Stockton, ORSORT All the rubidium fluoride cbtained from comimercial suppliers to date has contained considerably more than the specified quantity of cesium compounds, and therefore almost all of it has been retumed to 60 the vendors for further processing. However, the immediate need for small amounts of pure material for phase equilibrium studies and for rubidium metadl production has necessitated purification of some material at ORNL. Experiments '¢ have shown the feasibility of separating rubidium from cesium with the ion ex- change medium Amberlite IR-105, This resin was, accordingly, used in all the experiments conducted, The resin column consisted of a 3-in.,-dia glass pipe containing a 6-ft bed of the resin. The column was loaded for each run with about 350 g of crude RbF known to contain about 20% CsF. Elution of the rubidium was accomplished in about 35 liters of 0.5 M (NH ) ,CO,4 solution; the purity of the product was not appreciably affected by elution rates in the range 30 to 60 ml/min. Elution of cesium from the column, which was observed by use of Cs'37, required 65 to 70 liters of the (NH,),CO, solution; elution was not greatly improved by use of a 1.5 M solution. From a total of 1775 g of crude RbF charged in five different runs, somewhat more than 1 kg of RbF containing less than 1% CsF was obtained. The KF content (0.6% KF) was not appreciably affected by this treatment. be conducted on a large scale at reasonable cost if It is apparent that this purification con material sufficiently free from cesium cannot be obtained from commercial sources. CHEMICAL REACTIONS IN MOLTEN SALTS F. F. Blankenship L.. G. Overholser W. R, Grimes Materials Chemistry Division Chemical Equilibric in Fused Salts J. D. Redman C, F. Weaver Materials Chemistry Division The apparatus and techniques for experimental determination of edquilibrium constants for the q reactions Ce® + 2U'r:4 _ CrF2 + 2UF3 and Fe® + 2UF, == FeF, + 2UF, in molten NaZrFs were described previously, V7 and 161, R. Higgins, Chemical Technology Division, ORNL, work to be published. 17 6. Overholser, J. D. Redman, and C. F. Weaver, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL.-1692, p 56. values were given for the “‘equilibrium constants'’ obtained with UF, and the metal as starting ma- terials. During the past quarter, the values have been checked with various ratios of UF4 to UF, as starting materials in the NaZrF . solution. The apparatus and details of the experimental technique were the same as in previous experiments. The results of the recent series of experiments in which 2 g of hydrogen-fired {(1200°C}) Cr® and various UF ;-to-UF; rotios were used are shown in Table 5.4. In the table c?. . .C UF3 CrF2 C2 UF4 with the concentrations expressed as mole fractions by using the formula weight -NaZrF5 = 2 moles, It was necessary to calculate the final UF_ con- centration, since accurate analyses for U3 in the presence of Crtt could not be made; any error arising from this calculation would be important only at low UF; concentrations, The data show that at both 600 and 800°C the PERIOD ENDING SEPTEMBER 10, 1954 values of K _ remain constant at low UF 5-to-UF ratios but increase rapidly as this ratio rises above 3. This rapid increase in K may be due to in- complete removal by filtration of small amounts of chromium metal or to uncertainties in analysis of these frace quantities. Since, as shown above, the UF ;-to-UF, ratio in NaZrF . solution in equilibrium with uranium metal is about 9 at 800°C, it is not likely that disproportionation of the UF, is re- sponsible for the increase in K . In any case, the agreement between the meon of four values of the constant ot each temperature with previously published values is extremely good. Agreement among the various initial Uf""?;i'o:«U[':4 ratios seems to show that a state of equilibrium is reached in these experiments, When equilibrium data for the reaction Fe® 4+ 2UF4F~—"_——"'32UF3 + FeF2 were obtained by the addition of UF,, UF, and FeF, to NaZrF, in the same way as for the chro- mium reaction, the results obtained were not so reproducible. The data obtained are shown in Table 5.5, The mean of the values obtained at TABLE 5.4, EQUILIBRIUM DATA FOR THE REACTION Cr® + 2UF4V_—-..—:“£ CrF, + 2UF, IN MOLTEN NaZrFg Experimental Conditions Equilibrium®* Concentration Tem([.:ecr;ture ( IUF jkAdc:ed | UF, Added of Crtt K.>** moles/kg of melt) (moles/kg of melt) (ppm) 600 0.358 2080 3.0 x 1074 0.267 0.078 950 5.1 x 10~4 0.178 0.158 150 2.7 x 10~4 0.089 0.238 35 4.4 x 104 0.035 0.287 50 5.7 x 103 0.029 0.318 40 1.1x 1072 800 0.360 2660 7.4 x 10—% 0.280 0.081 1125 6.1 x 10™4 0.194 0.165 210 3.3 x 1074 0.111 0.244 55 4.2 x 10~4 0.064 0.295 30 1.8 x 1073 0.032 0.325 35 10-3 7.9 x *Yalues shown are mean of closely agreeing volues from duplicate experiments. Blank centained 200 ppm of cett, **K, values previously obtained: ot 600°C, K, =4.2x 10~4,; Keqcalculmed from ~AF°= RT aneqby using Brewer’s values for AF®: at 6OOOC,KGq = 1; at 800°C, K, = at 800°C, K_= 4.1 10~4, q 0.]- 61 ANP QUARTERLY PROGRESS REPORT TABLE 5.5. EQUILIBRIUM DATA FOR THE REACTION Fe° + ZUF“;“, ZUF3 + FeF, IN MOLTEN NaZrFs Experimental Conditions Concentration of Temperature UF, Added UF, Added FeF, Added Fe'™ in Filtrate Ky (°C) (moles/kg of melt) (moles/kg of melt) (moles/kg of melt) (ppm) 600 0.322 0.053 0.0319 620 1.2 x 106 0.322 0.053 0.0346 735 1.2 x 107¢ 0.326 0.083 0.0347 660 9.3 x 1070 0.326 0.083 0.0347 655 9.3 x 107¢ 0.321 0.036 0.0168 995 2.1 x 1077 0.321 0.036 0.0168 940 1.9 x 1077 800 0.242 0.113 0.206 8600 1.2 x 1073 0.242 0.113 0.103 3210 3.2 x 1072 0.242 0.113 0.103 3270 3.7 x 1073 0.277 0.077 0.103 4110 2.7 x 1073 0.277 0.077 0.103 4050 2.2 x 1077 0.0318 0.322 0.213 3110 3.4 x 106 0.0318 0.322 0.210 3300 2.5 x 1073 0.103 0.271 0.210 4530 1.1 x 1079 0.103 0.271 0.207 4645 3.7 x 1077 0,207 0.152 0.213 7940 1.3 x 1072 0.245 0.115 0.053 495 4.2 x 10-6 0.245 0.115 0.053 435 2.8 x 1076 0.245 0.115 0.053 265 1.3 x 1076 *Values previously reported: at 600°C, K = 1.2 x 10“5; at 800°C, K,=1.5x 10-8. Keq calculated from —-AF° = RT In K, by using Brewer’s values for AF°: at 600°C, K, = 2x 107% at 800°C, K = L4x 1074 600°C (1 x 10=7) agrees with the previously re- ported value of 1.5 x 10~3, although agreement among the individual determinations was not en- couraging. The values obtained at 800°C, however, do not agree well with the value of 1.5 x 10~ reported when UF, and ferrous metal were used as starting materials. In on ottempt to explain this discrepancy, the experiments in which UF | and ferrous metal were used as reactants at 800°C were repeated. The values obtained, os shown in Table 5.6, yielded a mean of 0.9 x 1078 and ogreed more closely with the previously reported values. It appears possible that the reoction of ferrous metal with UF, had not reached equilibrium in the reaction time allowed. 62 TABLE 5.6. EQUILIBRIUM DATA FOR THE REACTION OF FERROUS METAL WITH UF, iN MOLTEN NoZrF, AT 800°C UF4 added: 0.360 mole/kg of melt Concentration™ of Fettin Filtrate K, (ppm) 520 1.7 x 1078 390 5.4 x 1077 330 2.8 x 1077 480 1.1 x 1076 *Blank of 100 ppm to be subtracted from determined values in calculations. Stability of Chromium Compounds in Molten Fluorides J. D. Redman C. F. Weaver Materials Chemistry Division It was shown in previous experiments that CrfF, or some complex compound of divalent chromium was the corrosion product in the ZrF4ubuse fluoride melts of interest., However, when NoF-KF-LiF mixtures, with and without UF . were tested in equipment prepared from olloys containing chromium, the reaction products found were complex com- pounds of CrF . Since films that were apparently NaK,CrF . or similar materials have been observed in some experiments, attempts have been made to observe the stability ond solubility of CrF, and CrF, in typical salt mixtures of the two fypes. In the temperature range 600 to 800°C, CrF, is not stable in NaZrF. solution in contact with equipment of nickel. Examination of filtered speci- mens shows that the reaction ECrFB + Ni&=—= NEF2 + 2CrF2 The solubility of CrF, in the NaZrF g solvent seems fo exceed 6000 parts of Cr per million parts of salt at 600°C; solubility ot 800°C seems to be at least 15,000 ppm and may be considerably higher, In the NaF-KF-LiF system it appears that CrF, is not stable. action proceeds to essential completion, Whether the disproportionation re~ 3CrF2 ==y 4 ZCrF3 or some other mechonism is involved is not evident from the data available, It appears from preliminary evidence, however, that the solubility of Cr3% is several thousand parts per million at 600°C, Reduction of NiF, by H, in NaF.ZrF, Systems C. M, Blood H. A, Friedman G. M, Watson Materials Chemistry Division The use of hydrogen as a reducing agent in the removal of oxidizing impurities from fluoride melts has been routine for almost two years. One of the principal impurities removed is the Ni** introduced as a result of treating the melts with hydrogen fluoride in nickel containers at 800°C. Since hydrogen was first employed in this fashion, kinetic studies have been under way in order to supply data on which improved fluoride production procedures PERIOD ENDING SEPTEMBER 10, 1954 could be based, The Ni*" was found to reduce so rapidly thot the controlling step was the rate of removal of hydrogen fluoride from the melt; hence the chemistry of the process presents no problem of consequence from an engineering standpoint. With the completion of measurements af 800°C during the pust quarter, these studies have been discontinued, The experimental method used for the measure- ments at 800°C was the same as that used in previously described experiments ot 600 and 700°C.'8 Figure 5.5 shows a comparison of the effect of temperature on the rates of reduction for four concentrations. The slopes of the curves show that the apparent activation energy for the rate de- termining step is about 7800 cal. This energy of activation is in accord with the hypothesis that the reaction is heterogeneocus and is catalyzed by nickel surfaces. 186G, M. Watson, C. M. Bload, F. F. Blonkenship, ANFP Quar, Prog. Rep. Mar. 10, 1954, ORNL-1692, p 62. UNCLASSIFIED ORMIL-LR—OWSG 2525 TEMPERATURE 5 X 1072 2 X402 1072 [ 5 %4073 EGUIVALENTS OF NEF2 REDUCED PER liter OF B FASSED 2Rt L] | ] 10~3 e GODIRB Q0017 00016 0.0015 0004 GO 00012 A Fig. 5.5. Effect of Temperature and Concentra- tion on the Rate of Reduction of NiF, by H, in Molten NoF-ZrF, (53-47 mole %). 63 ANP QUARTERLY PROGRESS REPORT The data presented here were all obtcined at roughly the same gas flow rate (100 mi/min). However, since the rate of reduction is dependent on the flow rate and on the geomeiry of the appo- ratus, the results must be considered merely as representative and are subject to some voriation with conditions; they apply to 3-kg batches in 4.in,-dia pots with hydrogen bubbles rising from the notched end of a ]/4-in. vertical dip tube im- mersed to a depth of 5 in. Reduction of FeF, by H, in NaF-ZrF ; Systems C. M. Blood G. M, Watson Materials Chemistry Division Since the reduction of FeF, with hydrogen is the step in purification of NaZrk, mixtures for reactor use aond testing, an extensive study of the equilibrium constant of this reaction has been attempted. FeF, + Hye===2Fe® 2HF time-consuming The reaction is relatively easy to study, since the hydrogen fluoride concentration in hydrogen can be accu- rately determined, and, after filiration of the melt, accurate determinations of FeF, in the melt can be obtained. Since at these elevated temperatures the activities of hydrogen and hydrogen fluoride can be assumed to be equal to their partial pressures in the gas phase, then where Cp_ e is expressed in mole fraction, and from the relationships eq and the activity coefficient for FeF, at low concentra- tions in NaZrk equilibrium estimated by two different methods: method and on equilibration method. should be easily determined. The constant for the reaction has been a dynamic Dyncmic Methed. In preliminagry experiments in which H, was passed through molten NaF-ZrF, {53-47 mole %) contaminated with known quantities 64 of FeF, or CrF,, the HF concentration of the effluent H, was studied as a function of the H flow rate. The dota obtained are plotted in Fig. 5.6, in which the ratio of HF concentration at a given flow rate to that found at a standard flow rate (210 ml/min) is plotted against flow rate. As antici- pated, the ratio drops at high flow rates; however, the ratic shows no sign of leveling off at low flow rates, as would be expected. Although the curve shows no justification for the practice, the lowest flow rate found to be practicable (9 ml/min) was considered to be equivalent to a zero flow rate (thot is, the flow rate at which the HF concen- tration could be expected to be at equilibrium), and experimental values found in subsequent ex- periments were corrected to the concentration that they would have shown at such flow rate, UNCLASSFIED ORML- LR-DWG 2926 35— : —— e — 3.0 STANDARD FLOW RATE = 210 ml/min 800° 25 A FeF, +Hy 0% pe 4+ our 800°C O CrF, +Hy, s G 4+ 2HF 20 RELATIVE EF CONCENTRATION o —mm——MMM ; | 0 100 200 300 400 FLOW RATE (mi/min) Fig. 5.6. Eifect of Gas Flow Rate on Relative Saturation of Effluent Gas with HF When Possed Through Melten NaF-ZrF, (53-47 mole %) Con- tominated with Known Quantities of FeF2 or Cer, The experiments were performed by adding known quantities of FeF, to anNaF-ZrF, (53-47 mole %) melt in a clean nickel reactor and measuring the HF concentration of the effluent hydrogen passed at carefully measured flow rates. The cumulative total of HF removed was also determined by col- lection of the HF in caustic seclution and titration of the excess caustic, The Fel, concentration at any specified time could be calculated from the Fef, found by analysis on completion of the ex- periment and the HF yield after the specified time, This procedure, which is simple, rapid, and capable of high internal precision, yielded the values plotted as the lower curve in Fig, 3.7. The logarithm &f the HF concentration is plotted against the logarithm of the FeF, concentration at various flow rates of hydrogen. When these values are corrected to the ‘‘zero flow rate' (9 ml/min) by ~2 PERIOD ENDING SEPTEMBER 10, 1954 reference to Fig. 5.6, they vyield a consistent straight line after on initial peak. This initial peak is probably due to the reaction of HF at its initial high concentrations with the clean (hydrogen-fired) nickel apparatus and its subsequent re-equilibration with the more dilute HF-H, mixture. The straight- line portion of the curve shows the expected slope UNCLASSIFIED ORNL - LR-DWG 2927 VALUES CORRECTED TG GAS FLOW RATE OF 9 ml/min CONGENTRATION OF HF IN EFFLUENT GAS {moles/liter) 10 2 5 CALCULATED FROM FREE ENERGY DATA (BREWER) | EXPERIMENTAL VALUES AT VARYING GAS FLOW RATES C 210 ml/min A (AS LABELED) 3 10 2 5 10 CONCENTRATION OF Fef, (ppm Fe' ') Fig. 5.7. Equilibrium Concentration for the Reduction of Fel”"2 by H2 in Nul"'«ZrF’4 {53-47 mole %) at 800°C {Obtained by Dynamic Method). 65 ANP QUARTERLY PROGRESS REPORT of 0.5 and agrees closely with the struight line calculated from Brewer's value of AF° for the reaction, The values obtained for K and for YEeF, on the assumptlion that the activities of HF and H, are proportional to their partial pressures ot 800°C and that AFOFer = -65.4 kcal per atom of F and AFOHF = ~65.8 kcal per atom of are shown in Table 5.7, The agreement of the values obtained over a fourfold concentration range of FeF, is extremely good. It appears that FeF, in molten NaF-ZrF , is in nearly ideal solution, TABLE 5.7. EQUILIBRIUM CONSTANTS AND ACTIVITY COEFFICIENTS AT 800°C IN MOL.TEN NflF-Zer {53-47 mole %) FOR THE REACTION Hy o+ Ferr_—“‘\Feo + 24F Concentration P2HF of Fef, - VEoF, " K /146 (pom Fe) Py 'CreF, 1598 2.15 1.47 1513 2.15 1.47 1458 2.17 1.48 1372 1.98 1.36 1318 1.91 1.31 1233 2,37 1.62 1122 2.07 1.38 1067 2.00 1.37 983 2.1 1.44 900 2.18 1.49 857 2.11 1.45 760 2.17 1.49 606 2,19 1.50 550 2,24 1.54 481 2.19 1.50 Av 2.1310.22 1.46 £ 0.16 Equilibration Method, Since it was doubtful that the lowest flow rate found to be practicable (9 ml/min) actually tions, an attempt was made o determine the equi- librium conditions directly. A mixture of HF and H, was bubbled through a melt containing Fe® and FeF, until the influent HF concentration matched represented equilibrium condi- 66 that of the effluent. At that time a sample was drawn through a sintered nickel filter ond analyzed to determine the FeF, concentration, Since the decomposition pressures of molten KF-HF saturated with KF depend only on tempera- ture, 19 definite mixtures of H2 and HF could be obtained while hydrogen gas was passed over such mixtures held ot a constant temperature. The mixtures of H, thus obtained were used as influent for the reactor containing the salt mixture, In a typical experiment in which this method was used, a batch (~3 kg) of previously purified NaF- ZrF4 (53-47 mole %) mixiure was loaded into g clean nickel reactor. The mixture was heated to 800°C and further purified with hydrogen until a very low HF concentration (3 x 10~ mole/liter) in the effluent gas denoted high purity of the melt. The mixture was allowed to cool to rocom tempera- ture, and FeF , and iron wire were added to the cold, frozen mixture under an atmosphere of argon. The mixture was then heated, melted, and brought up to temperature while it was sparged with helium gas. The reactor containing the KF«HF or HF generator was heated separately to a convenient initic! Pure was passed in parallel through both reacters — the HF generator and the equilibration reactor — in order to clean out any NiF. films on walls and to precipitate nickel impurities. After 1 to 2 hr of parallel operation, the reactors were connected in series, ond the inlet and effluent gas streams were continuously sampled and ana- lyzed. The temperature of the furnace of the generator was then adjusted until the analyzer temperature and connected to the system. hydrogen showed H -HF mixtures of the desired composition, The gas mixture was allowed to bubble through the system for extended periods of time, and somples were taken intermittently for onalysis. When it was considered that the system had approached equi- librium, as indicated by a convergence of the in- fluent and effluent compositions, the HF generator was removed from the system and helium wos sub- stituted for the influent gas mixture. A filtering tube was introduced into the melt under an addi- tional argon atmosphere and o filtrate was obtained, The filtering tube with the frozen filtrate was re- moved ond analyzed for iron, This procedure was repeated with appropriate variation of the tempera- ture of the HF generator and consequent variation of the influent gas composition, 19G. H. Cady, J. Am. Chem. Soc. 56, 1432 (1934). The data obtained are presented in Table 5.8, which gives the concentration of iron found in the filtrate, the ratio of the MF concentrations in the influent and effluent gases, and the equilibrium constants ond corresponding activity coefficients for the FeF,. As is evident, the data in Table 5.8 are less precise than those obtained by the dynamic procedure., The lowered precision is probably due to the more complex apparatus and technique. How- ever, when the data are plotted (Fig. 5.8), the spread of the data is less thon the estimated un- certainty in the AF® values. tween the values obtained by the two methods is remarkably good, TABLE 5.8. EQUILIBRIUM DATA FOR THE REACTION FeF2 + H2"""“*"‘,‘...._,_FfeD + 2HF AT 800°C IN MOLTEN NuF-Zer {53-47 mole %) The agreement be- i Ratio of HF in 2 Concentration Influent Gas P HF of of FeF2 to HF in ; yFer (ppm Fe) Effluent Gas HZ'CFng 912 0,79 0.60 0.41 223 0.91 1.08 0.74 915 1.02 2,09 1.43 740 0.98 2.12 1.45 360 1,00 1.71 1.17 185 0.93 1.89 1.30 160 0.91 2,14 1.47 275 0.99 3.12 2.14 Av 1.34 1.25 Preparation of Various Fluorides B, J. Sturm E. E. Ketchen Materials Chemistry Division The simple structural metal fluorides have con- tinued to be used extensively for numerous pur- poses. A limited interest also has been maintained in several of the complex fluorides formed by the interaction of alkali fluorides with structural metal fluorides. Some effort also has been expended on the preparation of other simple fluorides, as well as on the purification of small quantities of the struc- tural metal fluorides. The identity and purity of these fluorides have been established mainly by chemical analysis supplemented by x-ray and petro- graphic data in some cases. The compound AgF, with an Ag-to-F ratio of 0.98, was prepared by heating Ag,CO, under HF PERIOD ENDING SEPTEMBER 10, 1954 at 150 to 200°C, An attempt to prepare AgF by reacting HF with AgCN was unsuccessful; most of the AgCN did not react. Several batches of an- hydrous ZnF, were prepared by heating ZnF »4H.,0 under HF at 600°C., The compound TeF, was prepared by reacting Te and HN03 and reacting the resulting TeO , with aqueous HF 1o give TeO,F ,-H 0, which was thermally decomposed to vyield T'eOZ, HF, H,0, ond the product TeF,. The TeF, wos collected in a condenser held at 130 to 140°C, Small quantities of Rb,CrF . and Li,RbCrF were prepared by heating the proper quantities of CrF RbF, and LiF in sealed capsules at 900°C. The Rb3CrF6 has properties approximating those of o corrosion product found in a rubidium-base fuel. The compound RbCrF, was prepared by heating CrF, and RbF at 800°C in a sedled capsule. Numerous batches of structural metal fuel- or coolant-type fluoride complexes were prepared. Among these (prepared by methods previously described) were: K,CrF,, K,NiF,, and KNiF,. The compounds NiF_, FeF,, and FeF, were pre- pared in a high state of purity by hydrofluorination of the corresponding chlorides, FUNDAMEMTAL CHEMISTRY CF FUSED SALTS EMF Measurements L. E. Topol Materials Chemistry Division Additional measurements of decomposition poten- tials of KCl and of various chlorides in molten KCl ot 850°C were muade. Morgonite (A1,0,) crucibles were employed with all the melts studied. Platinum and nickel were used as cathodes, and the anodes were of nickel, carbon (graphite), chromium, and zirconium. The solutions of various chlorides in KCl were prepared by melting 2 moles of KCI with 1 mole of the anhydrous chloride in a sealed, evacuated quartz tube, More dilute solutions were prepared by adding appropriate quantities of KCl to the 33.3% KCl mixtures so obtained. The data obtained recently are given in Table 5.9. The value of 1.5 v found upon electrolysis of KCI with on anode of electrolytic ‘chromium agrees with the value previously reported for chromium-piated gold wire in an atmosphere of helium,2? The more recent value of 1.27 v for electrolytic chromium in 20y | E, Topol, ANP Quar. Prog. Rep. Mar 10, 1954, ORNL-1692, o 43, 67 ANP QUARTERLY PROGRESS REPORT 1072 UNCLASSIFIED ORNL—-LR--DWG 2928 COMCENTRATION OF HF IN EFFLUENT GAS {moles /liter) . _ CALCULATED FROM FREE—ENERGY DATA (BREWER) | e — ] 2 T AF® FOR FeF,= ~65.4 % 1 keal | | | EXPERIMENTAL DATA _____ CALCULATED FROM FREE—ENERGY DATA (GLASSNER) AFC® FOR FeF, = -~68.0 kcal | 107 ‘ | . [ , 1 N 102 2 5 10° 2 5 104 CONCENTRATION FeF, (ppm Fe*™*) Fig. 5.8. Equilibrium Concentration for the Reduction of FeF, by H, in NaF-ZiF , (53-47 mole %) ot 800°C (Cbtained by Equilibration Methaod). a hydrogen atmosphere may indicate that a hydrogen electrode is formed. The reaction of UCI; with zirconium metal may be ascribed to the following equations: 2UC|3 + 3Zr = BZrClz + 20, AF° = 10 kcal, and 4UCL, + 3Zr = 3ZrCl, + 4U, AF° = +130 kcal. 68 Current-voltage measurements indicate the first reaction to be predominani. The E° of 0.4 v for the reaction of UCI3 with a chromium anode, that is, UCH; + 3Cr = 3CCH, + 2U E® (theoretical) = 0.90 v, seems to indicate strong complexing of the CrCl,. There seems, from studies by other groups, to be PERIOD ENDING SEPTEMBER 10, 1954 TABLE 59. DECOMPOSITION POTENTIALS OF VARIOUS CHLORIDES IN KCI AT 850°C Concentration of . Electrodes Blanketing Potential Observed Salt Salt in Solute . Cathade Anode Atmo sphere {v) {mole %) KCi 100.0 Ni Cr He 1.50 UC|3 33.3 Pt Zr He Spontoneous 33.2 Pt Cr He 0.40 UCI4 33.3 Pt C H2 0.95, 0.83, (0.75) 33.3 Pt C He 0.83, 0.95 33.3 P Ni He 0.60 to 0.80 3.0 Pt C He 1,84 to 1.87 3.0 Pt C H2 (1.15) 3.0 Pt Ni He 0.34 FeCl2 33.3 Ni C Me 0.80 10 0.83 33.3 Ni Ni He 0.35 to (.45 2.0 Ni C He 1.45 i.0 Ni C He 1.50 10.0 Mi C He 1.Q0 Ni(:l2 33.3 Ni C He 0,93, 0.82 3.0 Ni C He 1.13 to 1.17 1.0 Ni C He 1.20 little evidence for this behavior in fluoride melts, Dilute solutions of UCI, decompose into U and Cl, at inert electrodes upon passage of current, Since the measured E’s are somewhat less than the theoretical estimate of 1.99 v, it is believed that either some depolarization occurs or thot the estimate is too high, or that both ore responsible, Concentrated ‘UCl, solutions may undergo the fol- lowing electrochemical transformations at platinum- graphite electrodes: () 2UCl, - 2UCl, + Ci,, (2) ucl, - (3 sucl, E°= 118 v, ! U+ 201, E° - 1.9y, 1t 4UCl; + U, E° = Lév, Uranium metal is deposited on the cathode; so the first reaction does not occur alone, If this process takes place in conjunction with the transformation (4 4ucl, - 3ucl, + U, E° - 0.85y, a continuous depletion in the UCI, concentration, which is low initially, would result (more UCI, would be oxidized to UCI, than would be replaced at the cathode). With low concentrations of UCl, the reaction (5) 2UC13 = 2U + 3CI2 , E°® = 2.2 v, has been found to predominate. The high value of E® for reaction 2 eliminates its consideration at these voltages. The decomposition potential of reaction 3 is high also but may be cccounted for by an activity ratio of (UC-iS)/(UCld)S/“ that is approximately equal to 1073 if the estimated AF®'s of —191 and -183 kcal for UCI, and UCI, re- spectively, are correct. The reaction of dilute and concentrated UCI, solutions ond nickel anodes may be attributed fo 2UCl, + Ni = 2UCly + NiCl,, E°= 0.36v. Dilute UCl, solutions seem to show signs of re- duction in a hydrogen atmosphere. Concentrated ond dilute FeCl, solutions also undergo different reactions at inert electrodes, In concentrated solutions FeCl, is oxidized to FeCl, at the anode (E° = 0.35 v), whereas in dilute solu- tions decomposition into the elements occurs (E° = L11 v). With nickel anodes, both the fol- fowing reactions probably occur simultaneously: FeCl, + Ni = NiCi, + Fe, E° = 0.30 v, 69 ANP QUARTERLY PROGRESS REPORT and 3FeC|2 = 2FeCl,; + Fe, % = 0.35 v. The NiCl, solutions, as far as is known, undergo the same electrochemical reactions regardless of concentration. Thus the changes in decomposition potential with concentration may be assigned to differences in NiCl, activity alone, With the aid of the Nernst equation, the NiCl, activity at 33.3 mole % NiCl, may be computed to be 100 times that in the 3.3 mole % solution and 300 times that in the 1 mole % solution. Solubility of Xeaon in Molten Salts R. F. Newton, Research Director’'s Department D. G. Hill, Consultant The use of molten salts as reactor fuels offers the possibility of removing the gaseous fission- product poisons, of which zenon is the most im- portant. {f, for exomiple, the xenon concentration of the 60-Mw CFRE can be held at 10% of its equi- librium value (1.1 x 1078 mole of xencn per milli- liter of fuel), important perturbations in reactivity can be avoided. In on circraft reactor, however, it will be necessary to effect this removal of xenon with an absclute minimum of auxiliary eguipment. The available literature appears to contfain no measurements of the solubility of gases in molten salts. Accordingly, it haos seemed desirable to evaluate the solubility of xenon in molten fluorides over the temperature range of interest, Preliminary tests indicate that the following experimental tech- nique is satisfactory. The apparatus, which is constiucted of nicke! where it will be in contact with the fluoride melt and of glass elsewhere, consists of two sections that can be isolated by freezing a plug of fluoride in the U-tube connecting them, the apparatus the melt is allowed to saturate with In one section of xenon under predetermined conditions of tempera- ture and pressure. Melting of the fluoride plug permits the saturated sali to flow into the other section, while the U-tube seal prevents entrance of goseous xencn into this section; refreezing of the plug isolates the sections again. section, xenon is siripped from the salt by repeated circulation of the hydrogen contained in the appa- ratus through the solt and past a liquid-nitrogen- cooled irap. After the stripping process is com- plete, the hydrogen is removed, the xenon is allowed In the second to come to room ftemperature, and the amount col- 70 lected is ascertained by measurements in a modi- fied Mcleod gage. Preliminary experiments at 600°C ond essentially 1 atm of xenon indicate that the solubility in the NaF-KF-LiF eutectic is not greater than 107 mole of xenon per milliliter of salt, The method and technique proved to be feasible, but minor experi- mental difficulties necessitated modification of the nickel apparatus. More accurate data under these and other conditions will be available in the neor future, Meanwhile, in order to gain familiarity with the technique, the solubility of xenon in the KNO,- NGNOB eutectic (66 mole % NQNO3) has been measured in an apparaius consiructed entirely of glass. The solubility of xenon in this melt is about 8.5 x 1078 mole/ml at 280°C and 10-7 mole/ml at 360°C. This liquid apparently shows the positive temperature coefficient of solubility expected for gases in molten salts, X-Ray Diffroction Studies in Salt Systems P. A. Agron M. A, Bredig Chemistry Division Cesium Halides. [t has been suggested?1:2? that the increase in volume on melting of simple binary salts is indicative of structural changes in the melt, that is, of a decrease to a lower ionic coordination. Volume increases up to 30% have been indicated for the atkali halides. The x-ray data?® oa the thermal expansion of solid CsBr and Csl gave volume changes of approximaiely 26%, but on extrapolation of about 50°C to their melting points was required for the computation of these changes, It was there- fore cof interest to ascertain whether o solid-phase transition analogous to the one known to exist in CsCi occurs in the unexplored temperature interval. This would replace the assumed structural change in the melting process by one in the solid state, High-purity CsBr and Cs! were intimately mixed with fine nickel powder to provide an intemal standard whose thermal expansion is known accu- rately, 24 perature l.attice parameters as a function of tem- were obtained on the high-temperature 21y, w, Johnsan, M. A, Bredig, and W. J. Smith, Chem. Div. Quar. Prog. Rep. Dec. 31, 1952, ORNL-1482, p 32, 22[:"‘. A. Agron and M, A. Bredig, Chem, Semiann. Prog. Rep. June 20, 1954, QRML.-1755 (in press). 23, Wagner and L. Lippert, Z. physik. Chem. B3], 2563 (1934). 241 | Jordan and W. H. Swunger, J. Research Nat. Burn Standards 5, 1291 (19230). x-ray diffraction unit to within 5°C of the melting points, No transformation was observed in either salt to within 2 or 3°C of the fusion temperature, The volume changes for these solids from 25°C to the fusion point were 12% for CsBr and 11.2% for Csl. Comparison of the molar volumes of the solids with volumes extrapolated for these molten salts?3 down to their respective melting points shows that an increase of approximately 27% takes place on melting. These data indicate that a structural chonge, similar to the lowering of the ionic coordi- nation from 8 to 6 in solid CsCl, occurs on fusion. NMaF-ZeF,. It wos desirable to determine the magnitude of the volume change found in complex salts on melting. Nickel powder was admixed to o somple of a homogeneous phase having the com- position of NaF-ZrF4 (53-47 mole %). The lattice expansion was measured up to 491°C. The data are listed in Table 5.10. The volume expansion from 25°C to the melting point is 3.8%. The volume is on almost linear function of temperature., The thermal coefficient of volume expansion, «, is 7.75 x 1075 deg= !, Thus the molar volume of the solid at the melting point of 520°C gives a density of 4.00 g/cm?3, molecular weight of 1287.5 g, corresponding to the presence in the unit cell of 6.77 NaF 4+ 6 ZrF, molecules?® for this composition, On extrapolation The latter value is based on a 25t ierational Critical Tables, Yol. 3, p 24. 26p, A, Agron and M, A. Bredig, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 47. PERIOD ENDING SEPTEMBER 10, 1954 of the liquid density?7 of NaF-ZrF , (50-50 mole %) meosured ot above 600°C 1o 520°C, a value of 3.32 g/em® for the liquid is obtained, An adjust- ment of this to the composition of NaF-ZrF , (53-47 mole %), by assuming an almost constant molar volume, lowers the density to approximately 3.2 g/em?. Thus a relatively large density and volume change of 26% is indicated for this complex salt on melting. The volume increase may be considered as indicative of a decrease in ionic coordination in the fused state similor to that observed in the CsBr and Csl salts, The ionic coordination in the solid is assumed to be similar to that in NoUF ., for which Zachariasen proposed a ftluorite type of packing with equal coordinations for both the Na¥ and Zrit of the order of 8 or 9. It seems possible that the volume increase on melting results from a lowering in the coordination of the fluorines around Zr?*, which might also be considered as radical ion formation around zirconium; this point requires further study. Well-formed single crystals which originated from the central portion of a large melt of NaF-ZrF (53-47 wmole %) have become available, Single- crystal x-ray analysis has been initigted in an attempt to make a complete structure determination, especially of the positions of the fluoride ions, 273, 1. Cohen and T. N. Jones, A Summary of Density Measurements on Molten Fluoride Mixtures und a Corre- lation Useful in Predictin% Densities of Fluoride Mixtures of Known Compositions, ORNML-1702 (Moy 14, 1954), TABLE 5.10. THERMAL EXPANSION OF NuF-Zer (53-47 mole %} Hexagonal Cell O Temperature Dimensions (A) Molar Yolume Density™ Run No. o 3 3 Q) (cn™) {g/em”) a c -1 25 13,79 9.407 309.690 4.158 -4 190 13.84 9.453 313.38 4.108 -2 250 13.87 9.475 315.62 4.079 -7 278 13.89 9.482 316.64 4.066 -3 378 13.90 9.497 317.90 4,050 -6 395 13.90 3.504 318.14 4.047 -8 491 13.95 ®.525 321.18 4.009 Extrapolation 520 321.50 4.005 cell at this composition. *The density calculation is based on a unit cell containing 8.77 NaF and 6 ZrF , molecules per unit rhombohedral 71 ANP QUARTERLY PROGRESS REPORT Physical Chemistry28 E. R, Yan Artsdalen Chemistry Division The heat of combustion and the heat of formation have been determined for boron carbide, The com- bustion process in a high-pressure bomb is both incomplete and nonstoichiometric; it yields some free carbon, but the amount depends upon certain However, it is possible to obtain a fairly reliable value for the heat of the reaction extraneous conditions., B,C +40,—> 2B,0, + CO, by allowing for incomplete combustion and non- stoichiometry as determined by direct analysis of products. The heat of combustion is s‘\Hgga.] - ~683.5 + 2.2 kcal/mole., From this value and other established thermal data, it is found that the heat of formation is AHS. |, = —14.1 + 2.7 keal/mole and that the free energy of formation is AFJog 4. = -13.9 kcal/mole. The low-temperature heat capacity of molybdic oxide, MoOg, has been measured in the range from 15 to 300°K, and the results are in fair agreement with the work of Seltz, Dunkerley, and DeWitt, 29 who measured the heat capacity above 60°K, How- ever, low-temperature extrapolations by these authors according to the Debye T3 jaw are in error. Molybdic oxide has a layer structure, and its heat capacity between 15 and 60°K varies as T2 in the manner found for certain other crystals with layer |attice structure, such as boron nitride.3% The following values were obtained for MO, at 25°C: C% 29g.16 = 17.93 cal/molesdeg and S$9¢g 14 = 18.58 eu. Similar studies are in progress wit molybdenum disulfide, MoS,, another layer com- pound, Electrical conductivity and density measurements are nearly complete for the molten salt system, potassium chloride—-potassium iedide, which, in a number of respects, has been observed to resemble systems discussed previously, 3! 28D)atails of this work will be published in separate reparts and articles from the Chemistry Division. See also Chem. Semiann. Prog. Rep. June 20, 1954, QORNL.- 1755 (in press). 29y, Seltz, F. H. Dunkerley, and &. J. DeWitt, J. Am. Chem. Soc. 65, 600 (1943), 30A, S. Dworkin, D. J. Sasmor, and E. R, VYan Asrtsdalen, J. Chem. Phys. 21, 954 (1953), 22, 837 (1954), and Chem. Semiann, Prog. Rep. June 20, 1953, ORNL-1587, p 19. 72 PRODUCTION OF PURIFIED MOLTEN FLUORIDES F. F. Blankenship G. J. Nessle G, M. Watson Materials Chemistry Division Use of Zirconivm Metal as a Scavenging Agent C. M. Blood H. A, Friedman F. P, Boody F. W, Miles G. M. Watson Materials Chemisiry Division The most important contaminants of fluoride melts (HF, Fef,, and NiF,) can be removed by treatment of the melt with hydrogen, but this proc- ess requires long periods for nearly complete re- moval of the FeF,. Attempts have therefore been made to demonstrate the effectiveness of zirconium metal in removing these impurities in a short time, In NaF.Z¢F, Mixtures. Known concentrations of contaminants were added to 3.5-kg batches of pre- viously purified NaF-ZrF, (53-47 mole %) in nickel containers and allowed to remain overnight at 800°C in contact with a considerable excess of metallic zirconium (30 g); some stirring was assurad by continuous sparging with helium, No experi- mental difficulties were observed either in equili- bration or in filtration of the product, In one experiment, the 10 g of Fel", added gave an initial contaminant concentration of 1700 ppm Fe. After treatment, the contaminants present in the filtrate were 70 ppm Fe, 15 ppm Ni, and 15 ppm Cr. In a second experiment, 9 g of CrF2 was added to give an initial contaminant concentration of 1650 ppm Cr. The contaminants found in the filtrate in this case were 90 ppm Fe, 25 ppm Ni, and 15 ppm Cr. Thus these experiments indicate that the zir- conium metal addition is on effective means of preparing pure NaF-ZrF , mixtures, In NoF-KF-iLiF Mixtures. The purity of several 2-kg batches of NaF-KF-LiF eutectic after treat- ment with hydrogen and with zirconium metal ot 800°C is given in Table 5.11. It is evident thai scavenging with metallic zirconium can be sub- stituted for most, if not all, of the time-consuming hydrogen stripping process. Since the impurities in fluoride melts are initially present in low concen- trations, it is anficipated that the small amount of 31g. R, Van Astsdalen, ANP Quar. Prog. Rep. Dec, 10, 1953, ORNL.-1649, p 58; ANP Quar. Prog. Rep. June 10, 1954, ORNL -1729, p 57. PERIOD ENDING SEPTEMBER 10, 1954 TABLE 5.11. COMPARISON OF HYDROGEN AND ZIRCONIUM METAL FOR REMOVING REDUCIBLE IMPURITIES FROM NaF-KF-LiF MIXTURES Zirconium Hydrogen Used Contaminants Found in Product (livers) Metal Added —— : _ tx) r (wt %) Fe (ppm) Ni (ppm) Cr (ppm)- 500 None 0.0 2100 7 25 600 None 0.0 725 6 30 300 57 1.3 70 8 20 None 30 1.4 65 9 20 zirconium introduced will have o negligible effect on the physical properties of the melt, in these experiments at 800°C some alkali metal However, was detected in cold regions of the equipment and some attack on the gaskets and flanges was ob- served. If large-scale purifications were attempted, it would probably be necessary fo use a lower tem- perature for this treatment, ' Purification of NaF-ZrF , Mixtures by Electrolysis C. M. Blood H., A, Friedman F. P. Boody F. W Miles G. M. Watson Materials Chemistry Division In the effort to obtain improved fuel purification procedures it was shown that electrolysis in NakF- ZrF, melts ot 800°C with graphite anodes and nickel cathodes reduces dissolved iron and nickel to relatively low concentration in a much shorter time than is required for the customary reduction with hydrogen. Furthermore, the reduced impurities were collected on a removable cathode that could easily be withdrawn from pots with welded lids. This feature promises to be of marked advantage in processing large batches in equipment of a current design which ollows the reduced metal from each batch to recontaminate subsequent batches and thereby multiply the amount of reduction required. Careful attention was given to current-voltage curves at all stages during the electrolysis in the hope that “‘brecks'’ and apparent decomposition potentials would permit the progress of the reduc- tion to be followed. Filtered samples were removed during various stages and anq]yzed for iron and nickel to determine the actual amount present, To facilitate analysis of the current-voltage curves, the electrolysis pot waos fitted with “floating”’ anode and cathode probes, which were used to determine the potential between the melt and the ‘‘working’’ anode or cathode. The probes were made of I/8-in. nickel rod; their potentials with respect to each other and to both electrodes were followed through many cycles of increasing and decreasing applied voltage, but the results have not been completely interpreted, ' : J. A, Mcl.aren of the Chemical Technol‘ogy Division is of the opinion that measurements with the reference electrodes have shown that the depo- sition of iron takes place with very little over- voltage, At a current of 1 amp, iron was deposited with a difference of potential between the cathode and the nickel reference electrode of only 15 mv. After depletion of the iron, the cathode potential increased to 75 mv, The anode potential was found to have a marked discontinuity in the graph of current density and voltage, At some critical current density the anode voltage would increase suddenly. At the beginning of the electrolysis the jump was found only at high current densities, but as electrolysis continued, it was found at lower current densities. The anode jump was on the order of 0.2 to 2 v, an order of magnitude greater than the increase found on the cathode. All the electrolyses were carried out in NaF- ZrF, (47-53 mole %) mixtures in nickel containers, The electrodes were suspended from the lid of a 3-kg capacity purification readtor (4 x 17 in.) with modified spark plugs as insulating connectors; the electrodes were immersed fo a depth of about 4 in. The anode was o I/2-in.-dim C-18 graphite bar threaded to a short length of nickel rod which was welded to the spork plug. supplied from a Dresser Electric Co. 24-v 25-amp selenivm rectifier; measurements were made with the use of suitable potfentiometers, voltmeters, Direct current was 73 ANP QUARTERLY PROGRESS REPORT ammeters, and a recording ampere-hour meter. In the first experiments the nickel container was used as the cathode, and helium gas was continu- ously bubbled through the melt in order to stir and to remove gaseous electrode products., The melt was first thoroughly purified by a conventional HF treatment followed by reduction with hydrogen. Then current-voltage curves were token before and after the addition of 1980 ppm Fe*" as FeF,. The effect of the added impurity was readily noticeable. The purified melt gave a curve like plot D of Fig. 5.9, while the FeF ,-contaminated melt yielded a curve similar to plot A, Extrapolation of the curve obtained from the FeF ,-bearing melt gave « decomposition potential of 0.9 v. An elecirolyzing current of 5 amp was employed, but numerous current-voltage curves were also obtained so that at the end of 5 hr, 20.9 amp-hr had passed. This was approximately twice the theo- retical ampere-hours required to reduce the Fe'?; a sample removed at this time analyzed 50 ppm Fe and demonstrated that a fairly complete reduction had been achieved with at least 50% current effi- ciency. No evolution of fluorine was noted, nor was there an appreciable attack on the nickel con- tainer. The wall which had been in contact with the vapor appeared to be covered with a fine dust, presumably Nif . FFor the next series of experiments o new batch of freshly purified melt was used, hydrogen was substituted for helium, and a current of 10 omp was passed. Three purifications were carried out in succession with addition of FeF, and NiF, to simulate impurities, as shown in the first part of Table 5.12. Apparent current efficiencies based on data from Table 5.12 are deceptive because some of the reduction is accomplished by hydrogen, which gives high values, and also becouse the current efficiency for purification becomes very low when the amount of impurity to be reduced is very small, When the container was used as a cathode, 6 v was the maximum applied potential required to obtain a current of 10 amp. An HF concentration in the effluent gas of about 1073 mole of HF per liter of hydrogen was noted at 5 amp. In another series of experiments a removable cathode, insulated from the pot, was used. The cathode was a %8 x 4 in, cylinder of nickel gauze welded to a .I/B-in, nickel rod; the gavze was made of 15-mil wire, 25 wires to the inch. The dimen- sions were chosen to be in the correct proportion for a small-scale model of a 250-1b capacity pro- duction reactor. To obtain a relatively high level of initial impurity, approximately 20 g of FeF, and 20 g of NiF , were mixed with the starting materials. After the usual HF treatment before electrolysis, a sample was removed to determine the iron con- centration, The results of the trial are summarized in the second part of Table 5.12. At the end of 12.5 amp-hr the cathode was re- moved; it had gained 10.3 g of metallic iron (as computed from chemical analysis), representing a 97% yield, A new cathode was then used until a total of 37 amp-hr had passed; this cathode gained 0.2 g of iron, and thus the percentage recovery had been increased to 99%. A current of 10 amp gave current densities of 117 amp/cm? of anode surface. By doubling the nominal area of gauze used as the cathode to obtain the total surface area both inside and out, the current density at the TABLE 512, ELECTROLYTIC PURIFICATION OF NaF-ZrF ; MELTS WITH HYDROGEN STIRRING bimpurities Before Impurities After Cathode Electrolysis {ppm) Ampere-Hours Electrolysis (ppm) Arrangement Passed Fe Ni Fe Ni Nickel container 1980 2020 15.2 95 55 as cathode 1890 2000 23.6 Not sampled 1980 Necne 11.8 100 40 Removable nickel 4660 * 12.5 150 gauze cathode * % 37.0 8qQ *Unfortunately, a reliable analysis of the nicke! concentration was not obtained. **Continued without additional impurity, 74 0.0 | : PLOT A 8.0 | AMP-HR PASSED = PN : i : ! o DOWN | i ! ' } Li * | PLOT 8 ; | | . : PLOT G | ‘ 4!‘AMP—HR PASSED: AMP -HR PASSED: | o 1.7 (DOWN}) o 200 (BOWN) GURRENT (amp) UNCLASSIFIED QRNL-LR—DWG 2929 I I PLOT D AMP-HR PASSED: P 36.5 (DOWN) 8O — L a 122 {UP) L i az0R2MUP b LaFTO WP - & up N 5 : ' ‘ 5 ? : § ; ‘ § | : : ; ] | é 5 | 5 | L } : - — <~——y—— R i i i i : T \ | ‘ : P : : ' | L E ‘ L SN | E | \\*BREAK {UP): 3.5 amp | P \ | ‘ § i : e : : —— - | : | | i | —BREAK [UP): 1.4 amp | 74 ‘ ; i : i | o BReAK {DOWN): 0.6 amp T BREAK (DOWNI: 0.6 amp ‘ 4 20 1D 20 30 40 5C 80 70 20 3.0 40 50 50 APPLIED POTENTIAL {voits) 10 APPUIED POTENTIAL fvolts 20 i [ ~~BREAK {DOWN): G.4 amp = | 30 40 50 80 Fig. 5.9. Plots of Current vs Yoltage in Purification by Electrolysis of NaF-ZrF, Mixtures. pGCéL O d39WILdIS DONIONT GOIYId ANP QUARTERLY PROGRESS REPORT cathode was found to be about 8 amp/cm? of nominal gauze area, In all runs under hydrogen, the effluent gas was bubbled through KOM solution to remove HF. A sample of the gas after removal of HF was analyzed by mass spectrometry and found to contain 3.17% CH,, 1.25% H,0 (from the KOH solution), 0.73% N,, and only about 0.02% CF,. Evidently the predominant anode product is HF, The anodes disoppeared at o rate corresponding to a life ex- pectancy longer than 50 amp-hr. The cathode de- posits were expected to contain zirconium metal, and this was qualitafively confirmed by the high zirconium analysis (53.9 wt % for the cathode that gained 0.2 g of iron) in the scrapings of mixed metal and salt from the cathode. However, the zirconium metal could not be distinguished by x-ray diffraction, | ’x ‘ i ; | | 6.0 — v T t | INITIAL CONTAMINANT S0 LEVELS: o - T & Fe:= 4660 ppm /-FILTRATE = 3 Ni= &5ppmn 2 i X | | < 40 o o . o . L m II ; L ‘: O T 3 ! - 3.0 — --:UF:4 77 ANP QUARTERLY PROGRESS REPORT T TABLE 5.13. COMPLETENESS O Initial Melt Compasition {mole %) ZF, UF, NeF LiF KF Time 4 4 (hr) 63.7 363 4 47.0 53.0 5 47.0 53.0 5 45.9 40 501 3 459 40 507 3 39.9 6.5 53.6 4 39.9 6.5 53.6 4 4.5 95.5 50 4.5 95.5 26 4.5 10.6 43.8 40.9 3 A5 10.8 438 40.9 28 45 10,8 43.8 40.7 52 A5 10,8 43.8 40.9 28 A5 10.8 438 AD.9 2 5.3 0.7 435 405 28 .8 1.1 445 426 24 1.1 44.5 424 24 1.8 Equilibration #F REDUCTION OF JF 4 IN VARIOUS FLUORIDE MEL TS Eeducing Agent Ratio of U3+ to *Per cent of stoichiemetric quaniity -~ amount raquirsd to saiisfy: hue s Jur, = UFyor Y 2® 1 UF, = Uk, + ) ratios. All were vich in red crystals (no green ot all} dencted os “KF-UF3 complex compound.” No tetrovalent possibla quanfitiss) could be distinguiskad in this system, although tetravalent uranivim was found in both the | e & RELIT] uranivim {(othar frace Lit and the Zrf ,-bearing systems where the re- duction was vastly more complete, These results are guite surprising, since pelregraphic examination has generally been guite sensitive in oscertaining small amounts of UF, or U, in other systems. Additional studies in these and similar systems will be mads, (%*) Total U Temffg’;me W Ze° Expected Found 800 0 128 1.0 n.88 800 100 ¢ 1.0 0.89 800 100 0 1.0 0.88 800 0 25 (.25 0.14 800 0 25 0.25 .17 800 0 15 0.15 0.19 800 0 15 0.15 0.12 845 125 0 1.0 0.96 850 200 0 1.0 0.91 780 106 15 1.0 0.41 780 100 15 1.0 0.48 780 100 15 1.0 0.44 600 100 0 1.0 0.59 600 200 0 1.0 0.76 600 50 0 0.5 6.33 800 91.8 0 0.93& 0.30 860 921.8 0 0.9:2 0.42 ZrF Al cliMets! Fluoride Processing Fucility C. R. Croft 1. k. Eorgan J. P. Bickely J Truist Maoterials Chamistry Division The alkali-matal fluoride processing facility com- pletad May 15, 1954 is used chicfly for the develop- ment of suitable processes for the production of alkali-metal fluoride mixivies containing UF,, There have beon severa! preparstions, however, of special batches of other compositions to fulfill specific needa. Single batches of approximately 2.5 kg caon be prepared in this equipment, During the quarter, 27 preparations that yielded about 70 kg of material were maode. Repeated attempts to prepare NaF-KF-LiF mix- tures containing 14 wt % UF,.ond 1 wt % UF, by reduction of UF, with uranium metal yielded mix- tures containing 3 fo 7 wt % UF, by chemical analysis. These results, along with those reported elsewhere in this document, seem to show that in systems containing KF, reduction of UF, by urc- nium metal is markedly incomplete. Pending clari- fication of this point, production of material con- taining 3 wt % UF ; and 10 wt % UF is under way, Production Facility J. P, Blakely F. .. Daley Materials Chemistry Division During the past quarter, 1760 kg of processed fluorides was produced in the 250-Ib facility for external and internal distribution, A breckdown of the production according to composition is given below: Amount Processed {kg) NuF-Zer-U F4 (50-46-4 mole %) 721 NoF-ZrF , (50-50 mole %) 565 NaF-ZrF ,-UF , (50-43.5-6.5 mole %) 404 Pratt & Whitney Aircraft Division received 27 kg of NaF-ZrF, (50-50 mole %) and 113 kg of NaF- LrF ,-UF , (50-46-4 mole %), Battelle Memorial Institute received 46 kg of NaF-ZrF ,-UF, (50-46-4 mole %) and 68 kg of NaF-ZrF, (50-50 mole %). The remaining processed material was distributed to various requesters in the ANP Program, : The difficulties associated with the long stripping times ond large hydrogen volumes required for pro- duction of rigorously pure fluoride melts from the raw materials available have not yet been over- come.3? Attempts to shorten the time required by increasing the hydrogen flow rate from 4 to 15 liters/min ond thus decrease labor and maintenance costs were not successful, Under these conditions 320 R Blankenship and G. J. Nessle, ANP Quar. Prog, Rep. June 10, 1954, ORNL-1729, p 61. PERIOD ENDING SEPTEMBER 10, 1954 the gas inlet tubes plugged quite frequently and the HF concentration in the exit hydrogen feli much below the value obtained at lower flow rates. The over-all purification time was not appreciably shortened by the fourfold ‘increase in hydrogen passed. Sublimation of crude ZrF , at a temperature lower than that previously used by Y-12 personnel did not appreciably improve the purity of the ZrF, product. Accordingly, the NaF-ZrF , melts produced future will be prepared from the (NaF) .ZcF, that is available in moderately pure form from o commercial source; hafnium-free ZeF, will be used to adjust the composition as required, since this material is available in a high state of in the near purity. Electrolytic purification of NeF-ZrF, mixtures has been shown to be complete, and the process requires much less time than does the hydrogen stripping process. Accordingly, one of the units in the production facility is being modified slightly to test this method on a large scale. It is antici- pated that the use of the purer raw materials and the electrolytic purification process will consider- ably decrease the cost of fuel preparation. in-Pile Loop Loading J. E. Eorgan Materials Chemistry Division The first in-pile loop was loaded on June 11 with a fuel concenirate, NaF-ZrF ,-UF, (62.5-12.5-25.0 mole %), prepared from enriched uranium by Y-12 personnel. The looding apparatus and controls performed satisfactorily, and the transfer of ma- terial to the loop went smoothly. However, the electrical contact which should have indicated when the loop was filled to operating level was not octivated even after 120% of the caiculated charge had been added. _ | Subsequent examination of the loop showed that a weld had given way where the loop was connected to the pump bowl, and it seems certain that this leak was present during the filling operation. The U233 j5 being salvaged from the loop. Examination of the filling apparatus revealed that all but 68 g of a total of 5837 g of the material was transferred to the loop, 79 ANP QUARTERLY PROGRESS REPORT CHEMISTRY OF ALKALI-METAL HYDROXIDES L. G. Overholser . Kertesz Materials Chemistry Division Purificotion E. E. Ketchen Materials Chemisiry Division Eighteen batches of NoOH were purified by filter- ing the 50 wt % aqueous solution through a fine sintered-glass filter to remove the Na,CO; and by dehydration at 400°C under vacuum. Approximately one-half the runs yielded products containing less thon 0.1 wt % H,O and 0.1 wt % Na,CO;. The remaining runs yielded less than 0.1 wt % H,0, but the values for Na,CO; ranged from0.11 to 0.2 wt %, Four 0.5-1b portions of potassium were reacted with water, and the resulting solutions of KOH were dehydrated ot 400°C under vacuum. The resulting material contained less than 0.1 wt % K,CO,, 0.1 wt % H,0, and approximately 0.1 wt % Na, Reaction of Sedium Hydroxide with Metals H. J. Buttram . A, Knox F. Kertesz Materials Chemistry Division Work has been continued on the reaction of copper and sodium hydroxide. Data were previously re- ported33 on copper concentrations found in NoOH up to 800°C, A recent determination at 900°C shows 2500 ppm Cu in the melt after exposure. This value is higher than would be expected from the previous data taken from 600 to 800°C, A plot of log Cu concentration vs 1/T yields a siraight line from 600 to BO0O°C and thus suggests that true equilibrium values were obtained., Extension of this line would give a value of about 1500 ppm rather than the 2500 ppm Cu obtained, 3B, A Knox, H. J. Buitram, F. Kertesz, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 60, 80 Reactions of copper and nickel with sodium hy- droxide were previously investigated by determining the equilibrium hydrogen pressures developed. Monel, an alloy composed primarily of copper and nickel, has now been investigated by this method. The Mone!l reaction tube containing the melt was surrounded by a quartz envelope, to which was attached hydrogen pressure. a manometer for determination of the Establishment of equilibrium required long periods of time, and the values ob- tained were not completely reproducible. Therefore, only a range of equilibrium pressures is presented in the following for the sodium hydroxide-—-Monel reaction: Tempearature (°C) Pressure (mm Hg) 600 35 to 42 700 112 to 130 800 160 to 193 900 232 to 280 1000 280 to 360 In another series of tests nickel oxide (reagent- grade commercial heated to 700°C in vacuum) was added to sodium hydroxide in a jacketed nickel tube, At 350°C a pressure of 22 mm Hg was found, Mass spectrographic analysis of the gas revealed 5.6% hydrogen and the remainder to be essentially water vapor, The temperature of the test was increased to 500°C and then, in 100°C increments, to 1000°C, The pressure in no case exceeded 31 mm Hg, which is approximately the saturated ambient water~vapor pressure, Analyses of the gas showed it to be 97 to 98% water vapor, material with a little hydrogen. Under otherwise identical test conditions, no water vapor was found when nickel oxide was omitted. The net effect of the nicke! oxide is apparently that of oxidizing the hydrogen formed from the Ni-NaOH reaction to water vapor. PERIOD ENDING SEPTEMBER 10, 1954 6. CORRGSION RESEARCH W. D. Manly G. M. Adamson Metallurgy Division W. R. Grimes F. Kertesz Materials Chemistry Division The static and seesaw corrosion testing facilities were used for further studies of brazing olloys, special Stellite heats, Hastelloy R, Inconel, graph- ite, and various ceramics in sodium, fluoride fuel mixtures, and other mediums. For the fabrication of many reactor components it may be well to have a brazing alloy that is compatible both with sodium and with the fluoride fuels. Therefore several cor- rosion tests were performed on Inconel joints brazed with high-temperature brazing alloys. From this work it was found that the brazing alloy 67% Ni~13% Ge-11% Cr—-6% Si-2% Fe—1% Mn has good cor- rosion resistance in fluoride fuels and fair corrosion resistance in sodium. Two special heats of Stellite were tested that had a lower carbon content than those tested previously, and it was found that the fower carbon content increased the corrosion resistance in the fluoride fuels. ' High-density graphite was tested in sodium and in fluoride mixtures; it has very poor resistance to sodium attack but seems to have fair resistance to the fluoride fuels. Various hard-facing materials, ceramics and cermets, have been investigated from the corrosion standpoint for use as bearings, seals, and valve seats; Al,0,, MgAl,O,, ZrO,, and SiC were fested in sodium, a fluoride fuel, lithiom, and lead. In these tests, SiC showed the best corrosion resistance. in the thermal-convection loop corrosion studies, emphasis has shifted to fuels with the uranium present os UF3 and to loops of Hastelloy B, Hot- leg attack is not found in Inconel loops in which ZrF ,-base fluoride mixtures with the uranium as UF, are circulated. A deposit is, however, found on the hot-leg surface. Only preliminary information is available, but it appears that neither attack nor a hot-leg layer is found with alkali-metal-base fluo- ride mixtures containing UF,. Mixtures of UF, and UF, will result in a reduction in attack from that found with only UF ,, but some attack is present and in high-uranium-content systems it may be significant, The addition of zirconium hydride to the ZrF - base fuel mixtures containing UF4 was found to result in the formation of UF, and in a reduction in corrosion. However, to obtain a complete ab- sence of corrosion, sufficient hydride must be added fo cause a loss of uranium, Several Hastelloy B loops have now been suc- cessfully operated in both the as-received and the over-aged conditions. inboth cases, a considerable increase in hardness occurs during operation, With ZcF ,-bose ‘mixtures containing UF,, very little attack is found, even after 1000 hr. The mass transfer characteristics of type 316 stainless steel in molten lithium were studied by using thermal-convection loops. After 1000 hr of loop operation, there was no sign of plug formation, and only a small amount of mass transfer was found in one of the loops. The investigation of the cor- rosion and mass transfer characteristics of materials in contact with liquid iead has shown that certain alloys possess a much greater resistonce to mass transfer than the pure components of those alloys. The probable reason for this resistance to mass transfer con be related to the formation of inter- metallic compounds in these alloys, This has been proved in the case of alloys of 45% Cr-55% Co, Ni-Mo alloys, and the Fe-Cr-base stainless steels, The flammability of liquid sodium glloys has been studied, and it has been found that only bis- muth and mercury have an effect on the flammability of sodium. It was found that the degree of reaction of Na-Bi and Na-Hg alloys with air was not signifi- cantly changed when the pressure was varied from 0.25 to 1 atm. An investigation has been started to determine the amount of hydrogen liberated when NaOH is heated to 900°C in an inert environment, The container problem has been solved by using a large single crystal of magnesium oxide. A thermo- dynamic study of the alkali-metal hydroxides has shown that hydrogen is important in stabilizing NaQH at temperatures above 800°C, This is the temperature range at which moss transfer becomes significant in Ni-NaOH systems. These computa- tions link the formation of hydrogen with the forma- tion of unsaturated oxygen ions, which may be necessary for moss transfer to occur. 81 ANP QUARTERLY PROGRESS REPORT A plausible explanation has been developed for the evidence of a maximum in the corrosion of In- conel by molten fluorides in static tests at 800 to 900°C and for both void formation and chromium concentration in the melt being less at 1000°C than at B00°C. Static tests of type 316 stainless steel in NaZrF ¢ and in NaF-ZrF ,-UF , substantiated the phenomena observed with Inconel. Additional corrosion experiments with fuel mix- tures containing simulated fission products have shown that the experimental procedure being fol- lowed is entirely unsatisfactory. A controlled- velocity corrosion testing apparatus has been con- structed for the rapid transfer of mclten fluoride mixtures through both heated and cooled test sec- tions. This apparatus was designed to simulate conditions in reactor components. STATIC AND SEESAW CORROSION TESTS E. E. Hoffman W. H. Cook C. R. Brooks C. F. Leitten Metallurgy Division Brazing Alloys in NaF-ZrF ,~UF ard in Sodium Specimens of the three brazing alloys described in Table 6.1 were tested in NoF-ZrF -UF, and in sodium. The specimens were cut from 3-in. T-joints of Inconel which hod been brazed in a dry hydrogen The purpose of these tests was to find a brazing alloy with good corrosion resistance to both NaF-ZrF ,-UF, and sodium, since such « brazing alloy could be used to back up tube-to- header welds, In use, the brazing alloy would be exposed to the fluoride mixture, but resistance to attack by sodium would also be required if o fracture occurred in a tube-to-header weld. atmosphere. TABLE 6.1. RESULTS OF STATIC TESTS OF BRAZING ALLOYS IN NaF-ZrF -UF (53.5-40-6,5 mole %) AND IN SODIUM AT 1500°F FOR 100 hr Brazing Alloy Weight Weight Composition Bath Change Change Metallographic Notes (wt %) (g) (%) Alloy A 67 Ni, 13 Ge, 11 Cr, 6 Si, 2 Fe, NaF-ZrF -UF, +0.0003 +0,024 Very little attack; slight surface 1 Mn on Inconel (brazed at roughening to a depth of 0.25 1120°C for 10 min) mil Sodium ~0.0003 -0.017 Attack in the form of subsurface voids to a maximum depth of 3 to 4 mils Alloy B 65 Ni, 25 Ge, 10 Cr on Incone! NoF—ZrF4-UF4 ~0.0010 -0.077 No attack; slight uniform solu- {(brazed at 1140°C for 10 min) tion would account for weight loss Sodium ~0.0013 -0.116 Braze fillet attacked tc a maxi- mum depth of 8 mils Alloy C 72 Ni, 18 Ge, 7 Cr, 3 P on NGF-ZrF4-UF4 ~0.0005 -0.041 Braze fillet attacked to a depth Inconel of 2 tc 3 mils in the form of subsurface voids Sodium -0.0008 -0.058 Braze fillet ottacked to a depth 82 of 3 to 6 mils in the form of subsurface voids Three brazed Inconel specimens were exposed for 100 hr at 1500°F in %-in.-OD, 0.035-in.-wall Inconel container tubes to static NaF-ZrF ,-UF, (53.5-40-6.5 mole %), and three were exposed to static sodium. The brazed specimens were cleaned carefully before and after testing so that valid weight-change data could be obtained. The speci- mens were nickel plated after exposure to prevent rounding of the edges during metallographic polish- ing. Presented in Table 6.1 are the data obtained by mefullogmbhic examination of the specimens, Alloys A and B showed very good resistance to attack by the fluoride mixture, as shown in Figs. 6.1 and 6.2. Alloy A is also shown in Figs. 6.3 and 6.4 at low magnification in the as-brazed and as-tested conditions. Alloy C was attacked slightly by the fluoride mixture. All the alloys were attacked slightly by sodium, with alloy A showing the least attack. The maximum depth of attack of 3 to 4 mils Ni PLATE PERIOD ENDING SEPTEMBER 10, 1954 on alloy A is not serious, since sodium would be in contact with this alloy enly in the event of a tube-to-header weld fracture. It may be concluded from the results of these tests that further study is warranted on alloys A and B; both showed good re- sistance to attack by the fluoride mixture. Special emphasis will be placed on alloy A, since it exhibited superior resistance to attack by sodium, - Special Stellite Heats in NoF-ZsF ,-UF, and in Sodium Two special heats of Stellite, heats B and C, were requested from the Union Carbide and Carbon Metals Research Laboeratories for corrosion testing. The specimens were sand cast and had Rockwell C hardnesses as follows: heat B, 39 to 42; heat C, 42 to 43. The as-received microstructure consisted of Cr,C, corbide particles in a cobalt-rich matrix. The approximate compositions of these heats were Cov-ie 0.004 > o o 2 0.002 0.003 T L. = Fig, 6.1. Brazing Alloy A (Ni-Ge-Cr-Si-Fe-Mn) Fillet Surface After Exposure for 100 hr ot 1500°F in NaF-ZrF -UF ; (53.5-40-6.5 meie %). Etched with oxalic acid. 83 ANP QUARTERLY PROGRESS REPORT Ni PLATE 0,004 1000X o o 02 INCH 0.003 Fig. 6.2. Brozing Alloy B (Ni-Ge-Cr) Fillet Surface After Exposure for 100 hr ot 1500°F in NoF-ZrF - UF, (53.5-40-6.5 mole %). Etched with oxalic acid. UNCLASSIFIED | 0.005 0.040 150X e 2 o 0.020 INCH © Q n o Fig. 6.3, Brazing Alloy A on Incona! Telsint in As-Brazed Condition. Etched with oxalic acid. Fig. 6.4. Etched with oxalic acid. as follows: Composition (wt %) Heat B Heat C Carbon 1 0.30 Chromium 35 28 Tungsten S Mo lybdenum 12 Manganese 0.50 0.50 Silicon 0.50 0.50 Nickel 3 Cobalt 58 55.7 The specimens were tested in static mediums for 100 hr at 1500°F. used and duplicate tests were run. are presented in Table 6.2. Both Stellite heats were heawly attacked along the carbide network in NaF-ZrF - UF, (50-46-4 mole %), as shown in Figs. 6.5 and 6.6. The attack varied from a minimum of 7 mils to a maximum of 20 mils. The depth of attack varied with the orien- tation of the carbide network with relation to the Inconel container tubes were The test results Brazing Alloy A on Inconel T-Joint After Testing. PERIOD ENDING SEPTEMBER 10, 1954 0.03 CENCH: 0.04 Dark areas are not areas of attack, specimen surface, being deepest when the nefwork was perpendicular to the surface. None of the specimens were attacked in Sodxum however, in heat B there were no large equiaxed carbide pamcles in a 1- to 2-mil area adjacent to the surface. The carbide particles in this area ap- peared to have transformed to a very fine structure. There was also a large amount of fine precipitate in the cobalt-rich matrix near the surface. 4 Heat C had slightly better resistance than heat B to attack by sodium, and both heats had very poor resistance to attack by the fluoride mixture. Several of the specimens had shrinkage veoids near the surface, one that was 7 mils in diameter, and the attack was deeper in these areas. | Hastelloy R in Various Media Hastelloy R centerless ground and annealed rods, ?,’Q in. in diameter and 8 in. in length, were obtained from the Haynes Stellite Company. Each rod was machined info a specimen (h X ]/4 X ?j‘ in.), a con- tainer, and o plug. The specimen and the medium to which it was to be exposed were loaded into the 85 ANP QUARTERLY PROGRESS REPORT TABLE 6.2 RESULTS OF STATIC TESTS OF SPECIAL STELLITE HEATS IN SODIUM AND IN NunZer-UF4 (50-46-4 mole %) AT 1500°F FOR 100 hr Weight Change Metcllegraphic Notes Bath (Q/i"-vz) e Heut B Heat C Heat B Heat C As received Structure consisted of large, fairly Carbide particles which sur- equiaxed particles of complex rounded the dendrites of the carbide (Cr7C3) surrounded by a cobalt-rich solid solution had a eutectic structure in a cobalt- very fine structure; only a few of rich matrix the eguiaxed carbids porticles seen throughout Heat B could be found Sodium +0.008 +0.002 No attack; hawever, no globular No attack; very little difference in particles of tha carbide phase in as-received and tested structures a 1- to 2-mil area adjacent to the surface; carbide particles in this area completely transformed to a very fine precipitate; precipite- tion also occurred in matrix, es- pecially near the exposed surface +0.016 +0.003 Same as above Same gs above Nc1l"'-Zrl:4-UF4 -0,052 —0.045 Attack on corbide phase varied Carbide network attacked to a from 8 to 19 mils; minimum at- depth of 7 to 20 mils, depending tack of 8 mils occurred on speci- on orientation of the network men where dendrites of cobalt- rich solid solution surrounded by carbide particles were oriented parallel to the specimen surface; maximum attack of 19 mils oc- curred where dendrites were oriented perpendicular to the specimen surface; large amount of fine precipitation in matrix during test due to aging ~0.060 -0.057 Same as above Same as above container, and the plug was welded in place in a dry box under a purified helivm atmosphere. The nominal compositien of Hastelloy R is 65% Ni, 14% Cr, 10% Fe, 6% Mo, 3% Ti, 2% Al. The specimens were exposed to the various media for 100 hr at 1500°F, The test results are presented in Table 5.3. The specimens showed good resistance to attack by I‘«lc:l':-ZrF“-UF‘fii (50-46-4 mole %), Fig. 6.7, and by lithium. Surprisingly, this alloy had better re- sistance to attack by lithium than by sodium; the attack by lithivm on high-nickel-content alloys (for example, Inconel) at 1500°F is usually heavy and is greater thon the attack by sodium. The sample tested in sodium was attacked slightly, while the specimen tested in lead was heavily attacked. The specimen tested in sodium hydroxide was attacked throughout its entire thickness. 86 Inzone! in Melten Rubidium Tests were conducted on rubidium metal in con- nection with an ORSORT project. Rubidium has a low melting point, 92°F, and a low boiling point, 1270°F, and therefore is being considered as a possible reactor coolant. The rubidium metal used in the experiments was prepared by calcium re- duction of the rubidium fluoride salt by the Stable Isotope Research and Production Division, An initial attempt to determine the vapor pressure vs temperature relationship up to 1650°C was un- successful. However, the test equipment has been redesigned and another attempt will be made in the near future, lnconel has been tested in contact with static rubidium for 100 hr ot 1650°F, The at- tack varied from 0.5 mil in the vapor zone to a maximum of 1.5 mils at the liquid-gas interface, as shown in Fig. 6.8. The rubidium used was not dis- tilled and is known to have contained several per PERIOD ENDING SEPTEMBER 10, f1954 Q.00 1000% | o.002]| 0.003 ING &—{ - 0.004 Fig. 6.5. Special Stellite Heat B Before {(a) and After (b) Exposure to Static NnF-ZrF +UF, (53.5-40- 6.5 mole %) for 100 hr ot 1500°F. The area where the attack terminated is shown in (b) Efched with KOH-K ,F e(CN) . & o.00t Y-12593 0.004 Fig. 6.6. Special Stellite Heat C Before (a) ond After (b) Exposure to Static NaF-ZrF +UF, (53.5- 40-6.5 mole %) for 100 hr at 1500°F. The area where the attack terminated is shown in (b). Efched with KOM-K FE(CN) " 87 ANP QUARTERLY PROGRESS REPORT TABLE 6.3. RESULTS OF STATIC TESTS OF HASTELLOY R IN VARIOUS MEDIA AT 1500°F FOR 100 hr Weight Weight Bath Change Change Metallographic Notes (%) (g/in.?) NuF-ZrF4-UF4 (50-46-4 mole %) ~0.05 -0.0037 Specimen unattacked except in one small area where voids extended to a depth of 1 mil; no attack on vapor zone of tube NaOH +7.9 +0.6088 Specimen lfd-ina square attacked throughout entire thickness Sodium ~0.06 -0.0021 Specimen and bath zone oitacked to a depth of 0.25 to 0.5 mil in the form of voids; exposed surface of bath zone of container partly de- carburized to a depth of 1 mil Lithium -0.07 ~0.0047 Small voids to a depth of 0.25 mil in specimen and bath zone of tube; surface partly decarburized by lithium Lead 2 e Y + ¥ & L g ........ Specimen heavily attacked to a maximum depth of 3 mils {(average 2 mils); bath zone attacked uni- formly to o depth of 1.5 mils; phese change prob- ably occurred in attacked areas; vapor zone unattacked E Y.12619 : 0.004 4G00X 0.002 o T e 0.003 Fig. 6.7. Hastelloy R Exposed to Static NaF-ZrF ,UF, (50-46-4 mole %) for 100 hr at 1500°F. Etched with aqua regia. 88 Ni Fig. 6 8 PERIOD ENDING SEPTEMBER 10, 1954 LUNCLASSIEIED | Y1293 10,001 0.6002 0.003 INCH 0.004 lnconel Exposed to Stuhc Rub:dlum for 100 hr at 1650°F. Note decarburization in cflacked ared. Specnmen nickel piufed after test to pro’rec’f edge. Etched w:th glyceria reg(cl. cent sodium and some oxygen contamination. Ad- ditional static tests are under way with triple- distilled rubidium, and an Inconel thermal convection loop is being operated with boiling rubidium, Fig. 6.9. This loop has now operated for several hundred hours without difficulty. ' Carhunzahon of Inconel by Sodium it is well known that sodium, in addition to de- carburizing metals, can, in some cases, carburize them if the carbon concentration in the sodium is sufficiently high.. Therefore, an attempt is being made to determine whether small additions of carbon - would prevent decarburization of Inconel specimens during long-time creep tests in contact with sodium at elevated temperatures. A-nickel containers are being used for static tests ‘so that the ratio of inconel surface area to sodium volume will be small. The maximum solubility of carbon in nickel ot 1500°F is approximately ;0.]%_, and therefore the carburization of the nickel containers in these tests is very slight. The A-nickel used for the containers was found by analysis to contain only 0.05% carbon. The ratio of the Inconel surface to the sodium volume in. the tests performed to date was 0.76. The Inconel specimens used were 0.049-in. sheet reduced to 0.015 in. by cold rolling, and they were annealed for 2 hr at 1650°F, The carbon additions (1, 5, and 10 wt %) were made to the sodium in the form of small lumps of reactor-grade graphite. : The nickel containers were loaded with the In- conel specimens, the grophite, ond the sodium in a dry box in a purified helium atmosphere ond sealed under vacuum. As shown in Table 6.4 and Fig. 6.10, all the specimens were very heavily carburized and extremely brittle after exposure for 100 hr at 1500°F. . Also, they were partly covered v{rith a 89 ANP QUARTERLY PROGRESS REPORT g ——— |57 OF ~g— BOILER SECTION HEATER | J=e— 1085°F 1549F ¥-12930 CONDENSER SECTION\ _.fi',"‘ 833°F —t— | [QUID LEVEL ——- e 203 OF Fig. 6.9. Inconel Thermal-Convection Loop for Circulating Boiling Rubidium. green surface film that was identified by x-ray analysis to be Cr,0,. The oxygen source for forme- tion of this film was, at first, thought to be the graphite which had not been degassed prior to the tests; however, a Cr,C; film was later found in other tests with degassed grophite and standard tests with no graphite addition. Therefore prepa- rations are being made for obtaining oxygen-free sodium for use in these tests. TE. E. Hoffman et al., ANP Quar, Prog. Rep. june 10, 1954, ORNL-1729, p 70. 90 Special Tar-Impregnated and Fired Graphite A special tar-impregnated and fired graphite, known commercially as Graph-i-tite, has been tested for corrosion resistance to sodium and NaF-ZrF - UF, (53.5-40-6.5 mole %). Also, a comparison was made of the carburization of an austenitic stainless steel by contact with Graph-j-tite and with C-18 graphite (reactor grade). The Graph-i-tite was fabricated, as described previously,! by tar-impreg- nating and firing graphite 16 times. The repeated tar impregnations and firings produce a high density PERIOD ENDING SEPTEMBER 10, 1954 UNCLASSIFIED ; $20 UNCLASSIFIED y-12121 %TUnCLASSIFIED - Y-12120 Fig. 6.10. Carburization of Inconel Exposed to Sodium with Carbon Additions for 100 hr ot 1500°F. {a) inconel specimen before test. (b) Specimen tested in sodium plus 1 wt % carbon. (<) Specimen tested in sodium plus 10 wt % carbon. Specimens nickel plated to protect edges. Etched with glyceria regia. 250X. and a “‘tough skin,”" which, it was hoped, would reduce penetration of various liquid mediums. | An open Graph-i-tite crucible containing sodium and a sealed Graph-i-tite crucible containing NaF- ZrF ,-UF were tested at 1500°F for 100 hr in vacusm. The sodium completely penetrated the 0.38-in.-thick wall of the crucible, as in the pre- vious tesf,] and caused it to crack end crumble. The only macroscopically visible sign of attack on the crucibie that contained the fluoride mixture was that the inner surface had a higher gloss (Fig. 6.11) after exposure. Metallographic examination of sections of the crucibie indicated irregular pene- tration from 0 to 5 mils deep. No attempt has been made to determine whether reaction products ac- companied the penetration. This attack may be ?1 ANP QUARTERLY PROGRESS REPORT TABLE 6.4, CARBURIZATION OF INCONEL BY EXPOSURE TO SODIUM CONTAINING CARBON ADDITIONS FOR 100 hr AT 1500°F Carbon content of Inconel specimen before test, 0.056 wt % Carbon Added Carbon Content™ . ] Specimen to Sodium of Specimen Weight Change** Before Test After Test (%) (wt %) (wt %) o 1 0.882 1.09 5 1.09 1.34 10 1.24 1.63 *Average of three analyses. **On all specimens part of wzight change was due to tormation of Cr203 film on surface. compared with the 0 to 2 mils attack by NaF-ZrF - UF4 (50-46-4 mole %) in the previous tests.! The carbon content of the fluoride mixture did not change significantly in either test, A comparison of the carburization of austenitic stainless stee! by tar-impregnated and fired graphite and by C-18 graphite (reactor grade) was obtained through static tests of cylinders of the same size of each type of graphite in type 304 stainless steel tubes containing equal quantities of sodium in vacuum. Tests for 100- and 500-hr periods ai 1500°C were made on each type of graphite. Type 304 stainless steel was chosen because it carbur- izes more readily than other austenitic stainless steels, The results of the tests are shown in Figs. 6.12 and 6.13. In the 100-hr test, the tar-impregnated and fired graphite carburized the type 304 stainless steel to a slightly greater extent than the C-18 graphite did, and in the 300-hr test, it clearly car- burized the steel to a greater extent. After the 500-hr tests of both types of graphite, steel in con- tact with the sodium showed small subsurface voids; the voids were pronounced in the steel tested with the tar-impregnated and fired graphite. The depth of carburization of the type 304 stainless steel above the level of the molten sodium covering the graphite cylinders was 1 mil in all the tests, except the 500-hr test with a tar-impregnated and fired graphite cylinder in which there was carbu- rization of the steel to a depth of 2 mils. 92 ¥-12414 CONFIDENTIAL Y-12453 (b} Fig. 6.11. After (B) Exposure to NuF-ZrF4-UF4 (53.5-40-6.5 mole %) for 100 hr ot 1500°F. At the end of the test the crucible was inverted so that the fluoride mixture would flow to the top. The frozen fluoride mixture con be seen above the shiny inner surface of the crucible. Graph-i-tite Crucible Before (a) and PERIOD ENDING SEPTEMBER 10, 1954 20 S O > 10°0 UNCLASSIFIED Y-12110 | UNCLASSIFIED Tyaiz2i08 Fig. 6.12. Carburization of Type 304 Stainless $teel Exposed to Graph-i-tite (a) and to C-18 Graphite (b) in Sodium for 100 hr at 1500°F in Vacuum. (a2) Depth of carburization was 18 mils (concentrated) to 32 mils (traceable). (b) Depth of carburization was 16 mils (concentrated) to 32 mils (traceable). Etched with glyceria regia, 2 o 5 UNCLASSIFIED | Y.12429 . o $0°0 z <3 T £6°0 200 o < = +0° Fig. 6.13. Carburization of Type 304 Stainless Steel Exposed to Graph-i-tite {a) and to C-18 Graphite (b) in Sodium for 500 hr at 1500°F in Yacvum. (@) Depth of carburization was 40 mils (concentrated) to 49+ mils (traceable through the tube wall). (b) Depth of carburization was 33 mils (concentrated) to 49+ mils (traceable through the tube wall). Etched with glyceria regia. 93 ANP QUARTERLY PROGRESS REPORT Ceromics in Various Mediums Static corrosion screening tests of three ceramic oxides and one carbide (Al,05, MgAl,0,, calcium stabilized Zroz, and SiC) were made in a vacuum atmosphere at 1500°F for 100 hr in each of four mediums: sodium, NaF-ZrF ,-UF (53.5-40-6.5 mole %}, lithium, and lead. A fourth ceramic oxide was tested under the same conditions in sodium and in NaF-Z¢F ,-UF, (53.5-40-6.5 mole %). The Al 0, and MgAl,O, test pieces were cut from single, synthesized crystals; the MgO test pieces were cleaved from a single, synthesized crystal. The nominal dimensions of the test pieces and the re- sults of the tests are given in Table 6.5, Powder x-ray exdaminations were made of the specimens that were the least corrosion resistant in each medium, and the results are summarized in Table 6.6. The SiC specimens were, in general, the most resistant to corrosion, and the Al,C; specimens were the next most resistant. These single-run screening tests ore qualitative, and the results should be used with caution. Larger specimens ore to be tested so that more accurate data can be obtained oand metallographic examinations and chemica!l onolyses can be made. TABLE 6.5. SUMMARY OF STATIC TESTS OF CERAMICS IN YARIOUS MEDIUMS AT 1500°F FOR 100 hr Remarks Tested in Sodium Thickness Weight Material Tested Change Change (%) (%) a - - A|203 0.5 0.9 sich -0.9 a MgAl,0, 0.0 ZrD2 (CoOb stabilized) +8.6 MgO® -0.5 ~0.1 Specimen changed from a colorless transparent state to a white semitransparent state; a small crack oppeared on an edge Specimen essentially unaltered; brighter and cleaner; small part of disk edge broken off Originally colorless transparent specimen almost com- pletely changed to a black state; chipped slightly on one edge Specimen changed from light buff to blue-black; =dge chipped Specimen unaltered except for color change from color- less to light blue-gray Tested in Nc'!F-ZrF"“-UF4 (53.5-40-6.5 mole %) A|203 -31 SiC - 1.8 Specimen broken and covered with fluoride mixture; brown-red interface present between the fluoride mix- ture and the specimen Specimen broken; recovered pieces appeared to be un- attacked Recovered portion of specimen {more than one-half) covered with fluoride mixture and cpparently com- pletely altered ZrCB2 {CaQ stabilized) MgO A1203 94 No visible trace of specimen found No visible trace of specimen found Tested in Lithium No visible trace of specimen found - PERIOD END!NG? SEPTEMBER 10, 1954 TABLE 6.5, (continued) Thickness Weight Remarks Material Tested Change Change - 4 (%) (%) SiC MgA|204 ZrC’2 (CaO sfubilized) Some small particles found Mo visible trace of specimen found Specimen in many small pieces; buff color changed to charcoal-black Tested in Lead A|203 50,0 +0.3 5iC | +2.9 MgA1204 : (0,0 Zr02 {Ca0 s}obi!ized) . 0.0 +0.5 No visible attack; colorless transparent specimen changed to gray; small fspot found on specimen; specimen ap- parently covered with thin lead film No visible attack; specimen broken in loading No visible attac:ik; specimen changed from colorless to gray; broken in foading : Mo visible utraék; specimen changed from a light buff f;:) gray ?Disk cut from a single crystal; specimen 0.75 in. in diameter and 0. 021 i, thick, bDISl‘( 0.75 in. in diameter and 0.021 in. thick. “Cleaved from a single crystal; specimen 0.02 x 0.34 x 0.36 in. TABLE 6,6. POWDER X-RAY IDENTIFICATION OF THE PHASES OF SOME OF THE LEAST CORROSION-RESISTANT SPECIMENS ' Powder X-Ray Data on Specimen Speci men : Te st Med iUm "‘"-u"mw""""‘-; ---------------------------------------------------------------------------------------------------------------- - . Befcre Test ' After Test MgAl, O, ~ Sodium : Face-centered cubic, a =7.970 Face-centered cubic, a = 7.981 NaF-ZrF,~UF, Face-cenrered cubie, a =7. 970 NaF-ZrF ,-U F4 pottern (53.5-40-6.5 mole %) : Zr02 (CaQ ' Sodium ' Primarily face-centered cubic ZrOz; Seme as before test stabilized) : secondarily monoclinic ZrO2 ‘ Lithium : Primarily face-centered cubic 2502; Unidenfified;. face-centered cubicfi:, secondarily monoclinic ZrQ, : a=4.674 FLUORIDE CORROSION OF INCONEL IN THERMAL-CONVECTION LOOPS G. M. Adamson Metallurgy Division Effect of UF, in ZrF“-{Base Fuels Preliminary results of Inconel thermal- convection loop tests of the corrosive properties of UF 4 bacrmg fluoride mixtures were presen’red prevnousiy and compared with the corrosive properties of UF .- bearing mixtures. Additional tests with UF ;-bearing ZrF ,-bose mixtures have confirmed the reduchon or ehmmct:on of hot-leg attack, along with the formation of a hot-leg layer. The lack of attack and the hot-leg layer are illustrated in Fig. 6.14. 20, M. Adaomsan, ANP Uuar Prog. Rep. June 10, 19)4 ORNL-1729, p 72 95 ANP QUARTERLY PROGRESS REPORT The portion of a hot leg shown in Fig. 6.14 was taken from an Incone! thermal-convection loop which had circulated UF3 (2.1 wt % UF, in 2.4 wt % total U) in NaF-ZrF4 (53-47 wmole %) for 500 hr at o hot-leg temperature of 1500°F, The fluoride mix- ture was drained from the loop af the operating temperature to ascertain whether the layers pre- viously noted had formed during cooling of the loop. Since the layer wos again present, it is now con- sidered to have formed during operation. This conclusion was further strengthened when the metal lographic examination revealed o diffusion zone hetween the loyer and the base metal. An examination of the layer by a microspark spectro- grophic technique® showed it to be predominantly zirconium. Since this technique is not sensitive to uranium, the layer could also contain uranium. There is some indication that some of the UF dissociates fo produce uranium metal, which re- duces some of the ZrF4 to produce zirconium metal, The accuracy of the analytical determination for zirconium is not sufficient o revea! the reaction products postulated, since some zirconium is also lost by sublimation. In another Incone! loop in which UF., in NaF- ZrF, was circulated for 2000 hr at 1500°F, some voids were found to a depth of 4 mils upon metalle- 3Examinufion made by C. Feldman, Chemistry Division. Fig. 6.14. Hot-Lzg Surfoce of lncone! Loop After Circuluting UF3 in MaF-ZrF , (2.1 v % UF, in 2.4 wt % total U) for 500 hr at 1500°F. Loop drained while hot. Unetched. 500X. Reduced 27%. 96 graphic examination. The voids did not appear to be the same as those normally found, and it is thought that they were formed when brittle inter- metallic compounds were pulled out of the base metal during polishing. No hot-leg attack could be found, as shown in Fig. 6.15, but there was a layer 0.5 mil thick on the hot-leg surface. This layer was also reported to be predominantly zirconium, It has been definitely established that the use of UF; rather than UF, in ZrF -base mixtures will lower the corrosive attack and mass transfer in Inconel systems. However, it has been found to be impossible to dissolve sufficient UF, in ZrF ,-base mixtures to obtain fuels of interest for high-temper- oture reactors. To obtein a fuel with sufficient vranium, it would be necessary to use a mixture of UF, and UF,. Daia obtained from loops oper- ated with such mixtures and with standard UF .- bearing mixtures are presented in Table 6.7. The data show clearly that a mixture of UF; and UF, produces less attack than UF , alone, but the mix- ture does not eliminate the attack, as is the case with UF 5 alene, Effect of UF3 in Alkali-Meta! Base Fuels It has been found that more UF, can be dissolved in the alkali-metal-base fluoride mixtures than in the ZrF ,-base mixtures, and therefore additional tests were made with NaF-KF-LiF (11.5.42.0.44.5 mole %) containing UF,, UF,, and mixtures of the two, r . . ¢ MILS Fig. 6.15, After Circulating UF3 in MuF-ZrF4 for 20060 hr at 1500°F. Unetched. 250X. Reduced 36%. Hot-l.eg Surface of Inconel Loop PERIOD ENDING SEPTEMBER 10, 1954 TABLE 6.7. EFFECT OF MIXTURES OF UF; AND UF, IN NoF-ZrF, ON CORROSION OF INCONEL THERMAL-CONVECTION LOOPS OPERATED AT A HOT-LEG TEMPERATURE OF 1500°F Operating Loop Total U Uas UF, T Metcllographic Notes ime MNo. (wt %) (wt %) (hr) Hot-Leg Appearance Cold-l.eg Appearance 473 8.7 1.2 500 Moderate intergranular subsurface Occasional metal crystols voids 1o a depth of 4 mils 491* 12.8 2.5 500 Moderate to heavy intergranular Deposit to 0.3 mil thick subsurface voids to a depth of _ 7 mils 492* 14.2 1.8 2000 Moderate to heavy intergranular Deposit to 1 mil thick subsurface voids to a depth of 15 mils 469 B.S 0 . 500 Heavy general attack and inter- No deposit granular subsurface voids to a depth of 8 mils 462 13.9 0 500 Moderate to heavy intergranular Metallic deposit to 0.3 mil thick subsurface voids to a depth of 10 mils *These loops were filled from the same boteh of fluoride mixture, and the differences in uranium analysis cannot yet be explained, Two Inconel loops were operated for 500 hr with NoF-KF-UF to which UF, ond UF, had been added. Because of sampling and analytical dif- ficulties, the relative amounts of UF, and UF, are not known, but it appears that about one-third of the uranium was present as U’F3. One loop con- tained a total uranium content of 10.4 wt %, and after the test the hot leg showed light, widely scattered attack to a depth of 1 mil (Fig. 6.16); the hot-leg surface was quite rough. The other loop contained 7.4 wt % uranium, and the hot leg showed a void type of attack that varied from light to heavy, with a maximum penetration of 2 mils; again, the surface was rough, In both loops, a layer was found in the coid leg but not in the hot leg. These mixtures appear to be superior to the ZrF ;-base mixtures with UF, with respect to depth of the subsurface void type of attack, and the hot- leg deposit found with the UF -bearing ZrF ,-base mixtures did not form. More work is needed to de- termine whether the rough surfaces represent a different type of attack and whether the cold-leg deposits formed are NaK,CrF,, which would interfere with heat transfer. Fig. 6.16. After Circulating o Mixture of UF3 and UF, (4.3 wt % UF, and 6.1 wt % UF ) in NaF-KF.LiF (11.5- 42.0-46.5 mole %) for 500 hr ot 1500°F, Etched with modified aqua regia. 250X. Reduced 36%. Hot-Leg Surface of Inconel Loop 97 ANP QUARTERLY PROGRESS REPORT Effect of Zirconium Hydride Additions to Fuel Various amounts of zirconium hydride were added to NaF-ZrF ,-UF, (50-46-4 mole %) as a means of reducing the UF, to UF;. The hydride was added to small portions of fluorides taken from the same original batch, and filters were used when the small batches were transferred to Inconel thermal- convection loops. The data from loops operated with these batches of fluoride mixture are given in Table 6.8. lLayers were found in the cold legs of all loops to which the ZrH, additions had been made. The data show that to obtain sufficient reducing power by the addition of ZrH, to eliminate corrosion it may be impossible to prevent the loss of some uranium both in the treatment pot and in the loop. Effect of Uranium Concentration Two loops were operated with a high-purity NaF- ZrF ,-UF, (53.5-40-6.5 mole %) mixture. This mix- ture is comparable to the one to be used in the ARE and has a higher uranium content than the mixture normally used in thermal-convection loop tests. The heavy hot-leg attack in both loops was of the usual subsurface-void type with @ maximum penetration of 10 mils. This is slightly deeper than the 6 to 8 mils found with the lower uranium content mixtures. Thin metallic-appearing layers were found in the cold legs of both loops. The results obtained with these loops confirm those found previously with similar, but impure, mixtures. Effect of Inconel Groin Size Incone! pipe was annealed at two temperatures to provide specimens with different grain sizes. A series of loops fabricoted from the annealed pipe was filled from the same batch of NaF-ZrF -UF, (50-46-4 mole %) and operated for 500 hr ot 1500°F. Two loops were made from pipe annealed at 2100°F that had a grain size of 1 to ]]/2 gr/in.? at 100 X, while the loop fabricated from as-received lnconel pipe and the one fabricated from pipe annealed at 1600°F contained about 6 gr/in.2. Very little dif- ference in hot-leg attack was found in these loops. Those with the larger grains may have had slightly deeper attack, but the attack was heavier and more general and the deep penetrations were concentrated into fewer boundaries. FLUORIDE CORROSION OF HASTELLOY B IN THERMAL.CONVECTION LOOPS G. M. Adamson Metallurgy Division Loops fabricated from both as-received and over- aged Hastelloy B were operated satisfactorily. The operating mortality rate has been reduced from 90% to 0% in the last group of four loops. The increase in hardness during operation is not so great in the loops constructed with over-oged material as in the loops constructed with as- received material, but, with proper care, the loops of as-received material can be operated. Very little attack was found in a loop which circulated NaF-Zrf ,-UF, (50-46-4 mole %) for 1000 hr at 1500°F. The attack appeared as a few voids to a maximum depth of 1 mil, with possibly some increase in surface roughness. Most of the surface roughness was present in the as-received tubing, as shown in Fig. 6.17. TABLE 6.8. EFFECT OF ZrH, ADDITIONS TO NaF.ZrF o UF, (50-46.4 mole %) CIRCULATED IN INCONEL THERMAL-CONVECTION LOOPS AT 1500°F FOR 500 hr Loop ZrH, Uranium Content (%) Added Hot-Leg Attack Mo. (%) Before Test After Test 469 Heavy general atiack and intergranular voids to a depth of 8 mils 8.5 8.8 459 0.2 Moderate to heavy attack to o depth of 6 mils 8.6 8.5 470 0.5 Light to moderate attack to o depth of 3 mils 7.5 7.1 460 0.9 Thin hot-leg deposit; no attack 5.4 5.1 471 2.0 Hot-leg layer to 1 mil thick; no attack 4.0 4.0 98 MiLS MILS Fig. 6.17. As-Received Hastelloy B (@) and Hot- Leg Surface of Hastelloy B Loop (b) After Circu- lating NciF-Zrl""‘-l.lF4 (50-46-4 mole %) for 1000 hr at 1500°F. Etched with H,GrO, + HCL. 250X. Reduced 36%. Chemical analysis results now available for the Hastelloy B loop previously operated? for 500 hr confirm the low attack rate found metallographically, since neither the nickel nor the molybdenum content in the fluoride mixture increased. LITHIUM IN TYPE 316 STAINLESS STEEL E. E. Hoffman W, H. Cook C. R. Brooks C. F. Leitten Metallurgy Division Tests have recently been compieted on three type 316 stainless steel thermal-convection loops “AG. M. Adomson, ANP Quar. Prog. Rep. june 10, 1954, ORNL-1729, p 77. | PERIOD ENDING SEPTEMBER 10, 1954 in which lithium was circulated. These loops were constructed of 0.840-in.-0D, 0.147-in.-wall pipe. The hot and cold legs were 15 in. in length, and the 15-in. connecting legs were inclined af an angle of 20 deg. The welding and feading opera- tions on these loops were performed in a dry box in a purified helium atmosphere.? & these tests was there any indication of plug forma- tion. The operating conditions are given in Table 6.9, Macroscopic examination revealed no dif- ferences between hot- and cold-leg surfaces in loops 1 and 2. Only loop 1 has been examined completely; loops 2 and 3 have been sectioned and have been examined macroscopically. Loop 2 was very similar in appearance to loop 1, with no At no time dufing crystal deposition. Loop 3, however, revealed mass-transfer crystals attached to the cold-zone walls. These crystals did not plug the loop or noticeably affect the circulation. The crystal deposition was heaviest on the maojor radius of the exposed loop-bend wall in the cold zone. This loop has not yet been examined metallographically. TABLE 6.9. OPERATING CONDITIONS FOR TYPE 314 STAINLESS STEEL THERMAL.CONVECTION LOOPS WHICH CIRCULATED LITHIUM FOR 1000 hr Hot-Zone Cold-Zone Temperawre L.oop N Temperature Temperature Ditferential o. (°F) (°F) (°F) 1 1490 1220 270 2 1472 1355 117 3 1301 1094 207 Loop 1, which was operated at the highest temper- ature and with the highest temperature differential, had no mass-transfer crystals in the cold zone, and the moaximum attack in the hot zone was 1 to 2 mils (Fig. 6.18). Chemical analyses of the lithium and the amounts of crystals recovered are presented in Tcble 6.10. At present it is not understood why so little moss transfer occurred in loops 1 and 2, since in all three loops the same procedures and testing techniques were used and all were filled from the same batch of lithium. Some as yet un- discovered factor seems to have an effect on the rate of mass transfer. 5E. E. Hoffman et al., Met. Div. Semiann. Prog. Rep. Apr. 10, 1954, ORNL-1727, p 37. ?9 ANP QUARTERLY PROGRESS REPORT PLATE 0.005 200X 0.01i0 0.015 INCH Fig. 6.18. Hot-L2g Surface of Type 316 Thermal Convection Loop After Circulating Lithium for 1000 he at 1490°F. Specimen nicke! plated after test to protect edge. Etched with glyceria regia. FUNDAMENTAL CORROGSION RESEARCH G. P. Smith Metallurgy Division Mass Transfer in Liquid Lead J. V. Cathcart Metallurgy Division As previously reported,® the investigation of corrosion and mass transfer in liquid lead has indi- cated that certain alloys possess much greater resistance to mass transfer than their pure com- porients. For example, the time required for small thermal-convection loops containing types 410 and 446 stoinless steel to plug was from two to five times longer than that required for comparable loops confaining pure iron or pure chromium. It has been suggested® that the increased re- sistance to mass fransfer of materials such as the 400 series stainless steels might be related to a tendency toward the formation of intermetallic compounds in these alloys. To test this hypothesis 3. V. Cothcart, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 79. 100 TABLE 6,70. ANALYSES OF LITHIUM CIRCULATED IN TYPE 316 STAINLESS STEEL THERMAL- CONVECTION L.OGCPS Components Other Than Weight of Crystals Material Lithium Removed from Analyzed (ppm) Lithium (g) Fe Ni Cr Lithium, as-cast, 40 10 <10 before test Lithium fromloop 160 150 <10 0.0190 1, after test Lithium fromlcop 120 40 <10 0.001 2, after test further, a loop was operated which contained speci- mens of a specially prepared 45% Cr—55% Co alloy. This alloy composition corresponded to a two-phase region in the cobalt-chromium phase diagram, one phase being the intermetallic compound CoCr, The loop plugged after 768 hr of operation with hot- and cold-leg temperotures of 820 and 510°C, re- spectively. A transverse section of the hot-leg specimen is shown in Fig. 6.19. ' A survey of the alloys which have been tested revealed that the plugging time for all those in which intermetallic compound formation is a possi- bility [types 410 and 446 stainless steel, 2% Si~14% Cr~84% Fe, Hastelloy B (5% Fe-28% Mo~ 67% Ni), 25% Mo—75% Ni, 45% Cr—55% Co, and 16% Ni-37% Cr—47% Fe (austenite plus sigma}] was in excess of 500 hr, the only exception being the loop containing the 16% Ni-37% Cr~47% Fe specimens which plugged in 456 hr. The plugging times for loops containing the pure coristituents of these alloys were: Ni, 2 hr; Co, 80 hr; Cr, 100 hr; and Fe, 250 hr. No mass transfer was observed in a loop containing molybdenum specimens after 500 PERIOD ENDING SEPTEMBER 10, 1954 hr of operation. On the other hand, loops with specimens of alloys for which there is virtually no tendency toward compound formation (types 304 and 347 stainless steel, Inconel, and nichrome) all plugged in less than 150 hr. On the basis of these data, it was concluded that resistance to mass transfer in liquid lead is considerably greater in alloys in which intermetallic compound formation is possible. Future work will include tests of two alloys: one with the composition of 50% Cr—50% Fe and the other with approximately 50% Fe~50% Mo. These compositions correspond closely to the composi- tions of intermetallic compounds in types 410 and 446 stainless steel, and if the ideas presented above are correct, both should show relatively high resistance to mass transfer in liquid lead. e UNCLASSIFIED e e T Y gy Fig. 6.19. Tronsverse Sect‘ion of 45% Cr~35% Co Specimen Exposed to Liquid Lead at 820°C in the Hot-Leg Section of a Quartz Thermal Convection Loop. ' 101 ANP QUARTERLY PROGRESS REPORT Flammability of Sedium Alleys M. E. Steidlitz L. L. Hall G. P. Smith Metallurgy Division The studies of the flammability of liquid sodium alloys, which were reported previously,” were ex- tended. action of jets of sodium-bismuth and sodium-mercury alloys is not significantly changed when the pres- sure of air is varied from 0.25 to 1.0 atm, Additional data on the reactivity of sodium- bismuth solutions in dry air were obtained, and the results are summarized in Figs. 6.20 and 6.21. Figure 6.20 shows the limits of the region of no reaction on o temperature-composition diagram. A tt has been found that the degree of re- careful calibration of the flammability apparatus has shown that there wos a substantial error in the temperatures previously reported in these tests. [n this respect the earlier data have been corrected. The circles represent the temperature-composition values at which tests were conducted. The numeral by each circle gives the number of tests. The un- shaded circles represent tests which showed re- 7G. P.Smith and M. E. Steidlitz, ANP Quar. Prog. Rep. Dec, 10, 1953, ORNL-164%9, p 26, UNCLASSIFIED ORNI~LR-DWG 3030 Co 3 800 | -~ O OO SE— - | | [REACTION ‘ \ ! 2 2 | | | 5 o 0 5 a . i e ////4 _ MR R = 1 , ! v " ] % NO REACTION | / | . / | | ‘ ‘ 1‘ 2 2}/// 2 2 ‘ | ’ ! c o 0z 04 T e MOLE FRACTION OF SCOIUM Fig. 6.20. Temperature-Composition Diagram Showing Regions Within Which lets of Sodium- Bismuth Seolutions Did end Did Net React with Dry Ajr. 102 action, while the rest showed slight reaction. The cross-hatched band demarks the approximate limits within which the line separating the region of reaction from the region of no reaction lies. Figure 6.21 shows how the rate of reaction was found to vary with composition at constant temperature. The rate scale is arbitrary, with no reaction given the value zero and that of pure sodium given the value 4. Further work is being done to determine the flammability temperature for pure bismuth, UNCLASSIFIED ORNL-LR-DWG 3031 4 ] o —*“'7’—@{7*0 a 3 : o i = ; | o ; E o2 <7 Ll [0l i > !,_,__, K 7y M0 + Ko, =7 Ao , , = M0, AF3 Furthermore, the concentrations of oxygen in this equilibrium ond of hydregen in reoction 2 are related through the equilibriom K 1 - 8 o 8 H, + O, =8 H0 AF 105 ANP QUARTERLY PROGRESS REPORT The existence of these equilibria is compatible with Eq. 6a, as may be seen from the fact that AFS 5 = AFg — AF3, Ky 5 = Kg/K;. Hence it would have been possible to derive Eq. 6a by using equilibria 7 and 8. This is because the two sets of equilibria 1 and 2 and equilibria 7 and 8 are both equivalent to the one equilibrium 9) M0, + H, =12 M0 + H,0 AF?S Maba + My = Mgl + Hal 1,2 ° From Eq. 9 it will be noted that the pressure ratio (pHZO/sz)] , is determined by the competition between the oxide and hydrogen for oxygen, with the one trying to form the peroxide and the other trying to form water. In like manner, an expression analogous to Egq. 6a can be derived under the assumption that equi- libria 1 and 3 determine the water and hydrogen pressures, respectively. This expression is P H,0 13 (65) ot Py 1,3 L 1 AF?,:% = eXp { - —- ), s 6RT where AFS o = 3AFS - 2AFS 1/2 K KI 1,3 T T e ' 1/3 Ks 1/2 r 't 1,3 ~ oo ’ N 1/3 I’y By using the free energy data computed above for sodium hydroxide, the representative values shown in Table 6.11 were obtained. The pressure- activity coefficient products in this table give the pressures which would be obtained if the solutions were ideal. Figure (.23 shows the relative im- portance of equilibria 1 and 2 in terms of the logo- rithm of the pressure ratio as a function of the reciprocal of the absolute temperature. Figure 6.24 gives the free energy differences (AF‘I’ , and AF‘]"3) as functions of absolute fempemfure.' Examination of these data shows the following: (1) Apart from unexpectedly large deviations from ideality, equilibrium 1 will predominate at all temperatures. (2) Assuming ideality, hydrogen evolution from fused sodium hydroxide in an inert container will be small at all temperatures, but it should be measurable at 900°C. Likewise, the peroxide concentration at all temperatures may be small, but it should be appreciable at 900°C. (3) Deviations from ideality of a moderate amount could make hydrogen and peroxide formation either insignificant or greatly important at high temper- atures, UNCLASSIFIED ORNL—LR—DWG 3033 30 Lo — L 60 80 100 120 140 160 {1/ TEMPERATURE ] x40* k)™ Fig. 6.23. Decompesition Pressure Functions for NaOH, TABLE 6.11. DATA ON THE DECOMPQOSITION OF SODIUM HYDROXIDE .o -0 - T MMk R 25 35.3 76.0 . 10726 10736 s 600 21.4 64.4 4.7 x 1078 10718 10727 1.7 7.4 x 1078 900 13.5 59.9 3 x1073 8 x 10712 16718 42.0 2.1x107° 106 UNGLASSIFIED ORNL-LR-DWG 3034 Fe (keal/mote) I | 100 10 1 ‘ 200 HOO 1300 TEMPERATURE (°K) Fig. 6.24, Free Energy Differences as Functions of Temperature. There are no satisfactory heat of formation dota for any of the alkali-metal hydroxides other than those for sodium. Therefore, only the rela- tive importance of the decomposition equilibria can be studied. Some of the entropy values for the higher oxides have been estimated and are in the literature. However, the heats rubidium and cesium oxides reported of formation for are quite uncertain, and it is. presently believed that the accuracy of the AF‘{ 5 and [\I*1 5 values given in Table 6.12 would not be improved by use of these entropy estimates. Therefore, only enthalpy values will be used, but it will be shown that this restriction is not so serious as it might at first seem to be. Comparing equilibria 1 and 2 shows AF° 1,2 = AHA‘;20 + AF;fizo - AHM 0, T (Z,\ij All the data required in the above equation are available, except those for the entropy term. Since this term represents o difference in entropy for two similar substances, it may be supposed that its omission will not be serious. A similar situation obtains for AFI 3» Hence, the following approxi- mations will be used. PERIOD ENDING SEPTEMBER 10, 1954 AFS , X AHS . + AF? ~ — AHS , —o ~ A » . The degree of approximation can be checked by using the dato for sodium compounds. In this case AFS g s -35.5 kcal, when using the approximation, as compcred with fhe correct value of -40.7 keal. Likewise, AFS ~94.2 kcal compared with the correct value of -94 9 keal. The approximate values of f\[‘{ and AFS 5 at 25°C are tabulated in Table 6.12 for all the ‘alkali-metal hydroxides. Table 6.12 shows that hydrogen and the peroxide should be produced most readily in potassium’ hy- droxide and least readily in lithium hydroxide. Nevertheless, equilibrium 1 is predominant, and equilibrium 3 and hence equilibrium 4 are insigni- ficant. TABLE 46.12. APPROXIMATE /_\Ft]’ 2 AND AF‘; 5 AT 25°C O 20 {kcal) {kcal} Li ~47 MNa -36 —94 K =25 ~76 Rb —-34 ~73 Cs ~35 =70 Various investigations of corrosion and mass transfer have shown the fundomental role played by hydrogen. The above computations show that hydrogen is important in stabilizing sodium hy- droxide at temperatures above 600°C, as shown in - ASMzoz) Table 6.11, This is the temperature range at which mass transfer becomes significant. These compu- tations also link the formation of hydrogen with the formation of unsaturated oxygen ions. As will be shown in a subsequent report, the presence of these unsoturated ions may be necessary for mass 107 ANP QUARTERLY PROGRESS REPORT transfer to occur. As may be seen in Table 6.12, hydrogen and the unsaturated oxygen ions are most important for the hydroxides of potassium, rubidium, and cesium. Furthermore, these computations provide the first estimate of the thermal stability of the alkali-metal hydroxides (other than LiOH). These estimates confirm the frequently made assumption that the alkali-metal hydroxides are by far the most thermally stable hydrogen-containing liquids known. It must be emphasized that these computations are given as o progress report in a continuing in- vestigation. Thus, many of the valuesobtained, such as the standard free energies of formation, do not in themselves provide direct information about hydroxide corrosion, but they are essential values for any theoretical investigations of corrosion, Experimental Studies (with M. E. Steidlitz, Metal- lurgy Division). An investigation is under way to determine experimentally whether a significant amount of hydrogen is liberated when NaOH is heated to 900°C in an inert environment. The pre- liminary method being tested is tc use a pair of automatic Toepler pumps to concentrate any gaseous products liberated from the hydroxide. The col- lected gas is analyzed with a mass spectrometer, The most difficult aspect of this research has been the endeavor to provide an inert container for the NaOH, which reacts with all metals studied at 900°C (including the most noble metals) to pro- duce hydrogen. It reacts strongly with nearly all ceramics except magnesium oxide. Unfortunately, nonporous crucibles of pure MgO are not available because of the difficulty in sintering this material. Ordinary MgQ crucibles, which usually contain a small amount of binder such as silica, breck down due to attack on the binder. The problem was salved by machining out large single crystals of pure magnesium oxide prepared by the Ceramics Group. In the experiments which have been performed to date, appreciable quontities of hydrogen have been found in addition to much larger quantities of water. However, there was evidence of contamina- tion by organic material and the hydrogen may have come from this source. These studiesare continuing. 134. J. Butiram et al., ANP Quar. Prog. Rep. june 10, 1954, ORNL-1729, o 63, 108 CHEMICAL STUDIES OF CORROSION F. Kertesz Materials Chemistry Division Effect of Temperature on Corrosion of lncone!l and Type 316 Stainless Steel H. J. Buttram R. E. Meadows N. V. Smith Materials Chemistry Division The effect of temperature on the corrosion of Inconel by molten fluorides in static tests, as indicated by the extent of void formation observed and by chromium concentration of the melt, was discussed in a previous repart.'? In those studies there was soime evidence that a maximum in the corrosion could be observed at 800 to 900°C and that both void formation and chromium concentration in the melt were less at 1000°C than at 800°C. While it is possible to rationalize the decreased void formation on the basis that a high rate of dif- fusion for chromium in the metal may minimize the formation of voids, the decreased chromium con- centration in the melt is, however, more difficult to explain. During the past quarter NaZrF ; and NaF-ZrF ,-UF (53.5-40-6.5 mole %) were tested in type 316 stain- less steel under static and nearly isothermal con- ditions for 100 hr at 100°C intervals over the range 600 to 1000°C, as a check of the findings with Inconel. Metallographic examination of the cap- sules exposed at 600°C revealed light intergranular penetration up to 1 mil in depth. At 1000°C the attack was still light, but one isolated case of intergranular penetration to a depth of 18 mils was noted. When the chromium concentration of the melt was plotted against test temperature for the NaZrF ., a slight maximum in the curve appeared between 800 and 900°C; a similar plot for the UF ,- bearing mixture shows a plateau between 700 to 1000°C. However, more chromium appeared in solution in the NeZrF ¢ samples than in the UF,- bearing mixtures. The tendency toward less cor- rosion at 1000°C than ot 800°C never appears to be pronounced, but it has been sufficiently persistent in these tests to require explanation. A plausible explanation can be evolved in the following manner. The NaZrF, melt which has been used in these experiments gove rise to ab- normally high chromium concentrotions (1000 ppm) compared with equilibrium values (200 ppm) ob- tained in measurements carried out with carefully handled melts and pure chromium metal. It appears, therefore, that impurities were predominantly re- sponsible for the amounts of corrosion measured. The most likely impurity is HF, which con result from hydrolysis of adsorbed water or hydrated water in salts containing Na Zr,F,, or from inadequate purification. If HF were responsible, the ascending portion of the curve could be due to an increasing rate of reaction with increasing temperature. (There is little reason to believe that equilibrium conditions are reached in static capsules in 100 hr.) The descending portion of the curve may be ascribed to a marked decrease in corrosiveness of HF at increasing temperatures. Standard free energy estimates'* show that AF® for the reaction of HF with Ni to form NiF, becomes positive near 300°C, with Fe to form FeF, ot 800°C, and with Cr to form Crf, at 1400°C. In the experiments under discussion here, the melt at the end of a 100-hr test contains mostly Cr*¥, with very little Fe*" and Ni*™, There is some evidence that at the lower temper- ature ali three elements are attacked rather slowly and indiscriminately by HF and that the resulting Fe'' and Ni*' are replaced by Cr*™¥ in o fast secondary reaction. The effect of increasing temperature is to increase the degree of approach to equilibrium in 100 hr in a static capsule; how- ever, at higher temperatures the effect of a more unfavorable equilibrium constant becomes manifest. Iron and nickel are relatively unreactive toward HF at 1000°C, and the amount of Cr™" pickup in 100 hr could well be less than that noted at 800°C for three reasons: the free energy change for the reaction of chromium with HF is smaller, fewer Fe** and Nit" ions are present for reaction with Cr**, and there is mechanical interference by the unreacted nickel and iron. In other words, chromium is most readily oxidized from an alloy by HF if the accompanying alloy constituents are alsc attocked. Corrosion by Fission Products H. J. Buttram R, E. Meadows Materials Chemistry Division in a previously reported experiment,'® a fuel mixture containing simulated fission products ot 1000 times the concentration expected in the ARE My | Brewer et al., The Thermodynamic Propierties and Equilibria at High Temperatures of Uranium Halides, Oxides, Nitrides, and Carbides, MDDC-1543 {Sept, 20, 1945, rev. Apr. 1, 1947), PERIOD ENDING SEPTEMBER 10, 1954 was corrosion tested. Very heavy attack was observed; this was expected, not becouse normal amounts of fission products are particularly cor- rosive but because oxidizing agents such as RuF , MoBr3, and elemental tellurium were used in suffi- cient concentration fo cause excessive corrosion in any case, Additional experiments were performed with the use of the same additives, individually, in amounts as small as possible in an attempt to measure the relative activities as corrosive agents. Preliminary results, such as those previously reported which showed YF, to be extremely corrosive, and oddi- tional trials during the past quarter, which showed that such compounds as CsF were also very cor- rosive, made it obvious that the techniques em- ployed were unsuitable and that very misleading indications were being obtained. Free energy considerations, as well as general experience with fluoride systems in closed capsules, make it ap- parent that impurities such as HF and water were responsible for the effects noted. Hence it must be concluded that none of the experiments per- formed to date on corrosion by simulated fission products have heen sotisfactory for the intended purpose. Controlled-Yelocity Corrosion Testing Apparatys N. V. Smith F. A, Knox Materials Chemistry Division In order to more nearly simulate corrosion con- ditions of molten fluorides circulating through reactor components, a controlled-velocity corrosion testing apparatus has been constructed. This ap- paratus, similar to one previously described, '® al- lows the rapid transfer of molten fluorides through both heated and cooled test sections. The apparatus consists of two 4-in.~1D, 24-in.-high Inconel cylin- drical pots connected by means of three sections of 1/‘(in. inconel tubing with ¢ wall thickness of 0.035 in. The two outside sections are 5 t long and the center section, which was initially 1 ft long, was later increased to 3]/2 ft. The center section is cooled by either oir or water. Suitable furnaces and heaters permit holding the pots and transfer lines at desired temperatures. FEach Inconel pot has a gos inlet which allows opplication of helium Y5, 3. Buttram et al., ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, ¢ 65, 1614, W, Hoffman ond J. Lones, ANP (Quar. Prog, ‘Reb. Mar. 10, 1954, QRNL-1692, p 98. 109 ANP QUARTERLY PROGRESS REPORT at predetermined pressures to transfer the melt, A probe indicates when the desired liquid level is reached in the second pot and activates a relay which closes the helium inlet to the first pot and opens a corresponding one in the second pot to reverse the cycle. Velocities obtained with the apparatus have reached 5 fps, which corresponds to a Reynolds number of only about 1500. It is hoped that an improved design now being assembled will make possible turbulent flow. A 200°C gradient across the cooling section was desired; however, it has not yet been possible to achieve a differential of more than &0°C. The transfer cycle consists in having the melt contained in either pot at 700°C, being heated to 800°C in the first section of the transfer line, cooled in the middle section, brought back to 700°C in the third section, momentarily stored in the opposite pot at 700°C, and then the cycle is reversed, A complete cycle, that is, from one pot to the other and back, required approxi- mately 5 min. It is believed that this apparatus will yield corrosion data under conditions not easily obtainable by present means. Reaction Between Grophite cad Flueride Melt F. A. Kinox Materials Chemisiry Division Graphite has been used extensively as a con- tainer material during preparations of fuel and 110 coolant materials; therefore, an investigation was made of possible reaction between it and a fluoride melt. In the procedure used an lnconel reaction tube was loaded with either NaF-ZrF -UF, (53.5- 40-6.5 mole %)} alone or with the fluoride mixture plus graphite which had been treated under vacuum for degassing. The reaction tube was connected to @ manometer and then held at 1000°C until the gas pressure became constant. It was found that the rather large gas pressures obtained were due chiefly to SiF,. Analysis of the graphite indicated the presence of nearly 1% Si. The use of spectrographically pure graphite in this apparatus has yielded pressures very nearly the same as the pressure of the pure fluoride. No HF or CF, was detected during these tests. Lithium Fivoride Castings N. V. Smith F. Kertesz Materials Chemistry Division At the request of the Instrumentation and Controls Division, high-density lithium fluoride cylinders were cast from powdered material for use in an x-ray spectrometer. These cylinders were pre- pared by charging the powder to grophite crucibles and heaiing in an inert atmosphere to 900°C. The graphite liner containing the melt was then re- moved from the furnace and a strong air jet was applied to its periphery. The resulting cylinders were uniform and acceptable. PERIOD ENDING SEPTEMBER 10, 1954 7. METALLURGY W. D. Maniy Metallurgy Division Creep and stress-rupture testing have, in the past, been performed primarily on uniaxially stressed specimens. lests of this type are continuing, but the interest is now centfered on tube-burst tests in which a tube that is closed at one end is stressed with on internal gas pressure. The stress pattern introduced into the specimen in this test more nearly approaches the stress pattern that will be found in ANP-type reactors. The ratio of the tan- gential stress to the longitudinal stress is quite important in the ductility of the metal being tested, and therefore equipment in which the lengitudinal and the tangential stresses can be varied has been constructed. A theoretical analysis of a stressed cylindrical pressure vessel has been made with which a check on the experimental results con be obtained. Also, to determine the effect of tan- gential and compressive stresses on the corrosion rate of metal in contact with fused salts, an ap- paratus for loading a sheet specimen in bending was designed and constructed; tests are now being made. In the investigation of high-thermal-conductivity materials for fins for sodium-to-air radiators, stresse rupture and creep tests were made on copper fins with various types of cladding at stress levels be- tween 500 and 2000 psi at 1500°F. The tests have shown that for a 1000-hr exposure in air, stresses greater than 500 psi and less than 1000 psi are tolerable; that is, in this stress range there is no indication of brittleness in the core or oxidation of the core due to cladding failure, Thermal conduc- tivity measurements of a 6% Al-94% Cu aluminum bronze were made in the temperature range 212 to 1562°F, ' Tests of brazing alloys have shown that in braz- ing high-conductivity fin materials to Inconel tubing there are four alloys that can be used: low-melting- point Nicrobraz (LMNB)}, Coast Metals alloy 52, electroless nickel, and an Ni-P-Cr alloy. From the over-all considerations of melting point, oxidation resistance, dilution of fin and tube wall, formation of low-melting eutectics, and flowability, it was found that Coast Metals alloy 52 was the best alloy for the construction of radiators with high- conductivity fins. A sodium-to-air radiator with 6 in. of type-430 stainless-steel-clad copper high-conductivity fins was fabricated by using a combination heliarc welding and brazing procedure. The tube-to-fin sections were assembled and brozed with Coast Metals alloy 52. The tube-to-header joints were made by using the semiautomatic welding equipment and then back-brazing. The header sections were closed by manual heliarc welding. In the construction of a 100-kw gas-fired liquid- metal-heater system, packed-rod nozzle assemblies were needed as the inlets for gir and gas. These devices were made by brazing ]/B-in.-dicu stainless steel rods in o tight ossembly with a ductile, oxidation-resistant alloy of 82% Au-18% Ni. Work has started on the formation of duplex tubing in an attempt to prepare alloy composites that have good corrosijon resistance on the inner surface and oxidation resistance on the outer surface. Com- posites of copper and type 310 stainiess steel, Inconel and type 310 stainless steel, Inconel and Hastelloy B, and Hastelloy B and type 310 stainless steel have been prepared. Tubing produced from stainless-steel-clad molybdenum and columbium and from several special Inconel-type alloys was fabri- cated into thermal convection loops for fluoride corrosion testing. Several new alloys were produced for the corrosion tests with liquid lead. Attempts are being made to find new olloys in the nickel-molybdenum system that will have better high-temperature strength and fluoride corrosion resistance than Inconel has. Hastelloy B satisfies these requirements, but it has poor fabrication properties and oxidation resistance, and it loses its ductility in the temperature range of interest for application in high-temperature circulating-fuel re- actors. Investigations are under way in an effort to find a suitable melting and heating treatment that will increase the ductility of Hastelloy B in the temperature range of interest, Investigations of methods for producing boron carbide shield pieces of the required density and shape indicate that the pieces should be molded with nonmetallic bonding material by cold pressing followed by sintering. The bonding materials being m ANP QUARTERLY PROGRESS REPORT studied include sodium silicate, silica, silicon nitride, and boric oxide. The fluoride-to-sodium intermediate heat ex- changer, which failed in a life test after 1680 hr of cyclic service in the temperature range 1080 to 1500°F, was exomined, |t is probable that the failures in the tube-to-header welds were caused by unequal thermal expansion, which caused stress concentration at the roots of the tube-to-header welds. These stresses, combined with expansion and contraction of the tubes, would tend to propa- gate cracks through the walls. These failures emphasize the extreme desirability of using back- brazing as o means of minimizing the notch effect in tube-to-header welds. STRESS-RUPTURE TESTS OF INCONEL R. B. Oliver D. A. Douglas J. H. DeVan J. W. Woods Metallurgy Division The tube-burst test for obtaining information on the stress-rupture properties of Inconel has been studied intensively. The siress pattern introduced intfo the specimen in this test simulates to some extent the stress pattern that will be present in circulating-fuel reactors. The test consists of stressing a closed-end tube with internal gos pres- sure. In tests of this type reported previously, | it was observed that Inconel specimens stressed in this manner showed less ductility and much shorter rupture life than the uniaxially stressed specimens, and therefore an intensive study of the multiaxial stress system was initiated. W. Jordan of the Mechanics Deportment of the University of Alobama is investigating this problem. Part of the investigation consists of a study of the theory of stresses in cylindrical pressure ves- sels, with particular attention to the variations in siresses calculated by the thin-wal! formula vs the stresses determined by the more exact Lamé theory. The stresses under discussion are those in the walls only, with no consideration given to the end closure shape, except that the ends are assumed to be completely closed. The three principal stresses at a given point are the radial stress (o), the tan- gential (hoop) stress (o), and the longitudinal (axial) stress (o ). The theory of elasticity yields 2. B. Oliver, D. A. Douglas, and J. W. Woods, ANP Quar, Prog. Rep. June 10, 1954, ORNL-1729, p 89. 112 the following equations: o pir? - por?J ( 1 ) r?ri (PO Pz-) o = e L | ri — r? T2, Tg - T Piff"‘POTi | rro(po~{71) o, MHE (P rz -7 \f2 r — T pir? - porg (3) O‘a = _‘2’—;“ ’ reo— T where r is the radial distance to the point at which the stress is desired, 7 and r are the internal and external radii, and P and p, ore the infernal and external pressures, For the special case of the cylinder subjected to internal pressure only (po =0}, the equations reduce to: (4) Opi = Pi s Yo T 0, r2 + r2 2r2 5) _ Q 2 _ ( oti - pzl 2 2 ! g[o o pz —-2— ----------- ' r — 7. r r. o i o i 2 7i (6) Oaz’ = Gao = pl 2 ! r°- - r where the additional subscripts 7 and o on the stress terms indicate stresses ot the inner and outer surfaces. To simplify these equations for application to cylinders with thin walls, it is fre- quently assumed that the tangential stress does not vary across the wall of the vessel. This simpli- fication results in the following equations for in- ternal pressures only; (7) S,. =5 (8) s . =395 = 1"5; ’ where the letter S is used to denote stress and ¢ is the thickness of the wall (¢ = r- ri)' A simpli- fied expression for radial stress in thin-walled cylinders is not commonly used. In order to determine the error involved in using the formulas for thin-walled vessels, ratios of the stresses computed by the exact and the approximate theories are given below. ¢ U,. ' r. (9 e S, r t te z 2 4 — Ti 1o 2 (10) 2 , Sto 2 +_i_ 7 o (11) a2 Sa 2 +L Equations 9 through 11 are plotted in Fig. 7.1. The so-called ‘“‘exact’’ formulas (often called the Lamé formulas) are valid only while the stresses are elastic. The approximate formulas are based on equilibrium considerations and the assumption that the stresses are distributed uniformly across the wall. Under inelastic (including creep) con- ditions, it is believed that the tangential (hoop) stress distribution lies somewhere between those assumed in the Lamé and in the approximate elastic theories. With increasing stress, increasing temper- ature, and increasing time, the actual stress dis- tribution will depart from the Lamé assumption and approach the thin-wall theory assumption. Thus the two theories give the upper and lower bounds for the actual stress distribution. For the longitudinal (axial) stresses, the Lame and the thin-wall theories both use the assumption that the stresses are uniformly distributed over the cross section, Since the Lame theory uses the correct value for the cross-sectional area, while the thin-wall theory uses an approximate area, the Lamé theory is believed to be more correct in both the elastic and the inelastic ranges of stress. The majority of the tube-burst tests made to date have been of 0.010- and 0.020-in.-wall tubing, and thus the approximate formulao applies. Specimens are now being tested which have 0.060-in. walls, Since it is predicted that under the test conditions the PERIOD ENDING SEPTEMBER 10, 1954 UNCLASSIFIED ORNL~-LR-DWG 2848 1.28 : I ] | | { i /"2 i ’. - " e ! 1,24 — faa"‘“f/'a (\Sf'—PT A = N 1.20 B | 2.2 T T THIN-wALL 118 L. r,e - ,72 THEORY ] LAME THEORY Ve LI 1.08 | a = 108 Lo $.00 096 032 |- 0881 o STRESS COMPUTED 8Y LAME THEGRY - T~ | S STRESS COMPUTED BY THIN —WALL THECRY 084 | | J | o) 005 010 015 0.20 0.25 0.30 . I‘}' Fig. 7.1, Comparison of Elastic Stress in Cylinders as Computed by Lamé Theory and by Thin-Walled Pressure Vessel Theory. stress will be uniform across the wall thickness the approximate formula will be used for calculating the stresses, and the slight error thus introduced will be neglected, It is well known that in the case of a thin-walled closed-cylinder pressure vessel the ratio of the tangential (hoop) stress to the longitudinal (axial) stress is 2:1. Some investigators have shown that this stress distribution results in the minimum ductility for a given material, Therefore, in order to better understond the criteria for predicting failure of tubular specimens and in order to study the effects of anisotropy of the tubular material, it would be desirable to be able to test a series of specimens with varying stress ratios. Jordan has devised the apparatus described below with which it is possible to stress a tubular specimen purely tangentially, purely axially, or in ony desired ratio of these stresses. ' 113 ANP QUARTERLY PROGRESS REPORT The most obvious method of changing the stress ratic is by loading a tubular specimen axially with dead weaights ond producing an axial stress without the use of pressure, so that o /0 =0, By using movable pistons for the ends of the cylinder, as shown in Fig. 7.2, all the axial load caused by the pressure acting on the pistons is borne by the rod connecting the pistons. This will reduce the axial stress in the cylinder wall to zero (assuming no friction between piston and cylinder) without af- fecting the tangential stress. In actual practice, one piston can be reploced by a fixed end. By using @ piston of larger net area than the opposite cylinder end, compressive axial stresses can be introduced into the cylinder walls without altering the tangential stresses. This arrongement is shown in Fig. 7.3. By using thin-walled cylinder approxi- mations, the siress ratio is given by the equation vy 24 {d + 1) Ya a’2 -~ D2 UNCLASSIFIED ORNL-LR-DWG 2849 | ——— ——PISTON SNONCNY / ___—SPECIMEN o= PISTON ROD NN e PISTON — _Z\ Fig. 7.2. Apparatus for Producing Tangentia! Stress in o Tubular Specimen, 114 (Note that the diameter a of the piston rod does not enter into the calculation of the stress ratio.) In order to obtain a positive stress ratio of 0 /o, an arrangement similar to Fig. 7.4a can be used, The stress ratio is given by the equation In the event that it is necessary to extend the piston rod through the top end of the cylinder {to provide o thermocouple well, for example), the arrangement shown in Fig. 7.45 may be considered. The stress ratio is given by the equation Uz 24 (d + t) Ua D2 __.bz UNCLASSIFIED ORNL-LR-DWG 2850 /A - \\B e PISTON / AN = ) ———PISTON ROD A I / ———SPEGIMEN YN\ \\R 8 Fig. 7.3. Apparetus for Producing Compressive Axial Stresses in a Tubular Specimen Without Altering the Tangential Stresses. PERIOD ENDING SEPTEMBER 10, 1954 UNCLASSIFIED ORNL-LR-DWG 285t DIMENSION O MAY BE GREATER OR LESS THAN ¢ s X, 7 7 //. 7 /) i e 7 Y | e f A a b Fig. 7.4. Apparatus for Obtaining a Positive Stress Rafio of a,/oru in o Tubular Specimen. (o) With- out thermocouple well. (&) With thermocouple well. 115 ANP QUARTERLY PROGRESS REPORT As before, the stress ratio is independent of the diameter 2 of the piston rod. Since the piston rod is in compression, however, it should be made large enough to prevent buckling. The bleed lines shown in Figs. 7.4a and & should be kept open during testing to assure that no pressure leaks from the large-diameter chamber to the smaller one. A test rig has been built that imposes the con- dition 7 /o Strain measurements made with a strain-gage bridge fastened to the gage area showed that the friction in the system was negligible and that the only active stress was tangential. A speci- men is now being tested at 1500°F under this stress condition, = oo, There has been considerable speculation as to the possibility of o difference in the rate of cor- rosive attack of fused salts in contact with Inconel under tensile siresses as compared with lnconel in compression, In order to observe any difference, it is desirable that the compressive and tensile stresses be imposed on the same spscimen and be of the same magnitude. One approach to this prob- lem is to test a specimen under pure bending con- ditions. This would ensure that the magnitude of the maximum tensile stress would always be equal to that of the maximum compressive stress during a given test. An apparatus which would produce this effect was also designed by Jordan, and a specimen is now being tested in a fused salt medium. A drawing of the specimen and the loading opparatus is presented in Fig. 7.5, If it is assumed that no friction occurs between pins C and D and the specimen, all the force P will be transmitted across section |-l by the link L and no axial force will exist in the specimen. Likewise, since link L is loaded axially, it cannot transmit any bending moment and therefore the specimen must resist any bending moment existing across section |-l. The force in link L will be equal to P to satisfy the equilibrium of forces. The bending moment in the specimen will then be equal to the product of P and the horizontal distance between the centers of pins A and B. In order to minimize variations inthe bending moment (and therefore stresses) in the specimen during a test, the horizontal distance between A and B should be held as nearly constant as possible. Since the specimen deformsunder load, the ends will rotate and some change in the distance AB must occur, In order to minimize the change, the pins A and B are placed with their centers slightly above and below, respectively, ahorizontal 116 UNCLASSIFIED ORNL—LR—DWG 28582 VTR T SPECIMEN Fig. 7.5. Apparatus for Testing a Sheet Speci- men Under Pure Bending Conditions. diameter, This will allow some rotation to occur with very little change in the horizontal distance, The apparatus shown in Fig. 7.5 has been con- structed. Strain gages were placed on the tension and compression sides of a specimen, and load was applied. In the range of loads contemplated in a creep test, the tensile and compressive strains were essentially of the same magnitude; thus very little friction existed between the pins and the specimen, HIGH-COMDUCTIVITY-.FIN SODIUM-TO-AIR RADPIATOR Developmental work hos continued on o sodium- to-air radiator with fins of a high-thermal-conductivity material, The developmental effort includes in- vestigations of materials with high therma! con- ductivity, the development and testing of brazing alloys for use in fabricating the radiator, and the fabrication of radiators for service testing. Investigations of Fin Materials J. H. Coobs Metallurgy Division H. Inouye Stress-rupture and creep tests of type-310 stain- less-steel-clad copper fins that were 8 mils thick were made at 1500°F at stresses between 500 and 2000 psi. The material was obtained from the General Plate Div. of Metals & Controls Corp. and had an average cladding thickness of 1.87 mils per side. The maximum cladding thickness was 2.3 mils per side. The results of the several tests made show that some alteration of the copper oc- curs even though the specimen does not fracture. This was indicated by the brittleness of the normally ductile composite. For long-time exposures (1000 hours}, stresses greater than 500 psi and less than 1000 psi are tolerable and there is no indication of brittleness in the core or oxidafion due to cladding failure. In all the specimens which ruptured, oxi- dation of the copper at the point of fracture was complete, and the copper which did not appear to be oxidized was brittle. It was also noted that when the strain in 2.5 in. was 10%, the copper re- mained ductile, while strains above 20% (regardless of exposure time} produced brittleness. The data obtained are given in Table 7.1, - The aluminum bronzes are among the materials being considered as cladding for copper fins. Al- though extensive diffusion occurs between these alloys and copper, increases in the thermal con- ductivity of the fin can be realized by cladding the copper with these alloys. The composition gradient existing in a diffused composite results in an in- crease in the conductivity of the cladding material (because of the outward diffusion of aluminum) and a decrease in the conductivity of the copper core. Thermal conductivity measurements of two alumi- num bronzes were made by the Solid State Division and are reported in Table 7.2. The values given for measurements at temperatures up to 392°F com- pare favorably with the data of Smith and Polmer;z but for temperatures above 392°F they exceed the values extrapolated from their data. Also, the 2¢, S, Smith and E. W, Palmer, Am, Inst. Mining Met, fing%rg), Metals Technol. Tech. Pub. No. 648, 1-21 PERIOD ENDING SEPTEMBER 10, 1954 TABLE 7.1. 5TRESS.RUPTURE PROPERTIES OF TYPE-310 STAINLESS-STEEL-CLAD COPPER FINS AT 1500°F 2-mil cladding on each side of 4-mil copper sheet Rupture Elongation Stress (p5|) Tlme in 2.5 in. Remqus (hr) (%) 2000 96 39 Copper embrittled 1800 336 50 1650 528 40 Copper embrittlied Copper embrittled 1500 887 Copper embrittied 1400 1220 In progress 1000 27.5 Test terminated at 1000 hr; specimen oxidized throughout 1000 .20 Test terminated at 500 hr; oxide stringers on surface of specimen 500 10 Test terminated at IBOO hi; no indicatjon of failure 500 5 Test terminated at 500: hr; no indication of failure TABLE 7.2, THERMAL CONDUCTIVITY OF ALUMINUM BRONZES AT VARIOUS TEMPERATURES Therma! Conductivity Temperoture [C"'/SBC‘sz (DC/Cm)] (°F) For For 6.2% Al.-93.8% Cu 8.4% A1-91.6% Cu 212 0.178 0.176 302 0.240 0.210 392 0.280 0.250 1156 0.55 1562 0.77 estimated value of six times the thermal conductivity of stainless steel ot 1500°F was exceaded by the experimental value for the 6.2% Al-93.8% Cu alloy. The values given in Table 7.2 were not corrected for the volume expansion of the alloy with temper- ature and are therefore accurate only to within +5%. 117 ANP QUARTERLY PROGRESS REPORT Development of Brazing Alloys for Use in Fabricating Rediators P. Patriarca K. W. Reber G. M. Slaughter Metallurgy Division J. M, Cisar Aircraft Reactor Engineering Division Several alloys were investigated for brazing copper fins clad with Inconel, type 310 stainless steel, or type 430 stainless steel in order to es- tablish an optimum combination for fabrication of a sodium-to-air radiator, The evaluation of these alloys was bosed on metallographic examination of tube-to-fin specimens before and after exposure to static air for periods up to 600 hr, The alloys tested are listed in Table 7.3. TABLE 7.3, BRAZING ALLOYS FOR . o Brazing Alloy Name ~-omposition Temperature (et %) (°F) Low-melting-point 80 Ni, 6 Fe, 5 Cr, 1920 Nicrobraz (LMNB) 558i,3B,1C Coast Metals alloy 89 Ni, 55i, 4 B, 1840 52 ?2 Fe Electroless nickel 88 Ni, 12 P 1800 Ni-P-Cr 80 Ni, 10 P, 10 Cr 1800 A microsectionis shown in Fig. 7.6 of an Inconel- clad copper fin (2 mils of Incone! on each side of 6-mil-thick copper) joined to gflé-in.-OD, 0.025-in.- wall Inconel! tubing with 88% Ni~12% P alloy and then exposed to staftic air at 1500°F for 600 hr. Solution of the tube wall during the brazing cycle was negligible, but rather severe solution of the fin lip occurred. It is evident, however, that the oxidation resistonce of the joint was not affected. Examinations of specimens brazed with the other alloys listed in Table 7.3 revealed similar results, that is, minimum solution of tube walls and ex- cellent oxidation resistance of the braze joints. The effects of diffusion, however, were in evi- dence after each test, Color changes in the copper core, along with the formation of voids, indicate that the use of Inconel-clad copper in a radiotor would not be advisable, since the void formation 118 ond the change in core chemistry from copper to a copper-nickel alloy would seriously reduce the heat transfer efficiency of the fin. Other claddings, such as types 310 and 430 stainless steel, have been found to be relatively immune to diffusion effects, however, and a number of brazed joints were evalu- ated to determine the optimum alloy for brazing each of these materials to Inconel tubing. A joint of type-310 stainless-steel-clad copper brazed to Inconel tubing with the 88% Ni-12% P alloy is shown in Fig. 7.7. It may be seen that fin dilution and intergranular penetration of the resulting brazing alloy into the tube wall occurred. This penetration, undetected in the case of Inconel-clad copper, appears to be due to the formation of o complex eutectic resulting from the presence of an iron-base cladding material rather than from «a nickel-base material. Similar intergranular pene- tration was found when the 88% Ni-12% P ailoy A , UNCLASSIFIED Y-12604 0.01 |0.04 0.05 0.c8 Fig. 7.6, Incone! Tubing with 88% Ni.12% P Alloy ond Incone!-Clod Copper Fins Brazed to Exposed te Stotic Air at 1500°F for 600 hr. Note dilution of fin, hole formation in copper due to diffusion, and the excellent oxidation resistance of the bronze joint. As polished. 50X. Reduced 15%. was used with type-430 stainless-steel-clad copper, os shown in Fig. 7.8, In this case, the intergranular penetration was completely through the tube wall. The etfect of the intergranular penetration on the physical properties of the tubing has not yet been determined, but it is believed that it would be ad- visable to use a brazing alloy that does not react with the iron-base cladding material. | The tube-to-fin joints shown in Figs. 7.9 and 7.10 are, respectively, type-310 stainless-steel-clad and type-430 stainless-steel-clad copper fins that were brozed with Coast Metals alloy 52 to Inconel and then exposed to static air for 400 hr at 1500°F, A negligible amount of dilution and o relatively minor amount of boron diffusion into the Inconel tubing occurred. " Although low-melting-point Nicrobmz (LMNB) has been used successfully to braze all the proposed cladding materials to Inconel, its relatively high . UNCLASSIFIED ~ : S D.02 0.04 0.05 0.06 0.09 Fig. 7.7. Type-310 Stainless-Steel-Clad Copper Fins Brozed to Inconel Tubing with 88% Ni~12% P Alloy. Note dilution of fin and intergranular pene- tration into tube wall. As polished. 50X. Re- duced 15%. : PERIOD ENDING SEPTEMBER 10, 1954 flow point of 1925°F is so close to the melting point of copper that it cannot be used with ease, Since the oxidation resistance of the Coast Metals alloy 52 is comparable to that of LMNB and the effect of boron diffusion is the same for both alloys, the Coast Metals alloy was selected for the fabri- cation of the first high-conductivity-fin sodium-to-air radiator. ' ‘ The problem of edge protection of the clad copper fins has been resolved. Aluminum is opplied to the edges of fins by dip coating or spraying, or it is painted on as aluminum powder in a Nicrobraz cement slurry. The fin is then heated ot 800°C for approximately 3/2 hr. The diffusion of the aluminum into the copper core results in an aluminum bronze which exhibits excellent resistance to oxidation when exposed to static air for 500 hr at 1500°F, It appears that the depth of bronzing, which con- tinues during oxidation testing, is about 3/32 to 1/8 in. in 500 hr. The feasibility of simultaneously bronzing large numbers of fins is to be determined. NCLASSIFIED & ¥ 12602 T L 0.01 0.04 kL 0.05 f (o g.08 Fig. 7.8. Type-430 Stainless.Steel-Clad Copper Fins Brazed to inconel Tubing with 88% Ni-12% P Alloy. Note dilution of the fin and severe inter- granular penetration into the tube wall. As pol- ished. 50X. Reduced 19%. 119 ANP QUARTERLY PROGRESS REPORT UNCLASSIFIED Y-12601 0.04 0.05 0.06 Fig. 7.9. Type-310 Stainless-Steel-Clod Copper Brazed to Inzone! Tubing with Coast Metals Alloy 52 and Exposed to Static Air at 1500°F for 400 hr. As polished. 50X. Reduced 16%. Radiator Fabrication P. Patriarca K. W. Reber G. M. Slaughter Metallurgy Division J. M. Cisar Aircraft Reactor Engineering Division A sodium-to-air radiator with 6 in. of type-430 stainless-steel-clad copper fins wos fobricated by using a combination heliarc welding ond brazing procedure. The tube-to-fin section was assembled and brazed with Coast Metals alloy 52 at 1020°C, The split headers were then assembled and heliarc welded by wusing the semiautomatic equipment available in the Welding Labeoratory. A photograph of the radiator and welding torch after the welding of the 108 tube-to-header joints is shown in Fig. 7.11. The remaining header sections were then monually heliorc welded by utilizing the complete penetration technique. Although o pressure test indicated that six tube-to-header joints were not leak-tight, a 120 0. .02 )4 0.03 ©.04 £ 10.05 0.08 Fig. 7.10. Type-430 Stainlass-Steel-Clad Copper Brazed with Coast Metals Alloy 52 and Exposed to Static Air ot 1500°F for 400 hr. As polished. o0X. Reduced 17.5%. bock-brazing operation satisfactorily sealed the leaks ondremoved the effects of notches as sources of stress concentrations. The completed unit, as shown in Fig. 7.12, was helium leak-tight ofter the rebrazing operotion. This unit is to be used in the 100-kw gos-fired liquid-metal-heating system. NOZZLES FOR THE GAS-FIRED LIQUID-METAL-HEATER SYSTEM P. Patriarca K. W. Reber G. M. Slaughter Metallurgy Division J. M. Cisar Aircraft Reactor Engineering Division One of the design features of the 100-kw gas-fired, liquid-metal-heater system now heing constructed is the packed-rod nozzle assembly to be used for inlet cantrol of the gir and the gos flow. Test data and theoretical considerations indicated that ]/B-in.-dia stainless stee!l triangulorly packed rods would be satisfactory for use in this nozzle, |t was expected that these rods could be joined in o rigid bundle UNCLABSIFIED Y-12916 Fig. 7.11. Sodium-to-Air Radiator Showing Semi- automatic Heliore Tube-to-Header Welds. with @ minimum of plugged interstices by furnace brazing with a ductile oxidation-resistant high- temperature brazing alloy such as 82% Au-18% Ni. Experiments revealed that proper jigging was extremely important in obtaining the completed air- nozzle blank (Fig. 7.13). Each layer of rods was tack-welded for rigidity, ond the layers were then tack-welded to form the assembly. Since the de- termination of the exact amount of brazing alloy required was difficult, if not impossible, sufficient material was used to ensure good bonding at all capillary joints. The excess material was then removed from the nozzle interstices by placing a flat cross section against a metal sheet on which PERIOD ENDING SEPTEMBER 10, 1954 B UNCLASSIFIED ¥-13019 Fig. 7.12. Completed Sodium-to-Air Radiator Showing Tube-to-Fin Manifold Construction. many fine lines had been scribed. Upon rebrazing, these fine scratches acted as capillaries and the excess material was removed. The design of the natural-gas nozzle specified that a regular pattern of interstices should be selectively closed. The closures were obtained by inserfing wires into the designated holes ond brazing them to the rods. SPECIAL MATERIALS FABRICATION RESEARCH J. H. Coobs H. lnouye Metallurgy Division Stainiess-Steel-Clad Molybdenum ond Columbium Two thermal-convection loops constructed of stainless-steel-clad molybdenum were assembled by 121 ANP QUARTERLY PROGRESS REPORT UNCLASS|FIED Y-12934 o 1 Brazad Air-Mozzle Bilank. bonding at the points of tangency of the stainless steel rods. Fig. 7.13. Note using threadad molybdenum joints, which were then protected from oxidation by the use of welded cover plotes. Both loops failed early in corrosion tests because of fraciure of the cover plates and leakage frem the |oints. It is believed that overheating of the cover plates occurred while the loops were being initially heated by alectrical resistance., Since no metallurgical bond existed between the molybdenum and the cladding, separation became more pronounced with increased temperatyre, until finally all the current was being carried by the rather thin cover plates, Future molybdenum thermal-convection loops will be made from arc-melted molybdenum, which is weldoble, and the heavy nickel electrodes required for resistance heating will be welded directly to the molybdenum tube instead of to the cladding. Columbium clad with type 310 stainless stee! has been fabricated into a thermal-convection loop by heliarc welding in air. This loop will be tested as soon as the accessory parts are available., 122 Inconel-Type Allcys Eight thermal-convection loops of Inconel-type alloys were fabricated. The compositions and room temperature properties of the alloys are given in Table 7.4, The alloys are being studied as a part of the effort to obtain alloys with better high- temperature corrosion resistance than that of Inconel. TABLE 7.4, PROPERTIES OF HIGH-PURITY INCONEL-TYPE ALLOYS”® ¥ Nominal Tensile Yield - _ Neof Composition Strength Point or(u(c;c)mon e (wt %) {psi) (psi) ° 14 18 Cr, 7 Fe, 75 Ni 85,000 41,500 49 15 10 Cr, 15 Fe, 75 Ni 77,700 38,800 44 16 5 Cr, 20 Fe, 75 Ni 72,700 34,200 47 17 10 Cr, 7 Fe, 83 Ni 76,000 36,000 44 18 5 Cr, 10 Fe, 75 Nj, 10 Mo *Data obtained from the Superior Tube Company. Nickel-Molybdenum-Base Alloys The nickel-molybdenum-base alloys are also being studied as part of the effort to obtain alloys with better high-temperature strength and corrosion re- sistance than are offered by Inconel. These re- guirements are met by Hastelloy B; however, some difficulties have been encountered in attempts to fabricate this material. In castings and in fusion welds, o whole range of composition results because of severe coring during freezing, and thus the properties of the fobricated specimen vary from one area to another. Even long-time high-temperature aging treatments of such material will not produce an equilibrium structure. The aging treatment also results in o great increase in hardness with a cortesponding decrease inductility. At temperatures between 1200 and 1800°F, the wrought material exhibits its poorest ductility because of hot short- ness, which can arise from trace impurities or the precipitation of an age-hardening constituent, which is probably the case for Hastelloy B. Thus the temperature range of poorest ductility of Hastelloy B coincides with the temperature range of con- templated use in high-temperature circulating-fuel Hastelloy B is not recommended for use in air at temperatures above 1400°F because of reqactors, excessive oxidation. A potential source of trouble exists because of the molybdenum content; molybde- num oxide has been reported to cause catastrophic oxidation of surrounding structures, especially the steels. The solutions to these problems are being ap- proached in two major ways. First, attempts are being made to find suitable melting and heating treatments of Hastelloy B that will reduce the degree of age hardening so that both the cast and wrought materials will possess acceptable properties in the temperature range of interest. Second, Hastelloy-type alloys are being siudied. Both phases of the investigation will involve attempts to extrude seamless tubing, since, at present, Hastelloy B tubing must be made by welding strips. Aging studies of the following materials are now in prog- ress at temperatures between 1200 and 1700°F: commercial, wrought, Haostelloy B plate; vacuum- melted Hastelloy B, as cast; cast and wrought 20% Mo—80% Ni alloy; wrought 24% Mo—76% Ni alloy. Some of the properties of the 20% Mo--80% Ni alloy already obtoined indicate that this alloy will be easier to fabricate than Hastelloy B from an extrusion standpoint; sound tubing could be made, whereas none was obtained from Hastelloy B, The recrystallization temperature of a 60% cold- worked sheet of the 20% Mo-80% Ni alioy was ap- proximately 1900°F for an annealing time of 30 min. However, this alloy oxidizes in air at 1500°F at a faster rate than does Hastelloy B, although under a temperature cycle between room temperature and 1500°F the oxidation rates are similar. The scale formed spalis near rocom temperature. A sheet, 0.065 in. thick, could be heliarc welded without producing porosity, but an argon arc pro- duced porosity. In strength tests at 1500°F, the 20% Mo~—80% Ni alloy had a rupture life of 90 hr ot 8000 psi; when tested at 5000 psi for 600 hr, it had a 3% elongation; during a short-time test the alloy had atensile strength of 40,000 psi and an elongation of 10% in a 2-in. gage length. Oxidation and oxidation-protection studies are in nrogress to find suitable methods of protecting the nickel-molybdenum-base alloys during tests in ther- mal- convection loops and creep-testing apparatus. This work will be the ground work for further studies of the protection of reactor components, The coat- ings being studied are those that can be applied by methods readily available, that is, chromium electro- plate, chromium plus nickel electroplate, nickel PERIOD ENDING SEPTEMBER 10, 1954 plus aluminum spray, and aluminum spray. In constant-temperature tests, g chromium electroplate was found to be beneficial. In oxidation tests in- volving «cycling between room temperature and 1500°F, these coatings eliminate the spalling characteristics of Hastelloy B. The chromium elec- troplate and the chromium plus nickel electroplate afford the best protection. Duplex Tubing Attempts have been made to fobricate duplex seamless tubing that will have good corrosion resistance on the inner surface ond oxidation resistance on the outer surface. In the first ex- periments, attempts were made to deep draw duplex blanks 4 in. in diometer. The blanks were made by hot pressing and also by hot rolling. The following materials were combined by hot pressing 0.025-in,- thick sheets under an argon atmosphere with Al,0,- coated graphite dies at 1500 psi. Pressing Temperature (°Q Copper plus type 310 stainless 1000 steel Inconel plus type 310 stainless - 1200 steel Inconel plus Hastelloy B 1200 Hastelloy B plus type 310 stajnless 1200 steel Copper was also combined with type 310 stainless steel by hot rolling at 1000°C, For this, a three-ply composite with copper in the center and cover plates of the stainless steel was used. One of the cover plates was oxidized ot 1100°C for 2 hr to prevent bonding during rolling. | included in these experiments were some tests involving the fabrication of composites of molybde- num and Inconel. The purposes of these tests were to determine the conditions necessary to obtain a metallurgical bond by hot rolling, to determine the extent of directional properties due to the molybde- num, and to determine whether such a combination could be deep drawn. The first rolling experiments were made on an evacuated copsule at 1225°C with reductions of 10% per pass. The total reduction was 79% in thickness. No bonding was obtained by using nickel os an intermediate layer between the 123 ANP QUARTERLY PROGRESS REPORT Inconel and the molybdenum. In o second series of experimenis the rolling temperature was reduced to 1000°C, with reductions of 40% per pass and a total reduction of 85% in thickness., The lower rolling temperature reduced the tendency for inter- metallic compounds to form af the interface. Bonding was achieved by using nickel as the intermediate layer, and evidences of intermetallic compounds were found only by metallographic examination af high magnifications. The final thickness of the composite was 0.010 in. of molybdenum on 0.040 in. of Inconel. bination, the maximum ratio of molybdenum to Incone! without the directional properties of wrought molybdenum will be determined. All the composites mentioned above were suc- cessfully reduced from 4-in. blanks to 2‘/2~in. cups by the ‘“Guerin’’ forming process (male parts of die are hard rubber) at room temperature, Attempts to deep draw the composites are to be made. In future experiments with this com- Sigma-Phase Alloys For the continuing study of sigma-phase alloy corrosion by liquid lead (cf. Sec. 6, ‘“Corrosion Research’’), some additional tubing was made., Two ingots of a 48.2% Cr—51.8% Fe alloy and two ingots of o 45% Fe~40% r-15% Ni dlloy were vacuum cast into 1]/2-in.-did ingots 6 in. long. These ingots were then hot rolled in an air atmosphere at 1250°C to rod 5/B in. in diameter. After onnealing at 1100°C for 2 hr and quenching, the rod was readily machin- able info tubing. The transformation from the ferrite to the sigma phose will be accomplished by aging at 1425°F, Two of the tubes (Fe-Cr) were cold swaged 12% so that the effect of cold swaging on the distribution of the sigma phase could be determined. Boron Carbide Shielding The experimental fabrication of !fs-in.-fhick shield pieces by warm pressing boron carbide bonded with copper or silver has been discontinued in favor of the fabrication of pieces molded with nonmetallic bonding material by cold pressing followed by sintering to develop the desired properties. Non- metallic bonding materials that would be suitable from nuclear considerations and would be stoble af the expected service temperature include sodium silicate, silica, silicon nitride, and boric oxide. Arrangements are being made to secure samples of these materials for compatibility tests with Inconel under operating conditions. 124 The density of the proposed shield pieces is to be about 1.8 to 1.9 g/cm? of B ,C, equivalent to 1.4 to 1.5 g/cm?® of boron. They would be molded in the form of equilateral triangles or diamond shapes, 2 to 3 in. on a side, so as to cover a sphere with an integral number of pieces, Additional tests of the compatibility of Incone! with pure boron carbide are being made. Previous tests indicated the formation of a diffusion layer 5 mils thick in a 100-hr test ot the operating temper- ature of 1500°F, Tubular Fuel Elements Twelve tubular fuel elements are being prepared for further drawing experiments at Superior Tube Company. They are being assembled with high- fired UQ, in.the 30- to 55-p particle size range, as well as with very fine UO,, and with prealloyed and elemental stainless steel and iron core matrixes. Plans are also being made to try hot swaging of the tubes in an attempt to reduce stringer formation in the cores by reducing the amount of cold working. Hot swaging on a mandrel moy prove to be suc- cessful if bonding to the mandre! can be prevented by an oxide scole or coating. METALLOGRAFPHIC EXAMINATION OF A FLUORIDE-TO-SODIUM HEAT EXCHAMGER R. J. Gray P. Patriareca G. M, Slaughter Metallurgy Division The fluoride-to-sodium intermediate heat ex- changer, which in a life test failed after 1680 hr in cyclic service in the temperature range 1080 to 1500°F, was examined metallographically. All the tube-to-header joints were manually heliarc welded, and tests showed the heat exchanger to be helium leak-tight before assembly into the test rig. Visual examination indicated the presence of 19 fissures in the tube-to-header welds in two of the three adjacent headers at the hot sodium inlet end. The probable progressive propagation of these fissures is shown in Figs. 7.14, 7.15, and 7.16, The fissure shown in Fig. 7.16 probably extends to the surface of the weld, If so, there was a leak in the system at this point. It seems likely that dif- ferential thermal exponsion between the tubes and the casing caused stress concentrations at the roots of the tube-to-header welds. concentrations These stress would tend to propagate cracks through the welds in the course of thermal cycling, PERIOD ENDING SEPTEMBER 10, 1954 UHGLASSIFIED E y.raoe Fig. 7.15. Téube-io-Hender Joint witha $light Crack Extending into;the Weld, Etched. 100X, Reduced 16%. 125 ANP QUARTERLY PROGRESS REPORT UNCLASSIFIED Y-13004 Fig. 7.16. Tube-to-Header Joint Showing Severe Cracking. Etched. 100X. particularly since the columnar dendrites, which are typical of a weld structure, are aligned in such a way as to aid parallel fractures, This investigation emphasizes the extreme de- sirability of using back-brazing as a means for minimizing the notch effect and for reducing the possibilities of leaks developing in the system, 126 Corrosion tests in static sodium and in fluoride mixtures indicate that a 67% Ni-13% Ge-11% Cr—-6% Si~2% Fe-1% Mn alloy may be useful for this ap- plication. This alloy, which flows well at 2050°F, was found to be virtually unattacked in the fluoride fuel and corroded o a maximum of 4 mils in 100 hr at 1500°F in sodium. PERIOD ENDING SEPTEMBER 10, 1954 8. HEAT TRANSFER AND PHYSICAL PROPERTIES H. F. Poppendiek Reactor Experimental Engineering Division The enthalpies and heat capacities of NoF-ZrF - UF, (65-15-20 mole %) were determined; the heat capacity in the solid state over the temperature range 90 to 614°C was found to be 0.17 cal/g-°C, and the heat capacity in the liquid state over the temperature range 653 to 924°C was found to be 0.20 cal/g-°C. The enthalpies and heat capacitiés of LiF-NqF-UF4 (57.6-38.4-4.0 mole %) were also obtained; the heat capacity in the solid state over the temperature range 97 to 594°C was found fo be 0.23 cal/g-°C, and in the liquid state over the range 655 to 916°C it was found to be (.53 cal/g°C. The thermal conductivities of three fluoride mixtures in the solid state at normal temperatures were determined. The conductivity of NaF~KF—UF4 (46.5-26.0-27.5 mole %) was 0.7 Btu/hr-ft? (°F/#), that for KF-LiF-NaF-UF, (43.5- 44.5-10,9-1.1 mole %) was 2.0 Btu/hr-ft?2 (°F/ft), and that for LEF-KF-UF4 (48.0-48.0-4.0 mole %) was 1.4 Btu/hr-f12 (°F/ft). A new electrical con- ductivity device has been constructed and has been successfully checked with molten salts of known conductivity. Additional forced-convection heat transfer meas- urements of molten NaF-KF-LiF (11.5-42.0-46.5 mole %) have been made. Measured thermal con- ductivities and thicknesses of insoluble deposits on the inner walls of Inconel heat transfer tubes made it possible to calculate the thermal re- sistance of the deposit. The calculated value was in good agreement with values previously deduced from the heat transfer measurements. A device for studying the rates of growth of tube-wall de- posits has been successfully tested with a simple heat transfer salt. A hydrodynamic flow system to be used for studying the reflector-moderated reactor flow structure has been tested. A mathe- matical study of the temperature structure in can- verging and diverging channel systems that duct fluids with volume heat sources has been made. A study of wall-cooling requirements in circu- lating-fuel reactors was made. 'W. D. Powers and G. C. Blalock, Heat Capuacities of Compositions No. 3% and 101, ORMNL CF.54-8-135 (te be issued). PHYSICAL PROPERTIES MEASUREMENTS Heat Capacity W. D. Powers Reactor Experimental Engineering Division The enthalpies and heat capacities of twe fluo- ride mixtures have been determined with Bunsen ice calorimeters:! NaF-ZrF4—UF4 (65-15-20 mole %) Solid (90 to 614°C) fHy - Hpoe = =3 + 07T C, = 0.17 + 0.01 Liquid (653 to 924°C) Hy - Hypo = 22 + 0,207 Cp = (.20 * 0.02 LiF-NaF-UF (57.6-38.4-4.0 mole %) Solid (97 to 594°C) Hy — Hyor = 0.22(7)T + 0.0001772 C, = 0.22(7) + 0.000337 b Liguid (655 1o 916°C) Hp = Hpo = ~68 + 0.53T CP = 0.53 + 0.04 In these expressions H is the enthalpy in cal/g, CP is the heat capacity in cal/g-°C, and T is the temperature in °C. The experimental enthalpy data for LiF-NaF-UF, (57.6-38.4-4.0 mole %) are plotted in Fig. 8.1. Automatic Simplytrol units have been added to the calorimeter systems to more accurately control the furnace temperatures. This modification has increased the precision of the experimental data. Currently, enthalpies of a fluoride mixture are being measured with the copper block calorimeter, Density ond VYiscosity S. 1. Cohen Reactor Experimental Engineering Division Preliminary density measurements were made on rubidium metal, and the density was found to be represented by the equation plg/em®) = 1.52 — 0.00054 (T — 39°C) , 127 ANP QUARTERLY PROGRESS REPORT ORNL - LR-0OW6G 2747 450 [ 400 |- o : Co 350 | | 300 | = e e e e s ' | | i Jo——— b i | i : | 250 | - 200 { | ENTHALPY (cai/g) {OO S e e 400 600 80C TEMPERATURE ({°G) Fig. 8.1. Enthalpy vs Temperature for LiF-NaF-UF , (57.6-38.4-4.0 mole %). where T is the liquid temperature in °C. Measure- ments were made by the buoyancy principle. A calculated valve for the liquid density at the melting point, given in the Liguid Metals Hand- book,? falls within 3% of the value yielded by this study. This experiment was carried out at the request of the ORSORT group currently studying the use of rubidium vapor in an aircraft reactor cycle. Measurements were made on a sample of metal 2R. N. Lyon (ed), Ligquid Metals Handbook, p 52, AEC-Department of the Navy, NAVEXOS P-733 (rev), June 1952, 128 produced by the Stable Isotope Research and Pro- duction Division by reduction of a high quality fluoride which was prepared by the ORSORT group from RbF furnished by the Materials Chemistry Division. In order to obtain more accurate viscosity data it was decided to initiate a program of refining viscometry techniques. In particular, the influence of fluid density on calibrations of both the ro- tational and the capillary devices is being studied. Easily hondled fused salts are being used as calibrating fluids. The present studies indicate that some of the old calibration curves, which were based on low-density calibrating liquids, were shifted by about 30%. This means that some of the previously reported viscosity measurements are about 30% high. : Becaouse of significant improvements made in fluoride preporation ond handling, the mixtures which are now being obtained from the Materials Chemistry Division are exiremely pure, and they no longer cause the difficulties encountered in the early stages of the project, such as fouling of equipment and variations in physical character- istics. No doubt some of the earlier viscosity measurements were high because of the impurities in the materials. Thermal Conductivity W. D. Powers Reactor Experimental Engineering Division The thermal conductivities of three solid fluoride mixtures ot normal temperotures were determined by the transient cooling technique: Thermal Conductivity {Btu/hr-#2 (OF /i) NaF-KE-UF , (46.5-26.0-27.5 0.7 mole %) KF-LiF NaF-UF , (43.5-44.5- 2.0 10.9-1.1 mole %) LiF-KF-UF , (48.0-48.0-4.0 1.4 mole %) The thermal conductivity of solid LiF-KF-UF, (48.0-48.0-4.0 mole %) was measured by the steady- state flat-plate technique; a value of 1.5 Btu/hr.ft? (°F/ft) was obtained. The ratios of corresponding liquid to solid thermal conductivity measurements for the fluoride mixtures studied to date are near unity. Electrical Conductivity N. D. Greene Reactor Experimental Engineering Division The experimental current-potential conductivity cell has been successfully tested and stand- ardized. The electrical conductivity of a KNO, melt was determined to within a few per cent of the values reported in the literature. As predicted, the effects of polarization within this cell were greatly reduced by the use of the alternate method.® This reduction in the effect PERIOD ENDING SEPTEMBER 10, 1954 of polarization was further substantiated by the small observable increase of conductivity with frequency in contrast to the usual large depend- ence of conductivity upon frequency, as observed in the low-resistance conductivity cell. The measurements on KNO. were obtained during a single run without the necessity of having to replatinize the electrodes after each determination. A refinement of this current-potential cell which will permit measurements of greater accuracy is now under way. Preliminary measurements of several fluorides, the conductivities of which were determined previ- ously by an alternate method, were found to be in good agreement within the inherent limitations of accuracy. A summary of the preliminary measure- ments of the conductivities of KF-LiF-NaF-UF, (43.5-44.5-10.9-1.1 mole %), NaF-ZrF -UF, (50.0- 46.0-4.0 mole %), and NaF—'ZrF4-UF4 (53.5-40.0- 6.5 mole %), as well as the medsurements of several other salts, has been compiled in o sepa- rofe reporh"" Vapor Pressure R. E. Moore Materials Chemistry Division Yapor pressure measurements on the mixture N0F¢ZrF4 (25-75 mole %), which were begun last quarter,” have now been completed. The method and apparatus, originally described by Rodebush and Dixon,® were discussed in previous reports.” 8 The data, given in Table 8.1, can be represented by the equation ' 9368 T(°K) from which the calculoted pressures and the heat of vaporization {43 kecal/mole) were obtained. The approximate liquidus temperature, obtained from log,q P(mm Hg) = - + 10.57 , 3N. D. Greene, ANP Quar. Prog. Rep. fune 10, 1954, ORNL-1729, p 100. N, D. Greene, Measurements of the Electrical Con- ductivity of Molien Fluorides, ORMNL CF-54-8-64 {to be issued). R, E. Moore and C. J. Barton, ANP Ouar. Prog. Rep. June 10, 1954, ORNL.1729, p 101, SW. H. Rodebush and A. L. Dixon, Phys., Kev., 26, 851 (1925). 7R. E. Moore and C. J. Barton, ANFP Quar. Prog. Rep. Sept. 10, 1951, ORNL-1154, p 136, BR. E. Moore, ANP Quar. Prog. Rep. Dec. 10, 1931, ORNL-1170, p 126. 129 ANP QUARTERLY PROGRESS REPORT TABL.E 8.1. YAPOR PRESSURE IN NaF-ZrF4 (25-75 mole %) SYSTEM Temperature Observed Calculated ©C) Pressure Pressure (mm Hag) (mm Hg) 754 29 28 769 37 38 776 44 44 786 54 54 796 66 65 BDS 74 76 311 83 83 828 115 115 844 152 155 the intersection of the vapor pressure curve with that of pure ZrF ,, is about 760°C. FUSED-SALT HEAT TRANSFER H. W. Hoffman Reactor Experimental Engineering Division Additional data for NaF-KF-LiF (11.5-42.0-46.5 mole %) flowing for short-time periods in an elec- trically heated type 316 stainless steel tube have been analyzed. The results are in agreement with the general turbulent flow correlations for ordinary tluids. The thermal conductivity of K,CrF, was found to be 0.13 Btu/hr-ft? CF/f)° in the temperature range 100 to 200°F, Measurements from a photo- micrograph, Fig. 8.2, of a section of the tube used in the Incone! experiment show o film thickness of about 0.4 mil. The thermal resistance calcu- lated from these results, 0.00025 hr-ft2.°F /Btu, is in agreement with the measured thermal resistance obtained from the heat transfer experiments in the Inconel system. The forced-convection apparatus,'® which is to be used for studying the rate of growth of wall deposits, as well as for heat transfer measure- ments of fused fluoride mixtures, has been operated successfully for more than 100 hr with the heat 4. W. Hoffman and J. Lones, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 101, 1044, W, Hoffman and J. Lones, ANP Quar. Prog. Rep. Mar, 10, 1954, ORNL.-1692, p 98. 130 FILM; 0.4 mil ‘~—~—I Fig. 8.2. K3CrF6. incone! Tube with Film Deposit of transfer salt NaNOz-NqNfia-KNO3 (40-7-53 wt %) as the circulated fluid. The system is now being modified so that resistance-hected, as well as externally heated, test sections can be inserted in the system. With resistance-heated sections, larger axial fluid temperature rises can be at- tained. It is planned to study next a zirconium- uranium fluoride salt mixture in lnconel. TRANSIENT BOILING RESEARCH M. W. Rosenthal Reactor Experimental Engineering Division A study of the heat transfer phenomena which occur when a water-cocled and -moderated solid- fuel-element reactor suddenly becomes super- critical has been undertaken. The time required for generation of steam in water which is initially below the boiling point will be determined. The effects of water temperature, rate of power in- crease, air concentration, and surface condition will be investigated. Construction of the equipment for generation of power surges by electrical means in aluminum fila- ments has been completed, and the entire experi- mental apparatus has been assembled and tested. Satisfactory operation has been achieved. The transient-power generator consists of 100 thy- ratrons connected in parallel. A different bias can be put on the grid of each thyratron, and the system can be set so that a signal of increasing voltage applied simultaneously to all tubes fires them in a sequence that approximates the desired A peak current of about 500 amp can be produced, with the duration of the power surge ranging from 30 to 500 msec. function. The instantanecus values of voltage drop across the test filament and current through it are re- corded by a fast-response Hathaway oscillograph. A TFastax high-speed motion picture camera is used to photograph the event, and illumination is supplied by an Edgerton stroboscopic flash unit synchronized with the camera. A repetition rate of 6000 frames per second is obtainable with a flash duration of 2 usec. Closing a single switch initiates the operation of all the equipment; a time delay is provided in which the camera can ac- celerate to full speed before current is passed through the filament. FLUID FLOW STUDIES FOR CIRCULATING- FUEL REACTORS J. O. Bradfute L. D. Palmer F. E. Lynch Reactor Experimental Engineering Divisieon The flow system for the hydrodynamic studies of the reflector-moderated reactor core, shown sche- matically in Fig. 8.3, has been fabricated, as- sembled, and tested. A detailed sketch of the test section is also shown in Fig. 8.3. Since it will be necessary to insert, photograph, and then re- move a coordinate grid from time to time, the system was designed so that it could be readily assembled and disassembled. The cores are po- sitioned and supported by a brass sleeve, which was attached to the test section support along its center line before it was machined. The sleeve was carefully bored and lopped so that the shafts protruding downward from the cores would fit with minimum clearance and yet loosely enough so that they could be removed by hand. The upper surface of the test section support flange was machined perpendicular to the center line of the brass sleeve. This arrangement permits accurate align- ment of the core within the Plexiglas test section without supporting members in or ahead of the test region. Attempts are currently being made to obtain some preliminary fluid flow information for the reflector-moderated reactor core for the straight entrance condition shown in Fig. 8.3 by photo- graphing particle flow through the Plexiglas test section. The phosphorescent flow-visualization system, which was described previously,!! has been further modified. The ZnCd sulfide phosphor has been replaced by a ZnS (Cu-activated) phosphor whose luminosity was found to be greater after excitation. The duct work in the flow circuit has been im- 11 . D. Palmer and G. M. Winn, A Feasibility Study of Flow Visualization Using a Phosphorescent Particle Method, ORNL CF-54-4-205 (Aprit 30, 1954). PERIOD ENDING SEPTEMBER 10, 1954 proved so that higher Reynolds numbers are now attainable, and a diverging lucite channel has been constructed. Some of the flow features, such as asymmetry, flow separation, and the effect of entrance lengths, are to be observed for different entrance configurations in such channel systems. The phosphorescent flow-visualization method is also to be used in the reflector-moderated reactor core experiment. HEAT TRANSFER STUDIES FOR CIRCULATING- FUEL REACTORS H. F. Poppendiek L. D. Palmer Reactor Experimental Engineering Division The classical hydrodynamics study, conducted by Nikuradse,'? of converging and diverging channels was reviewed. It is felt that some of the fundamental flow features observed in con- verging and diverging channel systems will alsc exist in the reflector-moderated reactor. The results obtained by Nikuradse indicate that for the case of turbulent flow in convergent channels the flow becomes more stable or less turbulent in nature and that the eddy ditfusivities are signifi- cantly lower than in a parallel-plate channel system. For the case of turbulent, symmetrical flow in diverging channels the tlow is found to be more turbulent than in the parallel-plate system. When the total channel angle of a divergent system is increased beyond 8 deg, asymmetrical flow re- sults. For o 10-deg angle, a thin, low-velocity layer of fluid flowing in a reverse direction to the main stream is noted near one of the channel walls. This separation phenomenon becomes more pronounced at larger channel angles. Some radial temperature distributions were calcu- lated in the established-tlow region of a few con- verging and diverging flow channel systems con- taining volume heat sources within the ducted fluide The experimentel velocity and eddy dif- fusivity data obtoined by Nikuradse were used in the analyses. The results indicated that wall-fluid temperature differences can be significantly higher in converging flow channels than in parallel-plate systems; the converse is frue for symmetrical diverging flow channels. Asymmetrical divergent flow passages are difficult to analyze at this time because the diffusivity structure is not known. 12,, Nikuradse, Forschungsarbeiten YDl 289, 1-49 (1929). 131 el UNCLASSIFIED ORNL—LR ~DWG 2749 + L MANOMETER UPPER PIPL FROM WATER MAIN ——= % BUILDING WATER MAIN COLLAR — PARTICLE SLURRY STORAGE ROTAMETER ENTRANCE SECTION——= ~ @é / T T i A o : / - N TEST SECTION — / /\ \ CORE : \ : z-ni i } ! S \ 1 / TEST SECTION \\‘ 7 BRASS SLEEVE y TEST SECTION SUPPORT—— = LOWER PIPE TO DRAIN—== TO DRAIN Fig. 8.3. Flow System for Hydrodynamic Study of Reflector-Moderated Reactor Core. 14Oy SSFYO0Ud ATIFLIVNO NV However, it appears reasonable to conclude that high temperatures probably would exist in the thin, low-velocity layer of fluid in the 10-deg divergent channel because the flow there is probobly laminar in nature. A report is being prepared which will be useful to designers for predicting wall-cooling require- ments in circulating-fuel reactors which have pipe, parallel-plate, or annular fuel-duct geometries. The temperature solutions, which were presented previously,”"” have been tabulated in detail so PERIOD ENDING SEPTEMBER 10, 1954 that the superposition process outlined in those reports can be quickly effected. Cooling require- ments and resulting fluid temperature distributions have been determined for typical circulating-fuel reactor systems. By F, Poppendiek and L. D. Palmer, Forced Con- vection Heat Transfer in Pifies with Volume Heat Sources Within the Fluids, ORNL-1395 (Nov. 5, 1953). 4y F Poppendiek and L. D. Palmer, Forced Con- vection Heat Transfer Between Parallel Plates and in Annuli with Volume Heat Sources Within the Fluids, ORNL-1701 (May 11, 1954). 133 ANP QUARTERLY PROGRESS REPORT 9. RADIATION DAMAGE J. B. Trice Solid State Division A, J. Miller ANP Project Additional irradiations of lnconel capsules cone taining fluoride fuels were carried out in the MTR, Only one capsule containing a UF j-bearing fuel has been examined, ond it shows no corrosion, This is in contrast to the capsules containing UF ,-bearing fuel in which minor increases in penetration have at times been observed in the irradiated specimens as compared with the out-ofs However, it should be emphasized that even in the UF ,-bearing capsules, the ap- pearance of small deviations from the controls could be due to experimental difficulties. Inspection of the previously described LITR in- pile circulating-fuel loop, which failed prior to the startup, disclosed that the failure was caused by a bregk in the weld connecting the pump discharge nipple to the fuel line, Design revisions and refabrication of some components are in progress. pile controls. Developmental work continued on a smaller loop for operation in an LITR A-piece. Work on a loop for insertion in the MTR is described in Sec. 3, ‘‘Experimental Reactor Engineering.” Development of equipment continued for in-pile stress-corrosion tests on Inconel in contact with fluoride mixtures, and testing of the MTR tensile- creep rig neared completion. Detailed examination was completed of an Inconel loop in which sodium was circulated at high temperature in the ORNL Graphite Reactor, and no evidence of radiation- induced corrosion was found in the Inconel, Metallographic examinations were carried out in the hot cells on irradiated fuel plates for the Pratt & Whitney Aircraft Division, and studies were made on annealing.out of fission-fragment damage. Work continued on examination of wire and multiple- plate type units for GE-ANP, MTR STATIC CORROSION TESTS W. E. Browning G. W. Keilholtz Solid State Division A new series of capsules was irradiated in the MTR to compare NaF-ZirF, fuels containing UF, with those containing UF,. The property to be compared is the effect of fission on static cor- 134 rosion of Incone! by the fuels at 1500°F. The series includes five capsules containing 1,79 mole % UF,, 49.3 mole % ZrF,, and 48.9 mole % NaF and five capsules containing 1.74 mole % UF,, 48.2 mole % ZrF,, and 50.1 mole % NaF. Analyses are being run to confirm the purity of the UF; with respect to UF, contamination. For every capsule, there will be an out-of-pile control capsule. Each fuel generates 1100 w/cm® in the A-38 position in the MTR, Six capsules have been irradiated for the speci- fied two-week period. Five have been received at ORNL after irradiation; three of them have been opened and two have been examined - one con- taining UF ; and one containing UF ,. Bothcapsules had been irradiated with the metal-liquid interface at 1500 + 50°F., The UF;-bearing capsule had experienced one temperature excursion to 1660 + 50°F for less than 30 sec, but the temperature of the UF ,-bearing capsule never exceeded 1550 + 50°F. The fuel in each capsule was frozen two or three times during its history. No evidence of corrosion or film could be seen in the UF j-bearing Inconel There was subsurface void formation in the UF ;-bearing Inconel capsule to a depth of about 2 mils. No intergranular corrosion or concentration of voids at the grain boundaries was evident, Analytical samples of the irradiated fuel were taken with two sizes of drills. The uranium concen- trations identical, within experimental error (+3%), and were of the wvalue expected after burnup correction. Analyses for iron, chromium, and nickel gave the following results: Fe, 0.03 to 0.14 wt %; Ni, 0.01 to 0.05 wt %; Cr, less than 1074 wt %. There were no significant differences in the final iron, nickel, capsule, in the ceniral and outer samples were and chromium compositions between the two cap- sules. Analyses of the unirradiated fuels have not been completed. It would be unwise to draw conclusions on the basis of only one pair of samples; however, if these first results are confirmed in future capsules, UF, fuels may be considered quite desirable as far as radiation-induced corrosion is concerned. Two UF;-bearing capsules and two UF ;-bearing capsules in this series will be irradiated for six weeks each to amplify the apparent differences in the effects of UF; and UF, on Inconel. It is inter- esting to note that the UF ; capsule in this series is the first one run in the MTR and the first one run at 800 w/cm? or greater that has not shown a tendency toward intergranular corrosion. - Whether the differ- ence between eorlier UF -bearing capsules and the one in this series is due to improved temper- ature control or to uranium concentration effects will be determined by additional irradiations. FISSION PRODUCT CORROSION STUDY C. C. Webster G. W. Keilholtz Solid State Division Steps are being taken to perform corrosion tests out-of-pile with irradiated fuel in order to separately study the effects on Inconel of fission-product concentration and of Inconel irradiation. A water-cooled facility for irradiation of solid fuel was constructed and has been installed in position C-46 of the LITR., The unperturbed thermal flux is on the order of 4 x 1073, An Inconel tube has been constructed for casting solid bars of fuel 1 in, long and 0.1 in. in diameter. The solid bars of fuel will be transferred to Inconel capsules for irradiation. The capsule has been constructed so that it can be welded shut without the bars being melted and can be reopened in the hot cell for transfer of the fuel to the corrosion test capsule. Upon completion of an out-of-pile heating cycle, the test capsule will be opened in the hot cell. The fuel will be drilled out and divided into three samples: a portion for petrographic study to detect any oxidation or reduction, a portion for mass spectrographic study to determine burnup, and a portion for chemical analysis of the fuel. The capsule will be slit on the remote slitting machine described below for a metallographic study. At least two concentric samples can be taken from a capsule of this size by remote means. FACILITIES FOR HANDLING IRRADIATED CAPSULES C.C. Wébster G. W. Keilholtz Solid State Division it was observed metallographically, in some that when transverse sections were taken from a single fluoride fuel corrosion cases, several PERIOD ENDING SEPTEMBER 10, 1954 capsule the corrosion pattern was not consistent over the length of the capsule but appeared to be a function of the longitudinal temperature gradient. One method of obtaining a longitudinal section would be to grind away half of the capsule on the milling machine. However, by such a methed it would be difficult to ascertain that a section through the diameter of the capsule had been made. Also, the milling operation would create a serious contamination problem, The method being used consists in passing the capsule longitudinally by a 10-mil-thick silicon carbide fine-grit wheel and thus slitting the cap- sule down the center. The capsule is clamped into a vise and then fed into the wheel by means of a mechanical linkage which is to be replaced by a variable-speed motor and o contact cutoff switch, When the capsule is put into the gripping adapter in the vise, it is automatically aligned so that the wheel will cut in the plane of o diameter through the capsule, By making different gripping adapters, various sizes of tube can be slit lengthwise. Thus far the apparatus has been used on only MTR-type fluoride fuel capsules (0.100-in.-ID and 0.200-in.~ 0OD); it has worked very satisfactorily. : Since the abrasive wheel has a rubber base, it must be kept cool. The darea around the cut must also be kept cool to prevent any high thermal stresses from affecting the corrosion results. The wheel and specimen are over a tray and are covered with a splash shield so that a liquid coclant can be used. The coolant level is always kept above the bottom edge of the abrasive wheel so that by cutting upward into the specimen the coolant is carried into the cut; therefore the specimen need not be in the coolant, Carbon tetrachloride is used as the coolant because it does not react with the fluoride fuel at high temperatures. It is forced into the tray from outside ‘the hot cell through Tygon tubing by air pressure. The tray and splash shield are attached to the mount for the vise and they move with the specimen. Once installed within the hot cell, the remotely operated tube-slitting machine can be used with ease in cutting longitudinal sections from various sizes of tubing. Thicker abrasive wheels can be used for heavier walled tubing. The advantages of this method of cutting are that it permits visual observation of the amount of fuel left on the tube walls, it permits observation of constrictions without the necessity of removing or disturbing the 135 ANP QUARTERLY PROGRESS REPORY material causing the trouble, it gives o well-exposed surface for chemically removing the last traces of fuel, it permits a cut to he made through the bottom plug of the capsule so that an examination can be made for crevice corrosion, and it leaves nearly one half of the capsule for further and more com- plete study, if desired. After the fuel is removed chemically without attacking the metal wall, the specimen can be electroplated so that the fuel- metal metallographic polishing operation. The stability of the fivoride fuel under irradiation and the extent of the corrosion of the container by interface will not be rounded during the the fuel are determined, in part, by analyzing samples of the irradiated fuel for the concentration of the original components ond for the major con- stituents of the container material. The fuel samples are put into solution within the masier- slave hot cells and are then transferred to the Analytical Chemistry Division for analysis. Work on the necessary transfer shields has been coordi- nated with the design and construction of the lead barrier being provided in the analytical chemistry hot cells to improve transfer conditions and to reduce personnel exposure. Two transfer carriers weighing 150 Ib each and giving 2 in, of lead shielding have been constructed and are now in use. FEach carrier holds four 30-ml bottles that are held in place by spring clamps. A motor-driven decapper for removing the bottle cap and a remotely controlled pipet thai can be lowered into the bottle are provided. The bottles may be inserted into the carrier within the cells and can be left in the caorrier during all sample removal operations. The whole unit moy then be returned to the hot cells for reuse. Two lead storage units weighing 450 |b each and giving 3 in. of lead shielding were also built and are available for use when all the transfer carriers are in use. The lead storage units have a capacity of four 50-ml volumetric flasks and can be loaded within the master-slave cells. ANALYSIS OF iRRADIATED FLUORIDE FUELS FOR URANIUM M. T. Robinson G. W. Keilholtz Solid State Division One of the regular steps in in-pile static cor- rosion testing of fluoride fuels has been the anal- ysis of the irradiated fuel for uranium. However, there are several major problems connected with 136 sampling operations, described previously,'s? which have never been fully resolved. In the cutting and drilling operations it is possible for small chips of metal, either from the capsule or the drill, to get into the fuel sample. To minimize this possibility, drills used recently have been A search is also In the it was possible for dirt from the hot cells to inad- vertently enter the fuel samples during cutting and drilling. This possibility has been minimized in the newer equipment.? The transfer operations, in several of which the salt is exposed to con- tamination from hot cell dirt, pieces of rubber from manipulator grips, and the like, are all possible carefully honed to remove burrs. being conducted for new drill materials. early use of the present sampling technique, sources of contamination. The observation of opaque material, in some cases in large amounts, during petrographic examination of irradiated fuel samples® indicates that at one or the other of the stages discussed above foreign material may have been introduced info some of the samples. Such contaminants make the resulting uranium analyses u! 1 QV{G The samples obtained for chemical analysis have, at times, been very small, often being as small as 10 mg. They are weighed in tared weighing bottles on a conventional chain-type analytical balance, and, since they are so small, the precision of weighing is about 4% instead of the usual negligible amount. This lack of precision results in an obnormally high uncertainty in the uranium analyses. After being weighed, the samples are dissolved and the solutions are made up to volume in the hot cells. The resulting solutions often contain as little as 500 ppm of uranium, Such high dilution is undesirable, since the titer may change appreciobly on fong standing. It seems reasocnable to conclude that the uncertainty in uraitium analyses may be as high as +5% or more, even for the usually precise (+2%) potentiometric titration, An increase of ten times in sample size (easily attainable) would restore this technique to its normal precision. ]J. G. Morgan et al,, Solid State Div. Semiann. Prog, Rep. Aug. 3, 1953, ORNL.-1606, p 40, 2C. C. Webster and J. G. Morgan, Solid State Div. Semiann. Prog. Rep. Feb, 28, 1954, ORNL-1677, p 27. 3(3. D. White and M. T. Robinsen, Sclid State Div. Semiann. Prog, Rep. Feb. 28, 1954, ORNL-1677, p 28. Similar remarks apply, in part, to the mass spec. trographic analysis. The samples taken are weighed by the analysts on a microbalance to alleviate the effect of small sample size. However, the handling of the very small crucibles in the hot cells is riot so delicate an operation as might be desired. The possibilities of contamination entering the sample are high. No highly useful conclusions hove emerged from much of the analytical work done in the past, probably because of the difficulties described above. Recently, however, marked improvement has been made as a result of continved efforts to improve sampling techniques, sample sizes, and handling procedures, ond uranium analyses have been carried out in a satisfactory manner. Since the analytical results obtained recently show thot no measurable segregation of uranium takes place, the procedure of taking more than one sample from each capsule will be abandoned. This will allow very much larger samples to be obtained, will make the analyses more sure, and will minimize the effects of foreign additions to the salt. HIGH-TEMPERATURE CHECK VALVE TESTS W. R. Willis G. W. Keilholtz Solid State Division As part of the development program for a small in-pile pump loop for operation in a berylium A-piece in the LITR, an attempt was made to develop a high-temperature check-valve pump, Such a pump has the advantage of being o compact, completely sealed unit. A test rig was run with Inconel, stainless steel, and Stellite 25 check balls. None of these materials gave successful tests. Time of operation before failure varied from 30 min for the Inconel to 2 hr for the stainless steel. Except with the stainless steel balls, operation of the valves was not con- tinuous; they would occasionally stick in both open and closed positions. When failure occurred, it was abrupt, and therefore sticking of the valves was indicated rather than o gradual stoppage of the system, MINIATURE IN-PILE LOOP -~ BENCH TEST J. G. Morgan G. W. Keilholiz Solid State Division An exploded view of the in-pile pump described previously? is shown in Fig, 9.1, The RPM meter coil measures the rotation of the shaft on which PERIOD ENDING SEPTEMBER 10, 1954 is mounted the motor armature. The Graphitar bearing is mounted in the bearing casing above the impeller housing. The assembled loop is shown in Fig. 9.2. The pump contains an extra reservoir to enable metered volumes of salt to be pumped into the surge tank for calibrating the venturi. The furnace for heating the pump and reservoir is also From performance tests with this setup, pump speed vs head and pump speed vs flow have been determined. At 5000 rpm the flow was 4 fps, corresponding to a Reynolds number of 3000 in o 200-mil tube. At the 540 w/cm” generated in-pile, the temperature differential is expected to be about 100°F. shown, LIFE TESTS OM AN RPM METER, BEARINGS, AND A SMALL ELECTRIC MOTOR UNDER IRRADIATION J. G. Mergan H. E. Robertson G. W, Keilholtz Solid State Division Tests were conducted on an RPM meter, bearings, and o small electric motor as o part of the de- velopment of o loop for insertion in an LITR A- piece. In order to drive the pump, it is desirable to use a variable-spesed electric motor and to align the shaft with lubricated bearings. Also, o record of shaft speed would be desivable, The entire motor assembly will be just above the upper grid guide in the LITR and in an estimoted thermal flux of 5 x 1077 nevtrons/cm?.sec. Hole 53 of the ORNL Grophite Reactor provided a comparable flux intensity and was used for these tests, The in-pile apparatus, with a single lower bearing as in the second test, is shown in Fig. 9.3. A motor was mounted in g horizontal position with upper ond lower armature shaft bearings, In the first test two lower bearings were used. The motor and RPM meter were canned and inserted into the center line of the reactor. The variables measured were resistance of the motor winding, resistance of the RPM coil, the voltage applied to the motor, the current drawn by the motor, the temperature of the lower Fafnir bearing, the temper- ature of the motor field, and the speed of the motor shaft, The Delco ac-dec motor used is rated at 115 v, 0.9 amp, and }’15 hp and has insulated wires. Some substitutions of materials were made. In the first 4.!. . Morgan, Solid State Div. Semiosnn. Prog. Rep. Feb., 28, 1954, ORMNL-~1677, p 30, 137 ANP QUARTERLY PROGRESS REPORT UHMCLASSIFIED PHOTO 1273% g 0 RPM METER COIL @ BEARING CASING IMPELLER HOUSING Fig. 9.1. In-Pile Pump. Py UHTL ASSIFIED 1 PUOTG 12738 SURGE TANK (2 FURNACE B W T O W™ W ‘a PUMP AND EXTRA RESERVOGIR VENTURI rig., 2.2. Bench Test Loop. 138 ARMATURE - THERMOCOUPLt\\ e T SO PR O N, oA N \;\;\\\ \‘\ o \"\\ 2 NN o DN EARING, FAFNIR OW-8 MOTOR: DELCO, AC-DC, "5 hp, SERVICE NO. 504743t STATOR TAPE: GLASS BRUSH HOLDER PLATE: MICA RPM COIL: NO.'36 AW.G. CEROC, 200-5500 TURNS BEARING GREASE: RRG-2 (NLGI GRADE 2) ~STATOR \\THERMOCOUPLE PERIOD ENDING SEPTEMBER 10, 1954 UNCLASSIFIED ORNL-LR~DWG 20814 BEARING, NEW DEPARTURE 77R6 ) TLREM COIL INSTRUMENT LEADS Fig. 9.3. Motor and RPM Meter Test Assembly. test a brush holder plate of mica backed with a stainless steel plate was used. In the second test fired Lavite asbestos covered wire was used from the brush holder to the field coil. Glass tape was used around the field coils. The rest of the construction features were un- altered; that is, fish-paper-coil supports, plastic (paper-backed Bakelite) segment spacers, and cardboard armature shaft spacers were used. The upper bearing (New Departure 77R6) and lower bearing (Fafnir DW-6) were packed with California Research Corporation grease, identified as RRG-2 [NLG1 Grade 21. The RPM meter consists of a magnet fastened to the shaft which will rotate inside the gas-tight motor shell. A coil wound around a C-shaped iron core is mounted outside the shell. As the shaft rotates, pulses are generated in the coil and counted remotely by a frequency meter calibrated in rpm. The coil was wound with 36-g Ceroc 200 magnet wire covered with glass tape impregnated with Glyptol. Summary data on the test conditions are given in Table 9.1. The totors were examined, after irradiation, in the, hot cell, Assistance was given in hot cell examination by J. Noaks and J. Crudele, Pratt & Whitney Aircraft Division. A ‘withdrawal cask and its facilities were loaned by L. E. Stanford, General Electric Company. commutator TABLE 9.1. CONDITIONS OF TEST MOTOR AND RPM METER ASSEMBLY Test No, 1 Test No, 2 Motor load d-c voltage 60 33 Current, amp 0.45 0.20 Shaft speed, rpm 4500 4500 Fafir bearing 45 45 temperature, OC Motor field plate 65 40 temperature, °C Time of operation At 4500 rpm, hr 100 648 In flux,* hr 84 613 *Thermal flux, 5 % 10” neutrons/cmz-sec; ratio of fast to thermal neutrons, 1; gamma field, 4.4 r/hr. in the first test, motor load was abnormally great from the beginning and the field plate temperature When the motor failed after 100 hr, the stator temperature rose to 182°C while the power was kept on. The suspicion that there was a cocked upper bearing plate was confirmed upon postirradiation examination (Fig. 9.4). The sensing coil for the RPM meter functioned well; there was no resistance change or visual deterioration. was high. 139 ANP QUARTERLY PROGRESS REPORT BHCLASSIFED PHOTC 42903 Fig. 9.4. Motor lrradiated in the First Test. In the second test the motor operated as long as it would be expected to under normal conditions, However, postirradiation inspection shewed that the commutator segments had been loosened as a result of the deterioration of the plastic commutator segment spocers under irradiation (Fig. 9.5). The bearing grease proved to be adequate and was essentiolly unaffected, as evidenced by the constant power demand by the motor at constant rpm. The RPM coil also showed no resistance change or visible breakdown (Fig. 9.6). The brush wear appeared to be normal, A high-tempercture motor is being constructed for testing that will have glass-insulated-wire, mica, and high-altitude brushes. REMOVAL OF XENON FROM FUSED FLUORIDES M. T. Robinson . W. Keilholtz Solid State Division D. E. Guss United States Air Force W. A, Brooksbank Analytical Chemistry Division A program has been initiated to study removal of Xe'35 from fluoride fuels by means of helium 140 B uneLassIfIED | e PHOTO 12302 Fig. 9.5, Commutator Segments from Motor Ir- radiated in the Second Test. purging. A stainless steel capsule will be con- structed in a manner thot will allow bubbles of helium to pass through a filter plate and through the fissioning fuel in hole HB-3 of the LITR. Plastic mockups of the capsule are being used to study the characteristics of the helium flow. A four-channe! spectrometer has been constructed by the Instrument Department to monitor the Xe!3® concentration in the flowing stream of gamma-ray helium. CyF 6 UF, IRRADIATION W. E. Browning G. W. Keilholiz Solid State Division An experiment is being conducted to determine in a very rough fashion the effect of pile irradiation on a 20% solution of UF in C;F g This material is of interest tothe ShieldingSection of the Physics Division for simulating liquid fuels in dynamic shielding experiments, and the experiment is being conducted with the aid of personnel from that Fig. 9.6. Motor lrradiated in Second Test. division. Three nickel capsules were charged with the two components to make a 7-g solution in each. Normal uranium was used. Two capsules were irradiated in appropriate secondary containers in a water-cooled hole in the ORNL Graphite Reactor at a thermal flux of 6 x 10'7 neutrons/cm?.sec for 63 hr. This exposure corresponds to 25 times the fission density expected in the shielding experi- ment with enriched uranium, The capsules were opened in a special gos manifold, and the gas pressures were determined. The pressure inside the irradiated capsules was shown to have been greater than 50 and less than 100 psia. The pres- sure inthe unirradiated capsule was not perceptibly different from the vapor pressures of the two components (about 4 psi). The gas in the capsules was analyzed by infrared absorption spectrometry at K-25 and was found to contain CF,, C,F,, C,F ¢ and other fluorocarbon compounds not identified. No UF, could be detected in the gas sample. Pressure measurements at ~80 and —190°C were consistent with these analyses. The ir- radiated capsules were full of o nonvolatile (at 25°C) greenish powder. The powder was insoluble in C,F 160 It will be analyzed for uranium. The PERIOD ENDING SEPTEMBER 10, 1954 unirradiated capsule contained no residue except for a faint chalkiness on its inner wall. Additionai capsules are being irradiated for 1/100 of the accumulated exposure, LITR HORIZONTAL-BEAM-HOLE FL.UORIDE-FUEL LOOP W. E. Brundage C. Ellis C. D. Baumann M. T. Morgan F. M. Blacksher A. S. Olson R. M. Carroll W. W. Parkinson J. R. Duckworth 0. Sisman Solid State Division A loop for circulating fluoride fuel was inserted in hole HB-2 of the LITR and filled with fluoride fuel NaF-ZrF -UF, (62.5-12,5-25 mole %) during a protracted reactor shutdown. The pump was started as soon as sufficient fuel had been added, but circulation of the fluoride mixture became erratic after 5 to 10 min and finally stopped completely. Fluoride fumes in the off-gas line from the jacket around the pump indicated a leak,® and therefore the loop was removed before the reactor was started. The loop was carefully disassembled to aveid loss of uranium, since over 5 kg of enriched fuel mixture had been used to charge the loop. Recovery of the fuel from the ‘‘nosepiece’’ at the in-pile end of the loop had to be carried out in a hot cell because of neutron activation acquired when the unfilled loop was exposed to reactor flux before the fuel was added. The leak was found in-the weld joining the loop tubing to the discharge nipple ot the bottom of the pump bowl. Complete breaking of the weld occurred probably during disassembly of the loop or during cutting of the pump jacket, rather than while the pump was in operation, Before insertion in the reactor the jacket for enclosing the loop and the pump in an inert atmos- phere was tested for gas tightness., The loop, with the pump superstructure replaced by a blank flange, was vocuum tested at 800 to 1000°F with a helium leak detector. After the loop was inserted in the reactor, it was pressure tested at 1200 to 1500°F ot 16 psig, and there was no perceptible loss in pressure over a l-hr period, even though the pump superstructure with its rotating shaft seal was in place. 5W. E. Brundage et al,, ANP Quar, Prog. Rep. June 16, 1954, ORNL-1729, p 107. 141 ANP QUARTERLY PROGRESS REPORT The connections of the loop tubing at the pump have been redesigned for the second loop to provide a filling and drain line. This will permit charging the loop with a non-uranium-bearing flucride mixture, circulating it at operating temperature, and draining it before the loop is inserted into the reactor. The danger of tube rupture becouse of expansion of the fuel on melting necessitates the draining and refilling feature, if operating tests are to be carried out. The necessary changes in the pump shield to accommodate the drain and fill line are being made. The Metallurgy Division will supervise all critical welding and has provided specifications for an improved weld joint developed for Incone! tubing. The components fabricated previously for the second loop are being changed to conform to the specifications for the new weld joints. A new flange for closing the rear of the jacket has been made in order to give more space for making pipe connections to the large air lines required for the heat exchanger. During installation of the loop in the reactor an unexpectedly long time was required to connect heater leads, thermocouple leads, and air, water, and helium piping in the confined space available between the shielding and the instrument pane!, The piping at the reacter face is being revised to facilitate connection when the second loop is ready for insertion so that installation may be performed more ropidly. Modifications are also being made to remove lag in the flow alarms for the jacket cooling water, The oil systemfor cooling the pumps is being rebuilt to incorporate more reliable pumps ond improved flow and alarm instrue ments. A hydraulically operated door to match the with- drawal shield has been installed in o wall of a hot cell in preparation for handling the loop after A horizontal band saw has been modi- fied for almost complete remote operation and is being installed in the hot cell. Other remote han- dling tools for disassembling the loop have been made and are being tested. irradiation. ORNL GRAPHITE REACTOR SODIUM-INCOKNEL LLOOP W. E. Brundage W. W. Parkinson Solid State Division Metallographic examination was completed on an Inconel loop which had circulated sodium in hole 142 58-H of the ORNL Graphite Reactor for 218 hr with the irradiated section at 1500°F. The unperiurbed thermal flux in this hole is about 7 x 1011, and the flux above 1 ev is a factor of 10 lower. There was no evidence of radiation induced corrosion of the Inconel. The temperature of the pump cell section was held between 1025 and 1100°F, and that of the line entering the reactor was somewhat higher; the temperature of the line emerging from the reactor was about 1500°F. Before insertion in the reactar the loop had been operated at 800 to 1000°F for 80 hr while sedium was circulated through the loop and through a bypass filter circuit connected to the loop. CREEP AHMD STRESS-CORROSION TESTS W. W. Davis J. C. Wilson N. E. Hinkle J. C. Zukas Solid State Division Development of a suitable apparatus for in-pile stress-corrosion experiments hos been slow be- cause of difficulty in securing constant hegt- transfer rotes from the fuel to a heat sink where the fission heat can be removed. In the design shown in Fig. 9.7, fission heat is conducted from the fuel through the stressed specimen wall to a sodium-filled annulus. The sodium acts as an efficient heat conductor that exerts no constraint upon the deformation of the specimen tube as it creeps under applied stress. Part of the fission heat is removed by gaseous convection at the outer surface of the sodium-containing chamber and part by conduction through o metal ring to a water jacket. Simpler designs in which the sodium was contained in o constant-diameter annuius did not permit good temperature control. The temperature gradient that resulted from simulated fission heat input at the bottom and heat abstraction at the top caused heated sodium to rise discontinuously in the annulus, and temperature excursions of 100°F in 1 sec were not uncommon. A substantial temperature gradient (500 to 800°F per in.) is required to meet headroom requirements in the experimental hole and to prevent sodium distil- lation out of the annulus. Several designs of the sodium annulus were tried, such as baffles and spirals, ond ratios of outside-to-inside sodium-annulus diameters were varied but did not result in a constant temperature gradient, The apporatus shown in Fig, 9.7 appears to operate satisfactorily, but more operating time PERIOD ENDING SEPTEMBER 10, 1954 UNCLASSIFIED ORNL - LR-DWG 3109 FURNACE - L1 " \¢ G - INCONEL SPECIMEN — 3 ¥4 P l,u'_! BAFFLE ¥ ill X LIRS 22 | TR " “q i FIXED BASE~ \ FUEL CHAMBER .~ COCLANT CHAMBER == THERMOCOUPLE WELL Fig. 9.7. Stress-Corrosion Apparatus, must be accumulated before it will be certain that sodium distillation is not taking place in sufficient quantities to cause trouble. In the apparatus a tubular Inconel specimen con- taining 0.2 g of NaF-Zer-UFd {53.5-40.0-6.5 mole %) in the fuel chamber attached to the fixed base is eccenirically loaded through a lever arm by a weight. A stress pattern is produced across the horizontal cross section which varies from a maximum compressive stress. Surrounding the specimen is a sodium coolant chamber with baffies which reduce the magnitude of the thermal-con- vection currents and allow closer control of the specimen temperature. A furnace helps to maintain the temperature at 1500°F throughout the test. Two thermocouple wells project into the sodium chomber to a position neor the walls of the speci- men at the neutral axis. A double-walled, con- centric, sodium fill line, with only the outer tube welded to the chamber, and a vacuum exhaust tube complete the apparatus. All surfaces in confact with the fuel and sodium are made of Inconel. The assembly is hydrogen fired for 1 he at 1750°F, filled with fuel, and welded in a dry box wnder purified helium. The other parts of the container are welded into subassemblies, hy- drogen fired, and then welded together, with helium backing up the welds. The lever arm and weight are adjusted and spot welded onto the specimen, and the chomber covers are welded into place. Betore the rig is filled with sodium, it is filled with NoK and heated to 1000°F to remove any 143 ANP QUARTERLY PROGRESS REPORT The rig is then washed with butyl alcohol, which reacts with the NaK ond leaves the surfaces clean ond ready to be filled with sodium. The sodium is metered out inte a tube that has first been cleaned with NaK, sealed with Swagelok fittings, and provided with a helium valve. A micrometallic filter sealed into the line removes oxide remaining on the metal surfaces. any oxides of sodium that may be formed when the connection to the test rig is made. Once the con- nection is made, the apparatus is evacuated and the sodium is melted and forced into the chamber by helium pressure through the valve; a combi- nation of high surface tension and back pressure made filling the chaember somewhat more difficult without this procedure. For the bench tests, in sticks of sodium were used, the filled chamber was heated while a vacuum was pulled on the system, and a rise in pressure was noted which at about 525°F, presumably from dissociation of. sodium hydride. It remains to be seen whether such a procedure is necessary to ensure the purity of the sodium now available. The dry-box work is completed by cutting the outer fill tube, re- moving the inner (contaminated) tube, and crimping and welding both the fill and exhaust tubes. The entire assembly is welded into a stainless steel water jacket furnished with Kovar seals and a capillary tube through which helium flows throughout the test. A probe beneath the weight cuts off the current to the furnace at a prede- termined deflection of the lever arm and indicates the time required for a given degree of deformation. A gravity ‘‘tilt"" indicator aids in plumbing the specimen in the exposure can. A base plate with a conical cup furnishes a well for the sodium in case of rupture during operation, A suitable safety system is being developed. Approval by the LITR Experiment Review Com- mittee is being withheld until additional tests are made on the compatibility of sodium with the various component materials it might contact if a rupture occurred at the test temperaiure. Several tests aimed at a qualitative determination of the speed of reaction and the heat generated by such reactions should be completed within a short time. After an irradiation period of from two to six weeks in hole HB-3 of the LITR and a suitable decay period, the specimen is to be sectioned by the remote metallography group and examined for stress-carrosion effects. 144 Bench tests on the specimen tube alone are under way with the outside surface in air, The inside of the tubes contains either air, helium, or NaF-ZrF ,-UF, (50-46-4 mole %). An atmosphere chamber has been built for testing four specimens simultaneously in vacuum or any desired atmos- phere. One 200-hr test has been completed on barren NcF-ZrF4-UF4 (50-46-4 mole %) at 1500°F and about 1500 psi. No stress dependence of corrosion was found; another specimen was tested for 400 hr but has not yet been examined. A melting and overturning technique for removing fluoride fuel from the section that is to be ex- amined metallographically has been successful; this will greatly simplify handling of irradiated capsules. The MTR tensile creep rig is undergoing final testing and is scheduled for irradiation tests on Inconel in Qctober. REMOTE METALLOGRAPHY M. J. Feldman W. Parsley R. N. Ramsey A. E. Richt Solid State Division Work was continued on the examination of the solid-type fuel sandwiches of interest to the Pratt & Whitney Aircraft Division. In addition to the two experiments mentioned previously,® six other specimens have been exaomined. The results on three of them (capsules 1-4, 1-6, 1-7) have been published.” A report covering the other three capsules (1-5, 1-8, 1-9) is being prepared. Ex- amination of the eight capsules processed to date has indicated greater resistance to cracking and core-clad separation upon bending of the stainless- steel-matrix samples than upon bending of the iron-matrix samples, better resistance to cracking upon bending of the large UQ, particle size core than upon bending of the smaller particle size core, and a greafer increase in hardness upon irradiation of the small particle size core than upon irradiation of the larger particle size core. Work on solid-type fuel elements for the GE-ANP group has continued. Examinations of two ni- 6A. E. Richt, E. Schwartz, and M. J. Feldman, Solid State Div. Semiann. Prog. Rep., Feb. 28, 1954, ORNL- 1677, p 9. 7M. J. Feldman et al., Metallographic Analysis of Pratt and Whitney Capsules 14, 1-6, and -7, ORNL CF-54.5-41 (May 5, 1954). chrome V fuel elements have been completed.? Examinafions of two single-plate fuel elements® ond one two-stage multiple fuel plate test as- sembly!? were also completed. Work will continue on both wire and multiple-plate elements. Expansion of remote metallographic facilities is proceeding. ' FISSION-FRAGMENT ANNEALING STUDIES M. J. Feldmon W. Porsley Solid State Division Because of interest in the possibility of the removal of the radiation damage to fuel plates by annealing, a preliminary study was wndertaken. The eight samples in Pratt & Whitney capsule 1-9 (four were small particle size, <3-u UG, and tour were large particle size, 15- to 44 UO.Q) were selected for the study. Cne sample from each group (large and small particle size) was examined after the following treatment: (1) as irradiated, (2) 900°F anneal for 24 hr, (3) 1100°F anneal for 24 hr, and (4) 1400°F anneal for 24 hr, showing the results of this preliminary study is shown in Fig. 9.8, The conclusion drawn fron: this initial study is that a portion of the neutron A graph damage to the samples has been annealed out. Since the full annealing temperature for type 347 stainless steel is about 1800°F (for complete removal of the effects of cold work), a complete anneal for neutron damage at 1400°F was not expected. For the core, where the major portion of the damage is by fission fragments, it is felt that the curve shows a reduction in hardness be- cause of a partial anneal of the neutron damage with no reduction in the fragment damage. An attempt will be made to determine the anneal nec- essary to remove all the neutron damage from this type of material and, if possible, the fission- fragment annealing temperatures. HIGH-TEMPERATURE, SHORT-TIME, GRAIN- GROWTH CHARACTERISTICS OF INCONEL M. J. Feldman A. E. Richt Solid State Division W. Parsley As an aid in the analysis of the static corrosion capsules, a study of the short-time, high-tempera- ture, grain-growth characteristics of the Inconel 8M. J. Feldman et al, Metallographical Analysis of ?.E%bsfll;;’re Fuel Element No. 1, ORNL CF-54-4-8 {April PERIOD ENDING SEPTEMBER 10, 1954 stock used for the tests was made. Information is availoble in the literature on the usual times and temperatures for Inconel grain growth, but because of the nature of the experiment, with its possibilities of local short-time hot spots in the fuel, this study was initiated. Figure 9.9 is a graph of the data obtained. Of major interest were the definite temperature dependence of the carbide solubility and the extremely short times (relative to the corrosion test times) at which large grains could be produced. Details of the experiment have been published.!! BENL NEUTRON SPECTRUM - RADIATION DAMAGE 5TUDY J. B. Trice P. M. Uthe Solid State Division R. Bolt J. C. Carroll N. Shiells California Research Corporation V. Walsh Brookhaven National Laboratory A series of measurements were made in hole E-25 of the Brookhaven reactor to determine neu- tron flux energy distributions, The project was a cooperative effort with California Research Cor- poratien and Brookhaven National Laboratory. In- formation obtained from these measurements is to be used to correlate neutron flux intensity and energy distribution with radiation damage to cer- tain lubricants and organic compounds that have been irradiated in E-25 and examined by CRC. Data for the irradiated samples have been coded for the ORACLE by the ORNL Mathematics Panel so that the lorge number of arithmetical compu- totions usuolly required by o threshold detector experiment can be performed in o shorter period of time than is usual. ?A. E. Richt and R. N. Ramsey, Metallographical Analysis of Single Plate Fuel Elements GE-ANP 3B and 3C, ORNL (CF-54-3-42 (March 9, 1954), XOM. Jo Feldman et al.,, Metallographic Analysis of TwosStage MTR Test Specimen GE-ANP-1B, ORNL CF+54-7-77 (July 9, 1954), ”M. J. Feldman et al., Short Time—High-Temperature Grain Growth Characteristic of Inconel, ORNL CF-54- 6-70 (June 8, 1954). 145 ANP QUARTERLY PROGRESS REPORT $36° DIAMOND) £ESS (DPH HARDN 600 550 500 450 400 350 300 250 200 150 100 UNCLASSIFIED CORNL-LR-DWG 268TA | SAMPLE 17, SAMPLE 19, 15 TO 44 UO, 3 UO, .fu44fizy AT HTOOF ANMNEALED FOR 24 hr | Oy ~ 1 1 o Jkiifi__ B R < | & | < | a | el i x | oo x| eS8 L E | W W, 0 ql gk | zl o | %I ul < | Z | ] I I \\\\\ \\\\\\\\\\‘Lil\\\\\\\\\ //I: A rllfi/ ; Yy 1/1[/// AT 1400°F bbbk P T T T 71T IRRADIATED 633 hr TO~12.49, U%3d BURNUP et e e e— e ey e e — e E— — Fig. 9.8. Preliminary Annealing Study of Stainless Steel-U0, Fuel Plates in Pratt and Whitney Capsule 1.9, 146 UNCLASSIFIED ORNL—LR—-DWG 1552A & TiME {min) [es] - e " e " Py - - AN s LN AL AN o5 LA 1800 1900 2000 2100 2200 2300 2400 TEMPERATURE (°F) Fig. 9.9. Grain Growth Characteristics of In- conel. ' PERIOD ENDING SEPTEMBER 10, 1954 MTR NEUTRON-FLUX SPECTRA - RADIATION DAMAGE STUDY T. L. Trent J. B. Trice P. M. Uthe Solid State Division J. Moteff General Electric Company Fission probes with different fissionable ma- terials were used to maoke neutron flux traverses in MB-3 of the MTR. The fissionable materials used included U238, Y236 Np237 gnd Th2%2, whose fission threshold energies span the energy region between Y% and 2 Mev. The data from these traverses will be used with the results from acti- vations in HB-3 of other, nonfissionable, threshold detectors that are in the process of being irradi- ated. After measurements have been completed in HB-3 with the hole empty, several inches of beryllium will be introduced into the hole in such a way as to change the energy distribution of the trans- mitted neutrons. The new energy distribution will then be measured, and materials will be irradiated in both moderated and unmoderated neutron spectra. An attempt will then be made to correlate spectra and radiation damage.'? 12.1. B, Trice and P. M. Uthe, Solid State Semiann. Prog. Rep. Feb, 28, 1954, ORNL-1677, p. 14. 147 ANP QUARTERLY PROGRESS REPORT 10. ANALYTICAL STUDIES OF REACTOR MATERIALS C. D, Susano Analytical Chemistry Division J. M. Warde Metallurgy Division The primary analytical problem continues to be the separation and determingtion of trivelent ond tetravalent uranium in both NaZers- and NaF-Kf- LiF-base fuels. A successful potentiometric titra- tion of UF, in molten NaZrF . with metallic zir- conium was accomplished by meons of polarized platinum electrodes. The observed end point occurred ot a weight of titrant which corresponded to one-half an equivalent of zirconium metal per mole of UF,, This stoichiometry indicates reduc- tion of UF, to a mixed oxidation state. Polaro- grophic studies of uranium fluorides in reactor fuels dissolved in molten ammonium formate ot 125°C were initiated. Preliminary experiments revealed that tetravalent uranium was not reduced at the dropping-mercury electrode, formote appeors to be o useful solvent for fluoride- base fuels containing various oxidation states of uranium. The conversicn of UF, to UCIH in fluoride fuels was shown to be quantitative when the fuel was heated with boron trichloride at 400°C for 90 min. Trivalent uranium fluoride is converted to UCl; under these conditions. The solubility of UF, in NaAICI, was found to be 18 mg/g at 200°C as contrasted to a solubility of UF, of less than 1 mg/g. Molten NaAlCl, is expected to dissolve tetravalent uranium selectively from the fuels. Calibration measurements have been completed Molten ammonium on the apparatus for the determination of oxygen as metallic oxides in reactor fuels, The reaction in- the hydrofluorination of the oxide ond measurement of the increase in conductivity of liquid HF as a function of the water formed. Investigations were also made on the oxidation valves of UF, with oxygen at elevated temperatures. At 300°C, UF; is converted quantitatively to UF, and UO,. The mechanism of the oxidation of UF with oxygen is being studied. An oxide residue of about 11% of the original material was found when UF, was heated in oxygen at 300°C for 2 hr, An improved method for the determination of lithium in NaF-KF-LiF-base fuels woas developead. The chloride salt is exiracted with 2-ethylhexanol, and the chloride is titrated in the nonogueous mediyums, 148 Studies were also made on the solubilities of potassium, rubidium, and cesium tetraphenylborates in various organic solvents to ascertain differential solubilities, The feasibility of a spectrophoto- metric titration of fluoride in NaF-KF(RbF)-LiF. base materials by decolorization of zirconium or aluminum complexes was investigated, Determina- tion of the concentration of oil in helium was made by absorbing the hydrocarbons in petroleum ether and weighing the residue obtained after evaporation. It was found that prior bleeding of the line effec- tively eliminated contamination by oil. ANALYTICAL CHEMISTRY OF REACTOR MATERIALS 1. C, White Analytical Chemistry Division Reseorch and development were continued on the problem of determining UF,; and UF, in NaZrFs- and NaF-KF-liF-base reactor fuels. A number of new opproaches to this problem were investigated, including redox titrations of the fluoride mixtures in the molten state, for which polarized platinum electrodes were used, and pelarographic studies of fluoride mixtures in molten ammonium formate, Work was continued on the determination of oxygen in fluoride fuels and on the determination of alkali metals and fluoride ion in NaF-KF-LiF-base mao- teriols, Determination of Oxygen in Fluoride Fuels A, S, Meyer, Jr, J. M, Peele Analytical Chemistry Division The apparatus for the determination of oxygen as oxides in fluoride fuels, which was described briefly in a previous report, 'has been constructed. A schematic drawing of this apparatus is shown in Fig. 10.1, The hydrogen fluoride purification still, which was fabricated from mild steel, is equipped with a 24 x 1 in, packed column with on estimated efficiency of 25 theoretical plotes. Conductivity ]A. 3. Meyer, Jr, and J. M. Peele, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 110. THERMOMETER WELL -——— 3| | &= “‘LQ»—V—COPPER ColL N (\ PRESSURE GAGE ' > PERIOD ENDING SEPTEMBER 10, 1954 UNCLASSIFIED ORNL-LR~DW3 2724 MONEL NEEDLE VALVES () MONEL BELLOWS VALVES . Tfi\\ ___________________ e - < 8 \ :@::_ COPPER TURNINGS ——n __ %93 s :::{(. il Y J\fo W' \\ , | 3 | — rm."L__ THERMOMETER WELL ~ 5 REACTOR ANHYDROUS HYDROGEN FLUORIDE TO VACUUM PUMP s - L e 20 % KOH SODA~LIME TRAP Fig. 10.1. Apparatus for Determination of Oxygen in Fluoride Salts. measurements are carried cut ina 25-mi fluorothene cell fitted with platinized platinum electrodes, The reaction vessel, a modified Parr bomb, is silver plated on the inner surfaces, as are the lines and the fittings in the reactor-conductivity cell section, : In the procedure now being tested the somple is placed in the reaction vessel and the system is evacuated to valve 5 (see Fig. 10.1}. After the cell has been filled with a measured volume of the distillate, the conductivity is measured and the acid is transferred to the reaction vessel by distillation. When the oxides have been converted to water, the solution of water in hydrofluoric acid is distilled to the cell where the measurement of conductivity is made. Transfers of the acid are repeated until constant readings are obtained. The distillation apporatus hgs been tested and has been found to deliver anhydrous acid at a rate sufficient to fill the cell in less than 1 min. When the still was charged with 99% HF, the conduc- tivity of the distillate corresponded to a maximum concentration of water of approximately 1 ppm.. Weighed samples of anhydrous LiOH are being taken through the procedure outlined above as a Calibration measurements have been completed over the range 0.03 to 0.25% H,0 in HF, means of calibrating the opparatus, 149 ANP QUARTERLY PROGRESS REPORT Oxidation-Reduction Titrations in Fused Salts A, S. Meyer, Jr. Analytical Chemistry Division Experiments were carried out to determine whether the equivalence point of a titration in which UF dissolved in a molten fluoride fuel solvent is re- duced by the addition of small increments of an electrodeposition metal could be observed by one of the conventional electrometric techniques, The first system studied involved the reduction of UF, in NaZrF . by zirconium metal according to the proposed reaction AUF, + Zr° > 4UF, + ZrF, . Since no insulating material which is compatible with molten NaZrF. was available, it was not possible to devise o reference elecirode for meas- urement of the potentials of the solutions, Accord- ingly, polarized platinum electrodes were selected as indicating clectrodes. In order to obtain more information about the nature of the polarization phenomena, the area of one of the electrodes was made approximately 80 times that of the other. Thus, since the current density at the smaller electrode waos 80 times as great as thot ot the larger, the potential difference between the elec- was approximately equal to that of the smaller electrode. By reversing the direction of the polarizing current, it was possible to determine the potential at the cathode and at the anode, The solution of UF, in NaZrF g was contained in a 25-ml platinum crucible which was placed in the helium-filled quartz liner of o pot furnace, Agi- tation was accomplished by means of a mechani- cally driven platinum stirrer which was constructed of 0.05in.-dia platinum wire, Additions o the solution were made through o quartz tube which extended from the top of the liner to a point about 1 in. above the melt. In pure NaZrF. the onode was more strongly polarized than the cothode. At polarizing currents of 80 pa the potential of both electrodes exceeded trodes 1 v. Approximately ]/2 hr was required for the electrodes to reach equilibrium. Platinum elec- trodes could not be oppreciably polarized in NaF- KF-LiF-base fuels, When UF , wos added to the molten NaZrF, the potential decreased rapidly. For a solution which . L. Colichman, Polaro§rap!3y in Molten Ammonium Formate, LRILL.-117 (April 1954) 150 contained 1.25 mole % UF the potential of both electrodes was approximately 300 mv, with no significant difference between the potentials of the cathode and ancde. During the early part of the titration of the molten fluoride mixture with zirconium metal, the potential of the cathode decreased to values less thon 10 mv and remained constant throughout the titration. At the same time the potential of the anode increased rapidly ond attained potentials in excess of 1 v before the end point of the titration was reached, When an equivalence point was reached, the poten- tial of the anode decreased rapidly to about 50 mv and then decreased gradually on further addition of Only one breck in the potential vs The curve is shown reductant, titrant curve was observed. Fig. 10.2, N “w ORNL—LR—DWG 2725 TP . —— - Y —mseey s ' ; | ‘ -2 | 2 -~ < | r’/ i o 08 r T e i - LeJ S : ™ ¢ s | < 0.6 ‘ II C r - ’ — / I | O / i | 0 ‘I | o) ; &j i o N | = ® SO L O o - S . i 1 A AT a- 0 01 0.2 03 04 s 06 c9 10 EQUIVALENTS OF ZIRCONIUM PER MOLE OF UF, Fig. 10.2. Titratien of UF, in Molten NaZrF, with Zirconium Metal, The cbserved end points of two titrations occurred at a weight of titrant which corresponded to one- half an equivalent of zirconium metal per mole of UF,. This stoichiometry indicates the reduction of UF, to a mixed oxidation state that corresponds to the postulated formula U, F,. Polarographic Studies in Fused Ammonium Formate A. S, Meyer, Jr, D. L. Manning Analytical Chemistry Division Colichman? reported that when solutions of ucl, in molten ammeonium formate were reduced at the dropping-mercury electrode, three well-defined re- duction waves with halfewave potentials of 0.1 to ~0.2, =0.75 to —-0.80, and ~0.95 v (vs pool) were obtained, The following reactions were postulated to explain these waves: UM-EP—’# U+3£~> ) +2 and : : Ut e > U+3———%1 U+:Z %Ew} ue . He also reported that anhydrous hexavalent uranium gave no reduction wave until after an aging period during which the compounds were reduced by the ammonium formate to a lower valence state, probably guadrivalent, | Since it appeared that this techmque could be appiied to the determination of tetravalent uranium in the presence of large quantities of trivalent uranium, expérimmfs were carried out to obtain polarograms of solutions of UF , experiments it was found that UF, ond NaF-KF- L:F base fuels dassoived readily i molten ammonium formate ot 125°C to form green solutions containing as much as 1 mg of uranium per milliliter. Although UF, alone was found to be much less soluble, UF; which had been fused with NaZrF . was readily soluble, Similor relationships were found when molten ammoniom acetate was used as a solvent, : No U(lV) reduction waves were obtained when polarograms of solutions of UF, were recorded immediately after dissolution, and no reduction waves were observed in the polarograms of a sample of NaF-KF-LiF-base fuel that contained both UF, and UF,. When these solutions were gerated, a single reduction wave with a halk wave potential of approximately ~0.2 v was ob- served. The potential and the diffusion current of this wave corresponded closely to those recorded immediately after the dissolution of an equivalent quantity of K;UO,F, in molten formate. Since the U(IV) waves moy have been shifted to potentials outside the range of the solvent (0 to ~0.9 v) by the complexing action of the fluoride ions, oddi- tional polarograms were token of solutions con- taining UCi,. Except for a small reduction wave ot about —~0.2'v, no evidence of reduction was found, The diffusion current of this wave was increased by aeration to values corresponding to those ob- tained for hexavalent uranium, and therefore the increase was atiributed to the oxidation of UC!M It appears that in molten ammonium formate no oxidation state of uranium other thon hexavalent is reduced ot the dropping-mercury electrode. In preliminary and NaZrF - PERIOD ENDING SEPTEMBER 10, 1954 Conversion of UF,; and UF to the Respective Chiorides with BCI3 A, S, Meyer, Jr. D. L. Manning Anquhcal Chemlstry Division It was reported previously? that, on the bas:s of free-energy ' calevlations, UF,, UF, ZrF,, ‘and NaF are quantitatively converted to the chlorides when at equilibrium with boron trichloride. Since the exchange reaction in organic solvents at moder- ate femperdtures was found to be too slow for analytical application, tests were carried out with gaseous BCl, ot elevated temperatures. This tech- nique offered an addifional advantage in that the resulting chlorides couid be sepumfed by subllmu- tion. When o somple of NaZrF, UF: fuel was heated in a stream of BCl, and hehum a‘t 350°C, the bulk of the zirconium was sublimed as ZrCl, within 90 min, and all but fraces were removed ofier heating for an additional hour at 450°C, It has been re- ported? that uranium can be sublimed as Uucl, by treating UF:‘i with BCl, af 600 to 650°C, but in these tests only o small fraction of the UCl, was volatilized - when the UC! -NaCl residue, which fused at about 450°C, was heated to temperatures as high as 750°C. ‘ Analyses of the residues showed that when the conversion was carried out at 400°C for 90 min the vranium in - the NoZrF UF samples was quanti- tatively converted to UCl, ond that all the uranium remained in the residue. When the conversion was carried out at 500°C, 2% of the uranium ‘was volatilized, while at 750°C, 4% was volatilized. When UF, samples were converted to UCI, by this method, significant froctions of the uranium were oxidized to UC|,, It is believed that this oxidation will be eliminated when all traces of oxidizing contaminants are removed from the gase- ous reactants, & Salubiiiéi‘y of Tri- and Tetravalent Uranium Flyorides in Fused NaAIC], A, 5. Meyer, Jr. W, J. Ross Analytical Chemistry Division in the continuation of experiments to defermine the optimum conditions for the conversion of UF, 3A. S, Maver, Jr., D. L. Manning, and W. J. Ross, ANP Quar, Prog, Rep, fune 10, 1954, ORNL-1729, p 110, 4y, P. Caolkin 5, Divect Conwversion of TFfi to TCIA, CEW.TEC C.0.350.4 (Sept. 23, 1945). 151 ANP QUARTERLY PROGRESS REPORT and UF, to the corresponding chlorides,® it was noted that UF , was dissolved much more readily thon UF; by fused NoAICl,. Measurements of the solubilities were carried out by equilibrating sam- ples of the uranium fluorides with molten NaAICH, for periods up to 5 hr, under an inert atmosphere, followed by the separation of the undissolved solids by filtration ond calculation of the solubility from the concentration of uranium in the filtrate, Solubilities of 18 and 16 mg of UF , per gram of NaAICl, were found at 200 and 175°C, respectively. However, the solubility of UF3 decreased from > mg/g of NaAICl, at 225°C fo less than 1 mg/g at 200°C. Since UCI; was also found to be sparingly soluble at the lower temperatures, the above values are indicative of the true solubility of trivalent uro- nium in this medium rather than of incomplete con- version of the fluoride to chloride. At temperatures lower than 200°C, molten NaAlCi‘i is expected to dissolve tetravalent uranium selectively from UF, which has been complexed by potassium fluoride in the fuel. Since all available UF; samples are contaminated with tetravalent uranium, it has not been possible to carry out accurate solubility measurements at the lower temperatures by relying on the determina- tion of total dissolved uranium. An apparatus has been constructed for the determination of microgram quantities of frivalent uronium in the filtrates by hydrogen evelution, This apparatus, which is based on the measurement of evolved gases at reduced pressures, is now being calibrated against cou- lometrically generated hydrogen. In order to carry out dissolutions at lower temperatures, LiCI-AlCI fluxes which melt at approximately 120°C have been prepared. Oxidation of UF; with Oxygen A, S, Meyer, Jr. W. J. Ross Analytical Chemistry Division Preliminary investigations have been initiated to determine the extent of the oxidation of UF3 by oxygen at elevated temperatures. The reaction is assumed to be In the tests conducted to date the reaction products are refluxed with 2.5 w/v % ammonium oxalate solution, ond the soluble portion is analyzed for SW. J. Rodd and J. L. Mattern, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 111, 152 its uronium content, which is assumed to be UF ,. In these tests the reaction time was held constant at 2 hr and the temperafure was varied from 200 to 600°C, with the results shown in the following: Temperoture Reactivity (°C) (%) 200 30 300 100 200 91 600 25 At 300°C the reoction proceeds quantitatively as written, At higher temperatures, however, the con- centration of soluble uranium (UF4 in ammonium oxalate) decreased rapidly from the theoretical content with increasing temperature, probably as o consequence of a further reaction of oxygen with UF . The experiment was then repeated with UF, and oxygen. At 200°C the entire sample was soluble in ammonium oxalate, as expected, which indicated no oxide formation. At higher temperatures, a black insoluble residue was formed upon dissolution of The amount of residue increased at higher temperatures from 11% at 300°C to 27% at 600°C., The major portion of the sample dissolved rapidly, and a yellow filtrate was formed. It is to be noted that Kraus® reported that the following reaction occurs ot 800°C: UF, + 0,—> UF, + UO,F, . the somple in oxalate solution, The yellow color of the filtrate substantiates this reaction in part, but there is no way of accounting for the black residue, obviously an oxide of ura- nium. 1t is known,” however, that UF, is converted quantitatively to U,0g when ignited ot 900°C in air, Petrographic and x-ray siudies of the material will be made. Further tests are planned at various reaction times and temperatures. Determination of Lithium, Potassium, Rubidium, and Fluoride lon in NaF-KF{RbF)-LiF-Base Fuels J. C. White G. Goldberg B, L. MeDowell Analytical Chemistry Division A critical review of the analytical methods for the determination of lithium, potassium, rubidium, 6C. A, Kraus, Technical Report for the Months of April, May and June, 1944: Chemical Research — General, CC-1717, July 29, 1944, 7). J. Kotz and E. Rabinowitch, The Chemistry of Uranium, p 374, McGraw-Hill, New York, 1951. and fluoride ion is being made to ascertain those applicable to the ondlysis of NaF-KF(RbF)-LiF- base reactor fuels. Lithium. An indirect volumetric method for the determination of lithium in the presence of the large concentrations of sodium and potassium and/or rubidium which are found in NaF-KF(RbF)-L.iF-base fuels was developed. The method was designed specifically to provide a rapid, and yet precise, means of determining lithium. The alkali metals are first separated from uranium and then converted to chlorides. Lithium is extracted as the soluble chloride salt by heating ot about 135°C with 2-ethylhexanol® until the slightly soluble salts of sodium and potassium become *‘free-flowing”’ and no longer cling to the walls of the vessel, These salts are removed by filtration, and then the tiltrate is diluted with ethanol. The chloride in the filtrate is titrated by the Volhard method,? and the lithium concentration is calculated, The coefficient of varigtion is 0.5% for 1 to 50 mg of lithium. A topical report of this work is being written, Potassium and Rubidium. Since the concentration of either potassium or rubidium in NaF-KF(RbF}- lLiF-base fuels is of the order of 30 wt %, the de- termination of potassium or rubidium by means of the flome photometer has been neither sufficiently accurate nor precise, The conventional method for this determination is the perchlorate separation !0 of potassiumor rubidium perchlorate from the sodium and lithium salts in ethyl acetate. The methed is satisfactory but requires considerable operator time. |herefore the application of possible entirely substitute procedures which are specific for potas- sium and/or rubidium was investigated. cedure!! involving the precipitation of potassium bitartrate in ethanol-water and titration of the solution of the salt with standard base was tested. The method is extremely simple and rapid, and the bitartrate salt is readily filterable. Rubidium behaves similarly to potassium, but cesium does A pro- not precipitate as the bitartrate salt under these conditions. Further work is necessary to determine the effect of acid and salt concenirations on the 8E. R, Caley and H. D. Axilrod, Ind, Eng. Chem. Anal. Ed, 14, 242 (1942). 1. M. Kolthoff and E. B. Sandell, Testbook of Quanti- tlatéve Inorganic Analysis, p 477, MacMillan, New York, Y010:d,, pp 41415, YA, F. levinsh and Ya. K. Ozol, J. Anal. Chem. USSR 8, 57 (1953). PERIOD ENDING SEPTEMBER 10, 1954 solubility of potassium bitartrate before a decision can bereached as to the applicability of this method to NaF-KF-LiF-base fuels. A second method under consideration is the use of sodium ftetraphenyl boron'? as a reagent for the direct determination of potassium, rubidium, and cesium. The potassium salt is precipitated in aqueous acidic solution, filtered, dried at 110°C, and weighed as potassium tetraphenylborate. The method has the advantage of being simple and rapid, and it has a favorable weight factor for potassium, Themain interference in the use of tetraphenylboron is the ammonium ion, The reagent is expensive, but it is readily available at the present time, Preliminary work on the solubilities of potassium, rubidium, and cesium tetraphenylborates in non- aqueous solvents has shown some interesting dif- ferences in solubilities. The possibility of ex- ploiting differential solubility as a means of separating the salts is being explored, Acetyl- acetone, diethylketone, and methy! isopropylketone are three possible solvents for this separation, The order of solubility in organic solvents has been established as: KTPB > RbTPB > CsTPB (TPB = tetraphenylboron). Fluoride lon, The determinatian of fluoride ion in fluoride fuels was previously conducted by the pyrohydrolysis'® procedure which was particularly successtul for NaZrF -base fuels. Alkali metal fluorides, however, are resistant to pyrohydrolysis, and thus considerable time, of the order of 3 to 4 hr, is required to pyrohydrolyze completely 100 mg of NoF-KF-LiF-base material. Two alternative methods have been studied, therefore, for opplication to the determination of fluoride ion in NaF-KF-LiF-base materials. A pro- cedure, which was formulated by Chilton, ' was tested. In this procedure the fluoride ion is titrated with a standard solution of aluminum (potassium alum), and the resulting change in acidity is re- corded. Considerable difficulties have been en- countered, the foremost of which is the extremely small change in pH at the equivalence point. The method in its present form does not appear to be applicable to this work. The feasibility of a spectrophotometric titration of fluoride ion based on the decolorization of o 12, Geilmann and W. Gebaukr, Z. anal. Chem. 139, 161 (1953). 13), C. Warf, W. D. Cline, and R. D. Tevebaugh, Anal. Chem. 26, 342 (1954). 145, A Chilton, personul communication. 153 zirconium or aluminum complex, or lake, is also being investigated, In this method, the solution of fluoride is titrated with o colored zirconium com- plex, and the change in absorbance of the solution is plotted as a function of the concentration of in the complex. Zirconium alizarin sulfonate and zirconium eriochrome cyanine both react rapidly with fluoride ion and will receive zirconium further study. Preliminary results indicate that the reactions of these compounds with the fluoride ion are reproducible, A simple titration cell is being designed. Qil Contamination in ARE Helium G. Goldberg Anolytical Chemistry Division The contamination of the helium for the ARE with oil from valves and fittings was checked before the gas at the site was used, The method used for these tests was to pass the gas through dual scrubbing towers containing petroleum ether and then to evaporate the ether to dryness and weigh the residue. In the initial test on a tank car of helium, 360 pg/ft3 was found. A tower containing 70 in.3 of charcoal was placed in the gas line, but it had little effect in eliminating the oil. Further tests on the same tank car showed that bleeding the line before samples were taken markedly lowered For example, in three successive samples of 15.5 fi3 of gas each, the contamination decreased from 100 to 10 ug/ft3. These tests indicate that bieeding the line with 50 to 100 ft2 of gas before the equipment is coupled the contamingtion level, into the system will effectively remove oil from fittings and valves, PETROGRAPHIC INVESTIGATIONS OF FLUORIDE FUEL G, D. White, Metallurgy Division T. N. McVay, Consultant Petrographic studies of UF , in alkali-metal fluo- ride systems have shown that two phases exist in the KF-UF,; system. One is red and is neor 3KF-UF ;, and the other is blue and is near KF.UF . Only one binary phase is present in the NaF-UF, system, and it is probably 3NaF.2UF,, Lithium fluoride has not been found to combine with UF, in either the binary system LiF-UF,, in the ternary 154 systems LiF-KF-UF, or LiF-NaF-UF,, or in the quarternary system LiF-NaF-KF-UF, Cesium fluoride was found to form a complex with UF,, with a composition near 3CsF-UF,, Some work was done on the systems LiF-ZnF, and NaF-CsF, and no binary compounds were found, Part of a loop in which NaF-ZrF -UF, (50-46-4 mole %) had been circulated was received from Battelle Memorial Institute, Samples token from the wall contained about 50% ZrO,, and the balance was an NaF-ZrF, (53-47 mole %) compound. In addition, several loops containing NaF and ZrF, with reduced UF, have been examined, and no appreciable oxidation of the UF; could be found. SUMMARY OF SERVICE AMALYSES J. C. White W, F. Yaughan C. R. Williams Anglytical Chemistry Division Most of the somples received during the past quarter were from NaZrF .- and NaF-KF-LiF-base matericls. Other types of samples that were analyzed included silver fluoride, chromium fluo- rides, iron fluorides, beryllium, sodium, and rubidium metals, Inconel, and alkali-metal caor- bonates., A revised procedure was established fo determine UF; and UF, in NaF-KF-LiF-base fuels, Also, the method of White, Ross, and Rowan !5 was adapted successfully to the determination of oxygen and metallic impurities in rubidium metal. A total of 1,165 samples, involving 10,702 de- terminations, was analyzed., A breckdown of the work is given in Table 10.1. 15), C. White, W. J. Rass, and R, Rowan, Jr., Anal Chem. 26, 210 (1954). TABLE 10.1, SUMMARY OF SERVICE ANALYSES REPORTED Number of Number of Samples Determinations Reactor Chemistry 679 6,694 Fuel Production 38 395 E xperimental Engineering 379 3,406 Miscellaneous 59 207 1,165 10,702 Part Il SHIELDING RESEARCH 11, SHIEL.DING ANALYSIS E. P. Blizard J. E. Faulkner H. E. Hungerford F. H. Murray C. D. Zerby Physics Division H. E. Stern Consolidated Yultee Aircraft Corporation Theoretical analyses have been made so that an understanding of the air and ground scattering measurements: at the Tower Shielding Facility could be obtained and the attenuation in the side wall of an aircraft crew shield could be determined. In connection with the cir and ground scattering analy- ses, special attention was given to the calculation of the reduction of air scattering due to ground interruption and to the effects of multiple scattering. The gamma-ray slant penetration has been calcu- lated for the side wall of an aircraft crew shield, and the variation of radiation intensity throughout the crew volume has been briefly explored. SLANT PENETRATION OZF COMPOSITE SLAB SHIELDS BY GAMMA RAYS C. D. Zerby The total radiation dose in the crew compartment of an airplane is partially dependent on the gammo- ray flux penetrating the compartment shield. To obtain o fundamental, but practical, knowledge of the gamma-ray penetrations, the problem has been treated theoretically by using stochastie, or Monte Carlo methods. The problem was programed for solution on the ORACLE ond was arranged for investigating the effects of the variations of all the parameters in- volved. The shield was taken as o composite slab of two materials comprising a thick layer of Compton scattering material ' followed by a thin layer of fead. In each of many cases the initial incident photons were taken as monoenergetic and incident on the slab at a particular angle with the normal'to the slab. The stochastic process used the exact physical analog of the probability laws known to govern the life of a photon. ]In the range of energy considered, the Compiocn scatterer has opproximately the same physical chorac- teristics relative to a photon passing through it as does concrete or polyethylene (CH,). The parameters and their variations that have been investigated to date are listed below: Thickness of Compton SCOQ’ferEr,z in. 3,9, 15 Thickness of lead, in. ]/]0, :}/10’ }'2 Initial photon energy,3 me? units 2,6 Initial angle of incidence, deg 0, 30, 60 All combinations of these parameters were investi- gated. & The following solution of a typical problem indi- cates part of the information obtained: Initial Conditions Thickness of Compton scatterer 3in. Thickness of lead ]/]0 in, Initial photon energy 6 mc2 Initial angle of incidence 60 deg Results Fraction of initial energy 1. reflected 0.0234 2. obsorbed in the Compton scatterer owing 0.318 to scattering collisions 3. absorbed in lead owing to scattering 0.0568 collisions 4, obsorbed in lead owing to absorption 0.0515 collisions 5. penetrating shield 0.550 6. penetrating shield without being 0.435 degraded in energy Energy build-up factor 1.265 The energy spectrum for the reflected photons was also obtained in this typical problem. The spectra for the photons penetrating the shield were obtained in the angle intervals 0 to 15, 15 to 30, 30 to 45, 45 to 60, and 40 to 90 deg. These spectra will be available in a forthcoming report. 2 . The electron density of the Compton scatterer was token as the some as that for polyethylene, 3Mu|tiply mcz units by 0,5108 to obtain Mev. 157 ANP QUARTERLY PROGRESS REPORT AIR SCATTERING OF NEUTRONS In an effort to understand the air-scattered neutron intensities which have been measured af the Tower Shielding Facility, a number of calculations have been carried out with a variety of conditions and assumptions. The predominant difference between the cuirent work and that which was done pre- viousiy? is that the interference of the ground has The ground scattering is In an attempt to obtain a been taken info account. calculated separately. better fit to the data, such aspects as onisotropy of scattering and slant penetration at the detector tank (crew shield) have been taken into account in some of the work. Furthermore, measurements have been made of the angular distribution of the radiation leaving the reactor shield, and these data have been used in some of the calculations. Because the importance of the multiply scottered neutrons has appeared to be greater than was antici- pated, considerable effort has been mads to extend the calculations to beyond the singly scattered case. Some results ore reported for two scatterings, and o general method has been developed for all orders of scattering. Single Anisofropic Air Scattering in the Presence of the Ground (Shielded Detector) F. H. Murray For a cenvenient machine calculation, the at- tenuation in air was neglected, The detector was placed with 10 cm of water between it and the face of the crew shield which, when extended, contained the reactor source. The beam from the source was about an oxis through the detector, in theory, and contained terms cos a and cos® @ a calculation was made separately for the terms 1, cos o, and cos? ain ithe expansion of the source flux. I D is the reactor-detector distance and b the height of both above the ground, the flux at the detector is expressed as the sum of the terms: g w /2 and 12 _ o i h/~W 16a2p B=0 ¢=7/2 158 where Il gla + B) 1 +0.3cos(a+ ), 2 Aa) = a, 1; b, cos a; ¢, cos” a . The angle ¢ is the angle between the vertical plane taken as the face of the crew compartment and the plane friangle formed by the source, the scattering volume, and the detector. The results of this calculation® will be published later. Single !sotropic Air Scattering in the Presence of the Ground (Unshielded Detector) J. E. Favulkner A colculation was made of the reading of an iso- tropic detector caused by first-scattered neutrons in air as the distance above the ground was varied. The following assumptions were made: 1. The source and detector have the same alti- tude, 2. The ground is an infinite plone. 3. The source is isoiropic. 4. The scattering in air is isotropic. 5. There is no energy degradation on air scatter- ing. 6. The attenuation is pure inverse square. With these assumptions the reading on the detector may be shown to be proportional to »/(5) b \p) where D is the separation of source and detector, b is the altitude, and / may be expressed in closed e~1:25/sin Bsin ¢ gg g4 f e 1-25/sin Bsin ¢ df de¢ fflomfif(a) gla + B) do 4See, for example, H. Goldstein, Chap. 2.9, p 831, in Reactor Handbook, ed by J, F. Hogerton and R. C. Grass, Technica! lnformation Service, AEC, 1953, F. He Murray, Single Anisotropic Air Scattering of Neutrons in the Presence of the Ground (Shielded Detector), ORNL. CF.54-8-104 {to be issved). arc cot {D/h cos ¢ —~ cot ) f(a) gla + B) da, a=0 for0§<;c‘5§n/2; forn/2 . v v mm B § Y (@,) Y _(0,) doy L Y s Fob = [ e - 1Py} . p c be a + ik cos(k,r) The source function may be anisotropic; however, Let if an isotropic source is present or if the scottering , ' law possesses axial symmetry, some simplificafions = f ffe"zk"o C°s('0'k)f c(Q) C{VQ ’ occur. : e ? When the scaftering density has been obfmned by and then a Fourier inversion, one further infegration gives Lr o, Y& =Ere LA 1, v(). Equating coefficients of id m+l E qu pc Cy P where entical Y's gives e = L Pac P quc = Z Aqr: [pc * the flux inte an arbitrary unit volume from: any direction; if the detector sensitivity depends on direction, the total scattering to be counted at any point P “is represented by a convolution and can be calculated by another Fourier transformation and inversion. . The method is to be applied te air scattering measurements taken at the TSF and will make possible subsequent calculations for unusual reactor shield shapes, v Thus, if lec» is represented as a vector (leo,F?fifl’,,Fm’z, ...) for all m, Fm+1,0 Fm,O m+1,1 o, 1 N s, 5wl - al-lrl Fm+l,q ‘m,c Then, formally, | | en, tormd y Method for Isotropic Source and Scattering. |f (Pl = B 3;:0; the scattering is isotropic after each collision, the formula to be used for calculating the total flux at N [a ¥ : WLt = BEF I = B2 Fo the point P after the nth coliision is a convolution: F § = B™{F,} FPY = o [ f [F, {0 v, 0P 161 ANP QUARTERLY PROGRESS REPORT with WQRPQ SOP) =, 2 471R PO Since the Fourier tronsform of a convolution is the product of the transforms of the individual func- tions, for an isotropic source with no energy loss ofter each collision the transform TF becomes (Tqfi)”” O_Sn with 1 A+ po T =——1| et - b gt ()= s where ]/2 p =1 Cr)§+é)§+m§] . In taking the inverse of TF_,the path of integration is deformed inte the closed curve about the negative rea! axis to the left of —a. Calculations of these scattering functions for the first- and second- scattered flux give F | = 1.635, F, = 0,19167 . 2 These figures are to be multiplied by 1/4772 AD, for A = 130 meters, to obtain the number of fast neutrons at a distance D feet from the source. GROUND SCATTERING OF NEUTRONS A. Simon H. E. Stern An investigation is being carried out to derive formulas for the ground-scattered neutron flux to be obtained from a general source distribution. The model used is that of a source whose axis, or direction of maximum intensity, forms an arbitrary angle 0 with the source-receiver axis but lies in the same horizontal plane as the latter. The differential source strength is assumed to be a function of only the angle between the emergent ray and the source axis. The receiver is isotropic under the assumptions of no air attenuation and a constant albedo for the ground. The following expressions for the flux at the receiver are ob- tained: 162 For isotropic re-emission from the ground, N OA F . ffl (e} sin a da a=0 X f P 8722 7w/ 2 dep mi(m]x+m2y+w3z) e Ax,y,2) dx dy dz For cosine re-emission from the ground, NoA T _ F = —— fla) sin a da 4?72192 a=0 /.2 d X f ___qs__, D= mtt/ 2 [g(a,q'))]?’/z where N, = total source strength {(neutrons/sec), A = reflection coefficient of the ground (albedo), fla) = relative source strength per steradian at angle a between emergent ray and source axis, b = height of source and receiver above the ground, ] gla,g) - —— sin? a cos? b D -2 o cos ( eot a sec ¢ 2 D 9 / 2y - — | sin G tan & + | — b T\ ) = angle between source axis and source- receiver axis, D = separation distance between source and receiver. PERIOD ENDING SEPTEMBER 10, 1954 Alternatively, if a single-scatter approach with attenuation by a *‘removal’’ mechanism is used, the fol- fowing formula is obtained: F o= f fla) sinadaf - . 167252 Zr a=0 b —1/2 gla,d) 1 + sina cos & gla,d) where % = scattering cross section, Y = removal cross section, ORNL-LR-DWG 2547 A detailed report on this calculation is being prepared. 8 FOCUSING OF RADIATION IN A CYLINDRICAL CREW COMPARTMENT J. E. Faulkner A study of the focusing of radiation in o cylindri- cal crew compartment has been made, with particular attention being given to the case with the following conditions: | 1. The cylinder is infinite. 2. The along the inner eylinder walls. 3. The distribution of the radiation depends only on the angle between the direction of emergence and the normal to the cylinder. 4. The attenuation inside the crew space is pure surface radiation density is constant angulor inverse square. 5. The detector is isotropic. Under these conditions the angular distribution may be represented as an isotropic component plus a series in odd powers of the cosine. The contri- bution of each term at a given point in the com- ponent in the series may be expressed in an even polynomial in x (see Fig. 11.3), where x is the distance of the given point from the axis-of the cylinder divided by the radius of the cylinder. For pure cosine distribution the reading of an isotropic detector is independent of position in the crew compartment, 8 A. Simon and H. E. Stern, Some Calculational Methods for Air) and Ground Scattering, ORNL CF-54-8-103 (to be issued). T CROSS SECTION OF CREW COMPARTMENT x=h/R i cos 8 L ol bl L \ P = RADIATION PER cm® OF INSIDE SURFACE PER STERADIAN IN THE NORMAL DiREGTION O.B e L ,...!.,UA AAAAAAAA deinaienn ...____L ....... 0.6 [ofeoeereerfressseeferen G 2 E': “'“L*--fi_ \\\ — = T~ \ Z 0.4 B T~ z TN N 2 \< cos 8 \ \\ 0.2 ‘ -------- T~ o L L4 b Fig. 11.3. Variation of Radiation Intensity with Distance from Center of Crew Compartment. 163 ANP QUARTERLY PROGRESS REPORT 12. LID TANK SHIELDING FACILITY G. T. Chapman J. M. Miller W, Steyert D. K. Trubey Physics Division J. B, Dee Pratt ond Whitney Aircraft Division Preliminary work for further circulating-fuel reflector-moderated reactor shield experiments at the Lid Tank Shielding Facility (LTSF) has con- tinved with in the ORNL Graphite Reactor of a somple of the UF -C, 7 ¢ mixture irradiation that may be used to simulate the reactor fuel in the shield mockup. Also, a new effective removal cross section for carbon has been obtained for the case in which the carbon is distributed uniformly throughout the shield. Thermal- and fost-neutron measurements have been made around an array of three of the GE-ANP helical air ducts, and a 35-duct array is being Difficulties in fabricating the enriched uranium plate for the new LTSF source have delayed the completion of this project; however, it appears at present that these difficulties have been surmounted. It is anticipated that the installation will he completed within a month. assembled for further measurements. REFLECTOR-MODERATED REACTOR AND SHIELD MOCKUP TESTS J. B. Dee D. K. Trubey W. Steyert A second series of mockup tests for the circu- fating-fuel reflector-moderated reactor (RMR) and shield is being initiated at the LTSF.! For this series o larger tank (approximately a 6-ft cube) has been constructed to hold all the dry components of the configurations, as well as an exponsible plastic bag for containing boroted water. The beryllium blocks to be used in the mockups will also be placed inplastic containers for additional protection. The dry tank face adjacent to the source plate has a ]/B-in.-’rhick Incone! window, which corre- sponds to the RMR core shell. In order to determine the effect of the Inconel on the gamma-ray dose, TFor first series see C, L. Storrs et al., ANP Quar, Prog. Rep. Sept. 10, 1953, ORNL-1609, p 128. 164 gamma measurements were taken in pure water in the tank. The dose was higher by a foctor of 2 than that normally observed in the LTSF, and the increase agrees closely with the calculated 9-Mev capture gomma-ray dose from the Inconel. I addition to the presently planned static fission source tests for RMR designs, a dynamic fission source test is being considered for measuring the sodium activation from delayed neutrons released in the heat exchanger and the attenuation of gamma rays from shori-lived fission products. A liquid being considered is C;F, containing 20 wi % UF . [n cooperation with the Radiation Damage group of the Solid State Division, a sample containing natural irradiated in a fluorinated nicke! capsule in the ORNL Graphite Reactor for an inte- grated flux of 10'7 avt, which compares with an integrated flux of 10'3 nvt for the enriched fuel. After the volatile products had been removed by vacuum distillation, a precipitate containing most uranium was of the radicactivity and a large part of the uranium was found in the irradiated capsules; thus, this fuel mixture could not be used for long exposures. A similar capsule test is being prepared for a shorter exposure, and alternate solutions are being explored. EFFECTIVE REMOVAL CROSS SECTION OF CARBON D. K. Trubey Megsurements of the removal cross section of carbon have been made in a continuous carbon medium obtained by dissolving sugar (C,,H,,0,,) The solution (density = 1.312 + 0.001 g/cm3) contained 64.2 wt % sugar, which gave 0.354 g/ecm® of carbon and o hydrogen and oxygen in water, density that was 96% of that of plain water. It was contained in a large tank that had a I/B-in.-thick Inconel window on the source side. The thermal-neutron flux, the fast-neutron dose, and the gomma-ray dose are shown as functions of distance from the source in Figs. 12.1, 12.2, and Va CigHaply a0 ‘ | ‘ . /{"“_ J.::f;;; T CTTTIIT TV L CHRHOT THERMAL-NEUTRON FLUX {nv,, ) 0 26 40 60 80 100 120 140 60 Z, DISTANCE FROM SOURCE (em) Fig. 12.1. tion. Thermal-Neutron Flux in Sugar Solu- 12.3, respectively. Plain water curves are also shown for comparison. The gamma-roy water curve is higher than the normal LTSF water curve because of the high-energy capture gamma ray from the Inconel window of the tank (see preceding discus- sion). Since the medium contained almost as much hy- drogen and oxygen os does plain water and con- tained them inthe same ratio, no geometric correction was made in calculating the effective removal cross section, o . The average o_ for the range of 90 to 140 cm from the source was 0.750 barn. This com- pares with a value of 0.81 £ 0.05 barn froman LTSF measurement behind a slab of graphite that con- tained 51.3 g/cm?® of carbon, which corresponds to PERIOD ENDING SEPTEMBER 10, 1954 ORNL~LR-DWG 2649 & 2 - ~ 1. ¥ E 0 1™ Q 0 = 0 - B = 1 — 192 & N o 2 - — b fil I CyoHpz 0y M0 \\ 5l et \ e 7”......““”,.‘ ............................................. ‘,i.,vr...fi‘_‘}.. —_— 2 e e e il —_— — 13 ! 70 80 20 100 110 120 130 Zz, DISTANCE FROM SOURCE {(cm) Fig. 12.2. Fast-Neutron Dose in Sugar Solution. the sugar-water solution at 145 ¢m, fn order to observe the neutron attenuation in a medium removed from the source, the sugar-water solution is being placed in a 36-in.-long aluminum tank located 48.2 cm from the source. Thermal- neutron, fast-neutron, and gamma-ray measurements will be made in the tank. The measurements at the interface will indicate the difference in neutron age between water and the sugar solution for the more penetrating of the fission neutrons. 165 ANP QUARTERLY PROGRESS REPORT GE.ANP HELICAL helical air ducts at the LTSF. The ducts (Fig. AIR DUCT EXPERIMENTATION 12.4) were fabricated from flexible steel conduit J. M. Miller shaped around a 9-in. core. After removal of the core, the ducts were stiffened by Fiberglas wrapping. Thermal-neutron measurements have been made around 3-in.-dia by 4-ft (developed length) GE-ANP ORNL—LR-DWG 2651 — 10 4 ORNL—LR—DWG 2650 A 0 - —_ - e e e e JP i GAMMA-RAY DOSE {mr/hr} THERMAL-NEUTRON FLUX (nvih) 60 70 80 90 100 10 7, DISTANCE FROM SOURCE (cm) Fig. 12.3. Gamma-Ray Dose in Sugar Solution, YNCZLASSIFIED GT - GNP PHGTO U-32438 120 130 140 150 160 z, DISTANCE FROM SOURCE {cm) G e e Fig, 12.5. Thermal-Neutron Flux Measurements Beyond One and Three GE-ANP Helical Air Ducts — Fig. 12.4, GE-ANP Helical Air Duct. Horizontal Traverses. 166 PERIOD ENDING SEPTEMBER 10, 1954 Two configurations, o single duct and a friangular neutron traverses parcllel to the source plate. An array of three ducts (5 in. from center to center), array of 35 ducts is to be placed in a medium of were placed in the LTSF with one end adjacent to steel Raschig rings (35 vol %) and borated water the source plate. A comparison of the thermal- for a study of the effect of the addition of ducts to neutron flux beyond the two configurations is given gamma shielding, in Fig. 12.5, and Figs. 12.6 and 12.7 are thermal- g S ORNL--LR--DWG 2653 ORNL-LR—DWG 2852 ' B T I JnllJln{ ,,,,, T[ T I b ] ,,,,,,,,,,,,,, S TN SR ST PR SR e i i e S I O T N T ‘= o:c £ ( E £ .| HOCURVE AT £ £ = x 3 2 i@ L z - g E [ DA = R0 em T NGIN > wl . - = z H ‘e ‘ e e 4 3 e e N g 5 ‘ _ N & 5 5 e — 0 g N e L:"-:| u ~H,0 CURVE AT z =130 om | = = 9 /A 1 T,,_f_f [ER —_1 N e ] : _ ™ {/' —H,0 ATl z =140 cm \‘ i - ._..._._,‘WL_..-.._._..__ _...__.[....“L..n..______.“............_.t.._--___‘. -70 ~50 -30 -0 10 30 50 70 -70 =30 -3 -0 t0 30 50 70 ¥, VERTICAL DISTANCE FROM SOURCE AXIS (cm) ¥, VERTICAL DISTANCE FROM SOURCE AXIS (cm) Fig. 12.7. Thermal-Neutron Flux Measurements Fig. 12.6. Thermal-Neutron Flux Measurements Beyond Three GE-ANP Helical Air Ducts ~ Vertical Beyond One GE-ANP Helical Air Duct — Vertical Traverses. Traverses. 167 ANP QUARTERLY PROGRESS REPORT 13, BULK SHIELDING FACILITY R. G. Cochran F. C. Maienschein G. M. Estabrook K. M. Henry J. D. Flynn E. B, Johnson M. P. Haydon T. A Love R. W. Peelle Physics Division At the Bulk Shielding Facility (B35F) measure- ments were made of reactor radiations through thick slabs of graphite, In this experiment the fast-neutron spectrum through graphite and a re- moval cross section of carbon were also determined, The light to be given off from a nuclear-powered airplane has been further investigated, and some quantitative measurements are reported. In ad- dition, o method of determining the power of the ARE by means of fuel activation technigues is described. REACTOR RADIATIONS THROUGH SLABS OF GRAPHITE R. G. Cochran J. D, Flynn G. M. Estabrook K. M. Henry Measurements been completed for de- termining the attenuation of large thicknesses of graphite next fo a reactor, ! have These measurements are of interest for evaluating a graphite retlector as a shield component, and they also provide a direct comparison with LTSF determinations of the carbon removal cross section. Graphite thicknesses of 1, 2, and 3 ft were used, and the usual gamma-ray, thermal-neutron, and fast-nsutron dosc measure- ments were made behind each slab thickness., In addition, the fast-neutron spectrum (above 1.3 Mev) through 1 ft of graphite was measured, A graphite slab 1 ft thick and one 2 ft thick were constructed, and the 3-ft thickness was obtained by strapping thess two slabs together. When strapped together, there was no possibility of water getting between the slabs. Fach slab was constructed fargely of graphite blocks, 4 in. x 12 in, x 5 ft, stacked there could be no streaming of radiation through cracks extending through on entire slab thickness. in aluminum tanks in such @ way that Vihe details of this experiment will be reported in a memorandum by R. G. Cochran et al., Reactor Radi- ations Through Slabs of Graphite, ORNL CF-54-7-105 (to be issuzd). 168 Thin graphite shims were added to fill the tanks, which were sealed by heliarc welding and were pressure tested for leaks., The total aluminum wall thickness for each tank was 1.3 cm, Fulk Shielding Reactor loading No, 26 (Fig. 13.1) was used. This configuration is very similar to loading 22, which has been studied in considerable detail for other experiments,? but loading 26 uses less fuel to compensate for the presence of the graphite, which is a better reflector than the water it replaces. Since the presence of the graphite perturbed the neutron flux in the reactor core rather severely, the neutron flux distributions and thus the reactor power were determined in the usual way by means of cobalt foils, 2R, G. Cochran et al., Reactivity Measurements with the Bulk Shielding Reactor, ORNL-1682 (to be issued). OSNL-—LR 7056 2654 CREACTOR GR.O FLATE STANDARD FUEL ASSENSLY ,SAFETY RODS e S N o N M ELEMENT WAaTER REFLECTOR' ¥, CLEMENT : CONTROL ROD fig. 13.1. Reactor. L.oading 26 of the Bulk Shielding The thermal-neutron flux {(Fig. 13.2) was measured behind each of the three graphite slab thicknesses to a distance of 300 cm from the reactor face. Measurements were made with a 3-in. U?3> fission chamber, an 8-in. BF; counter, and a ]23’2-in. BF, counter, and the data were normalized to indium foil data. Fast-neutron measurements (Fig. 13.3) were made with o three-section neutron dosimeter, Gomma-ray measurements (Fig. 13.4) were taken ORNL-LR-DWG 265 S 3-in, FISBION CHAMBER, " B-in. BF, COUNTER, AND = 2% -in. By COUNTER; DATA NORMALIZED B T F — T THERMAIL-NEUTRON FLUX {n,, /watt) o g, W 1 21 OF ........ G AC’HITE” “,,u,.,,,,,...,.*,,,, ..... T TO INDILM FOIL DATA A ft OF FRAPH\TF’ TUNARON T T 400 DISTANCE FROM REAGTOR {cm) Fig. 13.2. Thermal-Neutron Flux Measurements Behind AGHT Graphite. PERIOD ENDING SEPTEMBER 10, 1954 behind each of the three graphite slab thicknesses, both to determine the magnitude and relaxation lengths of the gamma-ray dose through graphite and to ensure that none of the neutron detectors was being used in a too high gamma-ray field, which would have caused their readings to be high. The fast-neutron spectrum from 1.3 to 10 Mev was measured with the BSF fast-neutron spectrometer,® A spectrum with reasonable statistics could be obtained only through the 1-ft graphite slab; the neutron intensities through the 2- and 3-ft sections were too loew. The spectrum measured with the 3R G. Cochran and K. Henry, A Proton Recoil Type Fast-Neutron Spectrometer, ORNL-1479 (April 2, 1953). FAST-REUTRON DOSE {mrep/hr/watt} GRAPHITE SLABQ_", 7 I8 ALUMINUM TANKS A0 &0 30 GO DISTANCE FROM a0 20 REACTOR lam) 0 20 Fig. 13.3. Fuast-Neutron Dose Measurements Be- hind AGHT Graphite. 169 ANP QUARTERLY PROGRESS REPORT ORNL-LR-DWG 2657 10 .. 1t OF GRAPHITE . . BN S \_\( - 2 H OF GRAPHITE \ ST = 1 N - 3 ft OF GRAPHITE | et N ] < : : ,_,’% - o ! L e , | L A A | ; R \ a . . . P & .. .. .- i > Ve & ) /\‘\[/ efi I . . . . & 5 ‘ N 2 : 50-cc ION CHAMBER ; \i\ i ", \ h \ \‘(-\‘ | - - -fSLAE Ry - 2-fislAg | | SR v BOTH SLABS ’ | S . l . l ; i ‘ ‘ ‘ GRAPHITE SLABS | INALUMINUM TANKS LO-4 I l . ] - . - . oo 0 40 80 120 160 200 240 DISTANCE FROM REACTOR {cm) Fig. 13.4. Gamma-Roy Dose Measurements Be- hind AGHT Graphite. 1-ft graphite thickness between the reactor face and the end of the spectrometer collimator is shown in Fig. 13.5. The spectrum measured with the end of the collimator against the reactor face is also shown. Removal cross sections have been calculated for each thicknass of graphite by use of a method suggested by E. P. Blizard.* The resulting cross sections are 0,82 barn/atom for the 1-ft slab, 0.84 barn/atom for the 2-ft slab, and 0.80 barn/atom for the 3-ft slab, These removal cross sections are in good agreement with measurements made on graphite at the L. TSF, REACTOR AIR GLOW R. G, Cochran T. A. Love K. M. Henry F. C. Maienschein R. W, Peelle Attempts to theoretically determine the amount of visible light which may surround a nuclear-powered airplane in flight have resulted in widely differing 170 values,®s® Therefore an experiment was performed at the BSF to provide an experimental basis for future estimates. The end of an air-filled aluminum periscope tube was placed at the reactor face, and the amount of light produced in the tube was measured by a photomultiplier with spectral response to that of the average human eye. Other relative measurements were taken with a photomultiplier which was sensitive chiefly in the blue and near- ultraviolet range., The latter measurements are plotted in Fig. 13.6 as functions of the air pressure in the tube, Measurements with argon in the tube demonstrate that neither the approximate amount of light produced nor the exact spectrum emitted is strongly dependent on the atomic number of the gas. It is also interesting to note that the light production in a given volume of air appears to have a maximum at a pressure corresponding to an altitude of about 30,000 fi. It is demonstrated in Figs. 13.7 and 13.8 that the air glow is largely caused by gomma radiation rather than by neutrons. Figure 13.7 shows the decay of the light plotted along with the decay of the reactor gamma ion chamber current just after reactor shutdown. The attenuation by water of the radiation which produces the air glow is shown in Fig. 13.8. This attenuation rather closely follows that of gamma rays. similar The photomultiplier was used to compare the quantity of light given off in the air-glow tube with that from a small tungsten lamp mounted at the reactor end of the tube, This comparison showed that 7.2 x 107° lumen was given off by the glow for a reactor power of 100 kw and atfmospheric pressure, Presumably, the amount of light should be proportional to the integral of the gamma-ray dose rate over the volume of the air inthe measuring tube. This integral wos estimated to be 1.1 x 1010 (r-em)/hr. Therefore the effective light production per unit volume of air is L = 6.5 %1071 (lumen/cm3)/(r/hr) . If it is assumed thot all the light is given off at 4E, P. Blizard, Procedure for Obtaining Effective Removal Cross Sections from Lid Tank Data, ORNL CF-54-6-164 (June 22, 1954). 5T, A. Welton as quoted by C. E. Moore, Visual Detecrability of Aircrajt at Night, LAC-15, p 24 (Aug. 14, 1953). 6). E. Faulkner, Visible Light Produced in Air Around Reactors, ORNL CF-54-8-99 (tc be issved). PERIOD ENDING SEPTEMBER 10, 1954 ORNL-—@ 2658 FLUX (neutrons/cmZ/sec/watt/ Mev) 4-ft COLLIMATOR AGAINST i{-ft OF GRAPHITE 0 2 4 6 8 10 12 14 ENERGY (Mev) Fig. 13.5. Fast-Neutron Spectrum of the Bulk Shielding Reactor Observed Through 1§t of Graphite. 171 ANP QUARTERLY PROGRESS REPORT ORNL -LR-DWG 2859 EQUIVALENT ALTITUDE () 60,000 40,000 20,000 9 7 x10 T i T T | ; ‘ i \ \ —3 0O | E LIGHT CURRENT {amps) S A PRESSURE (cm Hg) Fig. 13.6. Variation of Glow with Pressure, 4600 /i, which is in the blue region, 7.2 x 10=3 lumen = 1,9 x 10~ 7 watt of visible light. The amount of air glow can then be characterized as the fraction of energy transformed into visible light: watts of visible light in tube watts of energy dissipoted in tube by gomma radiation 1.9 x 10~¢ T e = 6 x 107 for atmospheric pressure. 3.2 x 102 This experimental value of light yield is in disa- greement with the theoretical estimates of Welton® (0.05) and of Faulkner® (2 x 107%), In neither of these calculations was the effect of pressure considered, 172 The value of L may be used to calculate the light emission from a typical nuclear-powered aircraft, For a reactor power of 200 Mw and a reactor shield of 147 em of water (similar to the shield design of the 1950 ANP Shielding Soard?), the BSF measurements indicate that the gamma-ray dose just outside the edge of the shield would be 1.36 » 10% r/hr with a relaxation length of 22 cm. The dose integrated over space outside the shield is 1.2 x 10'5 (recm®)/hr. Multiplying this by L, the luminosity would be 75 lumens, which would have roughly the equivalent brightness to the eye of a 10-w (+7 w) incandescent light if the aircraft were far enough away to appear as a point source 7Report of the Shielding Board for the Aircraft Nuclear Propulsion Program, ANP-53 (Oct. 16, 1950). ORNL~LR--DWG 2860 —~8 107 L . ; Lo e ‘E ' ] I Iffifiifiifififi:[fifiiff’J'f'","fif",fffffifffififi.{fffifififfifi ............. o ] 5 | e Jo AIR GLOW — ] » e GAMMA RADIATION 1’? T B T ] 5 ; ] = ;': PRESSURE = 30cm Hg 2 - g \ = | T 5 ' r:_l:) o & {5 - M I = Qo 53 < o = = = = 19 w0 I 0 5 10 15 zZ0 25 30 35 TIME (min} Fig. 13.7. Decay of Air Glow and Reactor Gumma Rays. of light., This large uncertainty is primarily at- tributable to the low sensitivity of the eye-equiva- lent photomultiplier. A more detailed discussion of these preliminary velues will be given in a forthcoming report,® FUEL ACTIVATION METHOD FOR POWER DETERMINATION OF THE ARE E. B. Johnson When the ARE goes into operation, many measure- ments of its performance will be made. Among them will be the power level at which it operates, which will be measured in more than one way. A method suggested by J. L. Meem is based on the measurement of the relative activity of fuel samples exposed in the ARE and in a known flux in another reactor (BSR}). An experiment has been initiated at the BSF to implement this suggestion. 8F. C. Maienschein et al., Measurements of a Reactor- Induced Air Glow, ORNL CF-54-9-1 (to be issued). PERIOD ENDING SEPTEMBER 10, 1954 ORNL-LR—DWG 2661 1 oo AR GLOW . . § - ¢ (ATMOSPHERIS PRESSURE ) LIGHT CURRENT {amp) GANMMA~RAY DOSE [{r/hr) /watt] FAST NEUT ol (arbitrary units) N 1. o 10 20 20 DISTANGE FROM REACTCR {cm) Fig. 13.8. Attenuation in Water of Radiation Causing Air Glow, The activity resulting from the irradiation of fissionable material is not easy to predict because of the large number of isotopes for which calcu- lations would be required. However, for identical flux, exposure time, and waiting period after exposure, the specific activity of two samples should be the same. Furthermore, at low neutron fluxes the activity should be proportional to the flux. [t is contemplated that the ARE will be operated at a nominal power level of gpproximately 1 w for 1 hr, that it wiil then be shut down, and that a sample of the irradiated fuel will be withdrawn for counting and analyzing, The ARE experimental program will then proceed as scheduled. A decay curve will be run on this fuel sample from the ARE and compared with the decay curve obtained on o similor fuel capsule irradiated for the same length of time in the BSR in a known neutron flux of about the same magnitude. From the relative activity of the two samples, the operating power of the ARE 173 ANP QUARTERLY PROGRESS REPORT can be determined readily, For the power com- parison it will be necessary to know the uranium content of the sample and of the ARE, as wel! as the energy per fission.” A small correction will 95 L. Meem, L. B. Holland, and G. M. McCammon, Determination of the Power of the Bulk Shielding Reactor — Part [Il. Measurement of the Energy Released per Fission, ORNL-1537 (Feb. 15, 1954). 174 be necessary for the local depression of flux by the uranium sample. This factorhas been estimated by W, K. Ergen to be 0.80. Details for the appli- cation of this method are given in a separate repori’.]D 10¢, 3, Johnson, Fuel Activation Methods for Power Determination of the ARE, ORNL CF-54-7-11 (in press), PERIOD ENDING SEPTEMBER 10, 1954 14. TOWER SHIELDING FACILITY C. E. Clifford T. V. Blosser J. L. Hull L. B. Holland F. N. Watson Physics Division D. L. Gilliland, General Electric Company M. F. Valerino, National Advisory Committee for Aeronautics J. Yan Hoomissen, Boeing Airplane Company The Tower Shielding Facility (TSF) experimental program has, thus far, included measurements of ground- and air-scattered fast neutrons and the development of a new procedure for the determi- nation of the power of the reactor. Tests on the GE-ANP R-1 divided-shield mockup haove been started. FAST-NEUTRON GROUND AND AIR SCATTERING MEASUREMENTS T. V. Blosser J. Von Hoomissen D. L. Giililand F. N. Watson The performance of neutron and gamma-ray air scottering experiments that are free from an ex- cessive background of ground-scattered radiation is a primary objective of the TSF, Therefore, measurements of the scattered fast neutrons aos a function of reactor-detector altitude were neces- sary to determine the contribution of ground- scattered radiation to the total flux, particularly at the maximum altitude. These measurements will help to indicate the magnitude of the ground- scattered neutron background to be expected in future differential experiments, and they will aid in an understanding of the variation of ground and air scattering as the ground is approached. Measurements of the thermal-neutron distribution were token in the detector tank, which is es- sentially a 5-ft cube of water and which was, for these experiments, situated 64 ft from the reactor tank. The reactor was placed at an angle ¢ of 330 dey from the 4 axis (Fig. 14.1), and a BF, counter was moved along a line normal to and near the right side of the detector tank. In this region, contributions from other faces of the tank were negligible; thus the neutrons detected by fthe counter were the air- or ground-scattered fast neutrons which entered the side wall and were thermalized in the water near the detector. The reactor and detector tank altitudes were varied simultaneously, in discrete steps, from 0 to 195 ft. T R WA A ——— i A composite plot of the measurements {(Fig. 14.2) indicated only small differences in slope but ap- precicble differences in magnitude between the curves for the various altitudes. A plot of the flux vs the daltitude (Fig. 14.3) showed o .pro- nounced peck in the region. between the 15- and 20-ft altitudes, which indicated the importance of ground-scattered neutrons in this region. [t should be noted that the reading ot zero altitude was obtained with the reactor half-immersed in the ground pool, the upper half being shielded as before; so essentially half of the source was occluded, as was half of the scattering medium. Thus the air-scattered neutrons should be no more than half of the value at full altitude and no less than one quarter. Becnuse the intensities at the 150- and 195-ft altitudes were the same, it was concluded that the ground-scattered contribution at these altitudes was small, fn a preliminary analysis of the dota, it was estimated that this contribution was 2 to 5% of the total scattered neutrons. While the detailed caleulations of ground and air scattering being undertaken in connection with this investigation are not yet completed, o preliminary comparison of the dato with calculations carried out ot the Boeing Airplane Company for a not too dissimilar situation has been made. The qualitotive agree- ment is very good, although the pedk in the total flux, as measured, seems to be at o lower altitude. A report on this experiment is being prepared.! CALORIMETRIC REACTOR POWER DPETERMINATION D. L. Gilliland L. B. Holland 3L, Hull A procedure hos been developed and tested for a colorimetric determination of the power of the YeL B Clifford e al., A Preliminary Study of Fast- Neutren Ground and Air Scattering at the Tower Shielding Facility, ORNL CF.54-8.95 (1o bs issued). ANP QUARTERLY PROGRESS REPORT REACTOR TANK d — distance from the center of the reactor tank to the 0, 0, 0 point. b — altitude of both the 0, 0, 0 point and the center of the reactor. @ — horizontal angle between p and 4 axes. a — detector orientation angle, horizontal angle between 4 axis and a perpendicular to the broad side of the triplet BFF, chamber, UNCLASSIFIED CRNL-LR-DWG 843 DETECTOR TANK iLEFT SIDE DETECTOR RIGHT SIDE TOP VIEW OF DETECTOR TANK SHOWING DETECTOR ORIENTATION ANGILE @ 0, 0, 0 point location — on the outside of the de- tector tank 2.5 ft above the inside bottom of the tank. x, y, z — coordinates of geometric center of de- tector. Since the center of detection of the various counters will shift with the location of the detector in the tank, the x, y, z coordinates indicate the geometric center of the detector. p — thickness of water between outer reactor face and reactor tank wall, measured along a radius. Fig, 14.1. Geometrical Convention Adopted for Tower Shielding Facility Experiments. 176 RELATIVE THERMAL-NEUTRON FLUX (ARBITRARY SCALE) PERIQD ENDING SEPTEMBER 10, 1954 g 3 ORNL~-LR-DWG 2772 10 | I [ 1 i l | | _________ _ A = VARIABLE 8 = 330 deg CENTER OF TRIPLET BF, CHAMBER AT: __| - d = 64 ft = —90 deg x = VARIABLE y = 79.6cm S T p=452cm z=00cm T > CURVE ALTITUDE 1 15 ft ‘ , o 25 ft 3 8 fi o2 \\\ 3 4 4t SON 4 5 50 ft \\\\\\ e 6 96 f1 6 7 150 ft \i\\\\)K - 8 197 ft 3 \ \\\ 9 O ft N\ ‘\\\\\ N\ NN , N\ N \\ XA 1 10 X OO 3 87 A AN LT 5 N \\\\\>( - 9 N NN\ 5 N\ "“\\\\\\ \\ \\\\ \ \ 2 t 1 -75 ~65 -55 ~45 -35 X, HORIZONTAL DISTANCE FROM ¢ AXIS TQ DETECTOR CENTER (cm) Fig. 14.2. Attenuation of Scattered Neutron Flux in Detector Tank ot Various Altitudes. 177 ANP QUARTERLY PROGRESS REPORT SECRET s—0I-056-2- 72 O S T T |/ \ . . ' ‘ . | o % ) _ \ . ¥ =-67.5 cm , sl -8 S 5 ./ \ . LT v—m___L,,__,, _________________ % i w . > ' \ . x=-65cm [ R = e s & L T ; m / \o z Ll N : 2 h. - . . . > \__\. ¥=—60 cm % “—""*‘“"J—“'-—-uo----u—v " Lt }__ = _a = -390 deg . A _ =z = 54 ft o - - . . e & L g=sd _ A S 5 p =452 cm = . 6 = 330 deg ‘ _ ‘ , “ CENTER OF TRIPLET BFy CHAMBER AT: E x = SEE ABOVE tYy-in IRCN WALL AT x=-77.5cm) £ y=79.6 cm o . i . W z=02¢cm ‘ ' > < e o 01 : - e — 0 24 18 72 96 120 444 168 {92 215 &, ALTITUDE (i1} Variation of Scottered Neutren Flux Fig., 14.3. with Altitude, TSR, water are made while the water is being stirred suf- ficiently to ensure that no significant difference in temperature exists throughout the tank. The Thermocouple measurements of the bulk temperature in the 12-ff-dia reactor tank rate of temperature rise of the water when the reactor is started is then proportional to the power of the reactor. Small corrections must be made for the power input due to the mixing and for the heat loss. This procedure was first used for a 5 by 6 fuel element loading in the TSR. The reactor tank was lowered into the drained handling pool, and four mixers were placed at various positions within the tank. The turbulence created by the mixers was sufficient for thorough mixing of the water, but the power added by the mixers was not enough to give a measurable increase of the water temper- ature. Before the reactor was started, the temper- 178 ature remained constant for more than 1 hr with the mixers running. Soine heat insulation was obtained by creating, with a double layer of canvas, a dead-air space above the water and around the reactor tank. The temperature rise during a reactor run was measured by 12 thermocouples placed in the tank. The thermocouples had previously been calibrated at temperatures of 0, 25, and 50°C and agreed to within 0.1°C. The first run of the experiment was started at 24°C, and the total temperature rise was approxi- mately 14°C. higher temperature level but with the some heat input initial temperature for this run was dpproximately 15°C above the ambient temperature, and the rise was again 14°C. The cooling rate ot the conclusion of this run was 0.3°C/hr. In a third run the initial temperature was brought 15°C below the ambient by cooling the reactor tank with ice before the run. The results of the three runs were consistent to within 1%. A detailed report of this experiment is being prepared.? The experiment was repeated at a and total temperature rise. The GE-ANP R«1 DIVIDED-SHIELD MOCKUF TESTS T.V. Blosser J. Van Hoomissen D. L. Giliiland . N, Watson Tests on the GE-ANP R-1 divided-shield mockup (Figs. 14.4 and 14.5) will begin with traverses around the reactor-shield section in the handling pool for comparison with measurements made at the BSF.3 Measuraments will then be made in the TSI detector tank, which will be situated 64 from the reactor shield ot an altitude of 195 fi. [t is anticipated that the detector tank will be replaced by the crew-compartment mockup in mid- September. 2c. e, Clifford et ul., Calorimetric Power Determi- ngtion of the Tower Shielding Reactor, ORNL CF-54-8- 105 (to be issued). 3. E. Hungerford, Bulk Shielding Fuacility Tests on the GE-ANP R-1 Divided Shield Mockup, ORNL CF-54-8- G4 (to be issued). PERIOD ENDING SEPTEMBER 10, 1954 s o N O N T Sl e a ¥ . o oy 2L -ANP R-1 Divided-Shield Mockup. GE Fig. 14.4. Reactor-Shield Section of the 179 Fig. 14.5. Crew-Compartment Section (Ends Removed) of the GE-ANP R-1 Divided-Shield Mockup. 180 Part IV APPENDIX REPORT NO. CF-54-7-11 CF-54-7-143 CF-54-7-144 CF-54-8-171 CF-54-6-201 CF-54-7-1 CF.54-7-187 CF-54-8-200 CF-54-7-166 CF-54-8-199 CF-54-7-159 CF-54-8-180 CF-54-8-27 CF-54-4-188 CF-54-7-145 15. LIST OF REPORTS [SSUED DURING THE QUARTER TITLE OGF REPORT l. Aircroft Reactor Experiment Fuel Activation Method for Power Determination of the ARE ARE Operating Procedures, Part |, Pre-Nuclear Operation ARE Operoting Procedures, Critical Mass of the A Reactor Design P Part H, Nuclear Operation ARE Reactor 1, General Design arameter Study The Effect of the RMR Shield Weight of Varying the Neutron and Gamma Dose Components Token by the Crew and Comparison of the ized Shield Test Results and Design Comparisons for Liquid Metal-to-Air Radiators RMR Shield Weight to that for an ldeal- High Conductivity Fin Test Results I, Experimental Engineering Morse Silent Chain Drive Test on ARE Pump Hot Shakedown Test Stand Aircraft Reactor Experiment Fuel-Removal System Mockup IV, Critical Experiments The First Assembly of the Small Two-Region Reflector Moderated Reactor The Second Assembly of the Small Two-Region Reflector Moderated Reactor V. Metallurgy Preliminory Metallographic Examination of K-25 Heat Ex- changer Loop Vi. Heat Transfer ond Physica! Properties Physical Property Charts for Some Reacter Fuels, Coolants, and Miscellanecus Measurement of the Therma! Conductivity of Melten Fluoride Mixture No. 104 Materials (4th Edition) AUTHOR E. B. Jehnson W. B. Cottrell J. L. Meem C. B. Mills > . H. Fox et al. R. M. Spencer H. J. Stumpf H. J. Stumpf R. E. MacPhersen J. G. Gallagher J . E. Ahern A. G. Grindell J. Y. Estabrook Dunlap Scort Dunlap Scott R. S. Crause G. M. Adamson S. I. Cehen et al. W. D. Powers S. J. Claiborne DATE ISSUED 7-31-54 7-20-54 7-27-54 8-28-54 6-25-54 8-26-54 7-19-54 8.27.54 7-16-54 8-26-54 7-26-54 8-26-54 8-5-54 6-21-54 8-24-54 183 REPORT NO. CF-54-8-10 CF-54-6-4 CF-54-8-135 CF-54-6-164 CF-54-8-95 CF-54-8-97 CF-54-8-98 CF-54-6-216 CF-54-6-217 CF-54-7-35 CF-54-7-162 184 TITLE OF REPORT Measurement of the Density of Liquid Rubidium VYil. Radiation Damage Release of Xenon from Fluoride Fuels: Proposal for an Ex- perimental Program Heot Copocities of Compositions No. 39 and No, 101 Vill. Shielding Procedure for Obtaining Effective Removal Cross Sections from Lid Tank Data Preliminary Study of Fast Neutron Ground and Aijr Scattering at the Lower Shielding Facility Calculation of Fission Neutran Age in NeZrF¢ Age Measurements in LiF I1X. Miscellansous ANP Research Conference of May 19, 1954 ANP Research Conference of June 8, 1954 ANP Research Conference of June 29, 1954 ANP Research Conference of July 20, 1934 AUT HOR S. 1. Cohen T. N. Jones M. T. Robinson W. D. Powers G. C. Blalock E. P. Blizard Tower Shielding Group J. E. Faulkner J. E. Faulkner A. W, Savolainen A. W, Savolainen A, W. Savolainen A. W. Savelainen DATE ISSUED B-4-54 6-2-54 8-17-54 6-22-54 8-23-54 8-31-54 8-31-54 6-25-54 6-25-54 7-2-54 7-26-54 THE AIRCRAFT NUCLEAR PROPULSION PROJEGT THE OAK RIDGE NATIONAL LABORATORY SEPTCMBER 1, 1954 C. R BALLOCK® A R, sITESY 8 ANP PROJECT DIRECTOR W. H. JORDAN RD CO-DIRECTOR 5. J, CROMER RD ASSOCIATE DIRECTOR R, I, STROUGH PW ASSISTANT DIRECTOR A ) MILLER RO D. HILYER, SEC, RD REPORTS ASSISTAMT TO DIRECTOR AL W. SAVOLAINEN ARE L. 4, COOK ARE P, HARMAN, SEC. ARE o H. MCFATRIDGE, SEC, ARE LITERATURE SEARCHES A, L, DAYIS ARE MRCRAFT REACTUR ENGINEE RING DIVISION 5. 1. CROMER, DIRECTOR RD H. MCFATRIDGE, SEC. ARE SUPPORTING RESEARCH ¥, H. JORDAN A J.MILLER ASSISTANT TQ DIRECTOR L L. M. COOK ARE 1 | i i { ! I i i i REACTOR PHYSICS ENGINEERING DESI Gal POWER PLANT ENGINEERING EXPERIMEN TAL ENGINEERING ARE OPERATIONS STAFF ASSISTANT STAFF ASSISTANT | STAFF ASSISTANT STAFF ASUSTANT STAFF ASSISTANT STAFF ASSISTANT W. K. ERGEN ARE H. C. GRAY P¥ A, P. FRAAS ARE H. W. SAYAGE ARE E.5. BETTIS ARE W, R. GRIMES WC W.D. MANLY " l E. P. RLIZARD P A, J. MILLER RD W. K, ERGEN ARE H. F. POPPENDIEK REE J. L. MEEM ARE J. BENGSTON ARE A. A. ABBATIELLO ARE C. 5. BURTNETTE USAF ¥. G, COBB ARE P. HARMAM, SEC. ARE R. K. OSBORN ARE J. Y. ESTABROOK ARE R. W. BUSSARD ARE 3. A CONLIN ARE R. E. HELMS ARE W. T, FURGERSON* K-25 C. P. COUGHLEN ARE G, A, CRISTY EMm COMPUTERS G L. HOLLIS ARE M, E. LAVERNE ARE R DREISBACH P R. E. HARRIS EM CHEMISTRY METALLURGY SHIELDING RESEARCH RADIATION DAMAGE REACTOR CALCULATIONS HEAT TRANSFER AND PHYSICAL F. L. MAGLEY ARE F. W, MCQUILKEN ARE A. G, GRINDELL ARE E. R MANN® i W, R. GRIMES re . PROPERTIES RESEARCH A. FORBES ARE C. A MILLS ARE C B MLLS ARE /. HELTON ARE DAY SHIET T - £ D BLIIARD z T e « & A ChaRPE ® H. F. POPPENDIEK REE M. TSAGARIS ARE J. A. GLSON P G, SAMUELS ARE . R, HUNTLEY ARE P. A, AGRON ¢ G. M, ADAMSON M L. S ABBOTT P i . - H. R. ROCK PW W. L. 5COTT ARE J. W, KINGSLEY ARE w. B, COTTRELL, SUPV. ARE C, J. BARTON MC c R BOSTON M J.EI FAULKNER P R. R. COVEYOU ARE CONSULTANTS W, J. STELZMAN ARE E. M. LLEES ARE J. G. GALLAGHER ALC J. P BLAKELY MC w' H’ BRIDGES M . HI C. D. BAUMANN 85 C. L. GERBERICH" MP J. 0. BRADFUTE REE B. M. WILNER AGC R, E. MacPHERSON ARE H. E. HUNGERFDRD P F.F. BLANKENSHIP MC "o R Do M F. H. HURRAY P W. E. BROWNING 55 M, . GIVEN* P S, 1. COHEN REE T. ANDERSON M. M, YARDSH ARE L. A, HANN ARE $. C. SHUFORD P €. M. BLOGD we O e " ne.steRn & ¥. E. BRUNDAGE 55 W, E. KINNEY ARE N. D, GREENE REE A. NOHEL, GEQRGIA TECH, E. M. EISSEMBERG, SEC. ARE %. B. MCDONALD ARE A, L, SOUTHERN éERE g‘ :. :g;g:fl ME v AR ARF " R, RICKMAN, SEC 5- :‘ g:sgou- gg W. C. SANGREN MP g- 5 :S?LLJ&N :EE W.R. C. STORRS P, M / . . , SEC. W ™ CONSULTANTS o DesoRN ARE - I . BRATCHER MC R B, CLAUSING . M. J. FELDMAN 5§ B LYNCH REE DL L H. . E. . D, PALM A. H. FOX, UNION COLLEGE O. F. SALMON 4RE ECHMICIARS ol e . H, COOK " CONSULTANT R fifixms e CRITICAL EXPERINENTS W, 0. POWERS Ree W, LOWEN, UNION COLLEGE P.G. SMITH ARE 4 5. ADDISON ARE - X 1. H. DE VAN M H. A, BETHE, CORNELL UNIVERSITY U IR e MW, ROSENTHAL REE R L. MAXELL E.$TORTD ARE J. R. CROLEY ARE D. R. CUNEQ e . DO AnaRE W C.L.HIL A. . CALLIRANY P . ¥ UNIVERSITY OF TENNESSEE W T TUNNELL ARE F. J. SCHAFER ARE F.L.DALEY e o n BOUGLAS . m hE :E:E:guz g: T. K. CARLSMITH, SEC. REE 6. F. WISL ICENUS D. R, WARD ARE C. F. WEST ARE J. E. EORGAN W 2 L HLESTAND " CONTRACTORS 8. % KEWLHOL 5 . 5. CRUDELE PW TECHNICIANS PENNSYLVANIA STATE URIVERSITY D. ALEXANDER, SEC. ARE EVENING SHIFT H. A PRIZOMAN e E.E. HOFFMAN M METAL HYDRIDES, INC. M. T. MORGAN 55 9. L. DREENSTREET ARE C. 6. BLALOCK REE D. HARRIS, STENO. ARE F. KERTLSZ He H. INOUYE " NUCLEAR DEVELOPMENT ASSOCIATES, . OLson « V. G, HARNESS* P R W BURNETT REE D. STOREY, REC. CL. ARE G. D. WHITHAN, SUPY. ARE E.E.KETCHEN MC C.F.LEITTEN " iNC. 3 e 1. TN e . J. CLAIB A. MONTGOMERY, CL. TYP. ARE K. LESLIE ARE F. A, KNOX e R. B. OLIVER M M. . OSRORME 5 . H. MARABLE* P - J. CLAIBORNE REE E. L. MORRISON USAF S. LANGER MC N B TRIAGCA " W. W. PARKINSON 55 1 W NOAKS Pw T. N. JONES REE TECHNICIANS F. A, FIELD UsAF R, E. MEADOWS MC o n H. E. ROBERTSON 55 E. R. ROHRER* P J. LONES REE T. ARNWINE ARE G. C. ROBINSON ARE R, P WMETCALE Mg K. W. REBER " LID TaNk W o s D. SCOTT, JR. ARE i M G. 5. CHIL F.W. MILES MC - - - > R. M. SPENCER USAF - ! aRE TECHNICIANS R, E. MODRE M C- . SLAUGHTER " G. 7. CHAPHAN P DSCAR SISMAN 5§ DV, P WILLIAS® P . M. CUNNINGHAM ARE T.E. CRABTREE ARE G, J. NESSLE e Wt STEIOLITZ " L. . TEMPLETON 5 E. L. ZIMMERMAN® P J. R. DUCKWORTH ARE R. E.DIAL ARE R. F. NEWTOH RD A TABODA " 1. B. DEE PW C.C. WEBSTER 55 M. L. RUEFF,* SEC. P %K. R, FINNELL ARE ., . JENKINS ARE L. G. OVERHOLSE R Us W, WOODS " J. M. MILLER P R. A WEEKS 55 W, D, GHORMLEY ARE £. . PERRIN ARE 4. 0. REDMAN M "1 THOMAS, SEC M D. K. TRUBEY r 3 & wieon o R. 1. SHEIL ue - : .C. C. J, GREEN e < o L :EE NWGHT SHIFT M.V, SMITH e TECHNICIANS TECHNICIANS J. C. ZUKAS S5 R. A. HAMRICK ARE R. G, AFFEL, SUPY, ARE B. L STURM M W. A. ANDERSON M E. BECKHAM P TECHNICIANS 8. L. JOHNSON ARE 4. W. ALLEN ARE 1. E. SUTHERLAND MC G. D. BRADY “ R, TAYLOR e LLis 55 W. 1. MAYNARD ARE R.L. BREWSTER ARE R, E. THOMA e 2. TLEAST M 1%, WAMPLER . B D. E. MCCARTY ARE J. M. CUNNINGHAM ARE L. E. TOPOL MC M. GOMZALEZ M J. SELLERS Ic F. M. BLACKSHER 55 6. E. MILLS ARE 4. J. HARSTON REE 1. TRUITT MC L L HALL M C. C. NANCE ARE - G. #h WATSON MC 5. D, HUDSON W CONSULTANT J. 1. PARSONS ARE TECHNICIANS C. F.WEAVER MC R. W. JOHNSON M D. F. WEEKES, TEXAS A & M COLLEGE M. A, REDDEN ARE T. L. GREGORY ARE D. E. CALDWELL, SEC. MC .G LANE " BULK SHIELDING FACILITY g- ?;T%i':féw ARE g- :2:5 ke TECHNICIANS 5. MCOONALD ¥ 1 &6 cocHRAN P . G AR . . MCNA * B. C. WILLIAMS ARE R, K. BAGWELL MC B Hotnet : F. €. MAIENSCHEIN P J. M, DIDLAKE MC . CONSULTANTS F. A DOSS ARE ‘L: ;' ?’;‘é‘fig " G. M. ESTABROOK 3 4. F. BAILEY, UNIVERSITY OF TENNESSEE CONSULTANT 1. P, ELBANKS MC R Thomas " M. P. HATDON* P ¥, R. CHAMBERS J. H. BUCK, WELL SURVEYS, INC. B, F. HITCH MC W ALKER . 1.0, FLYNN P J. F, HAINES W JEHNINGS MC - K. . HENRY P W. K. STAIR, UNIVERSITY OF TENNESSEE F. G. KILPATRIC MC E. B, JOKNSON P G, A, PAIMER MO METALLOGRAPHY C. BOUNDS, 5EC. P REACTOR COHTROL B, C. THOMAS MO CONTRACTOR ) ) C H. TIPTON e R. 1. GRAY M TECHNICIANS YITRC CORPCRATION E.P.EPLER < R. A, WILEY ARE R. 5. CROUSE I H. JARVIS Ic T. M. KEGLEY M P e F. P, GREENT I CONSULTANTS C.L.LONG L R, th SIMMONS P 5. H. HANAUER® ic J. M. CARTER, CARTER LABORATORIES TECHNICIANS S. SMIDDIE Ic E. R. MANN® Ic D, G. HILL, DUKE UNIYERSITY G. G. STOUT ic W. F. MRUK! Ic H, INSLEY LR " H. WEAVER P L. C. GAKES* Ic : . F. LG SaKES, Ic T. N, MCVAY, UNIVERSITY OF Al ABAMA L P zeeRALD " R. S, STONE* Ic CONTRACTORS B C. LESLIF M C.S. WALKER® ic TOWER SHIELDING FACILITY P. GROOVER " SEC. e AME'Sr CULE MR STITUTE CONSULTANTS C. E. CLIFFGRD P NOTE: THIS CHART SHOWS ONLY THE LINES OF TECHMICAL COORDINATION OF THE ANP PROJECT, THE H. A, BISHOP,* DRAFTSMAN IC g:gékfig‘ggfi%‘fl'—lgg ! - YARIOUS (MDIYIDUALS AND GROUPS OF PEOPLE LISTED ARE ENGAGED EITHER WHOLLY OR PART TIME ON METAL HYDRIDES, INC, N. CABRERA, UNIVERSITY OF V”‘G““T'Ar T. ¥. BLOSSER P RESEARCH AND DESIGN WHICH IS COORDINATED FOR THE BENEFIT OF THE ANP PROJECT IN THE MANNER MOUND LABORATORY e A S n TUTE F. N. WATSON P INDICATED ON THE CHART. EACH GROUP, HOWEVER, IS ALSO RESPONSIBLE TO IT5 DIVISION DIRECTOR FUEL REPROCESSING A R NIZHOLS, SANDIEGD STATE ). L. GREGS, CORNELL UNIVERSITY L. B poLLAND k FOR THE DETAILED PROGRESS OF ITS RESEARCH AND FOR ADMINISTRATIVE MATTERS, —] F R BRUCE LNIVERSITY OF ARKANSAS 2 LNS:Ii;ES 3. VAN HOOMISSEH BAT . F. . M. F. VALERING NACA M. R, BENNETT cr RENSSFLAER FOLYTECHNIC INSTITUTE B L. GILLILAND GE THE KEY TO THE ABBREVIATIONS USED IS GIVEN BELOW. G. | CATHERS cr W. F. SAVAGE, E. MCBEE, SEC. P D. E. FERGUSON cr ANAL YTICAL CHEMISTRY RENSSELAER POLYTECHNIC INSTITUTE C. E. LITTLEJOHN T . A E. C. WRIGHT, UNIVERSITY OF ALAGAMA TECHNICIANS ANP STEERING COMMITTEE AC ANALYTICAL CHEMISTRY DIVISION ~ ORNL T. LONG T ©. D. SUSANOD < P, C. SHARRAH, 1 N, MONEY P T R. P, MILFDRD CcT 8. £. YOUNG,* SEC AC UNIVERSITY OF ARKANSAS AGC AEROJET.GENERAL CORPORATION O b WATSON T E.D. CARRO:‘- ic W. H. JORDAN, CHAIRMAN ALC AMERICAN LOCOMOTIVE COMPANY TECHNICIAN 1. C.WHITE AC G. G. UNDERWOOD '« ARE AIRCRAFT REACTOR ENGINE ERING DIVISION — ORNL s CaLowect o D.E CARPENTER re CONTRACTORS E.S. BETTIS BAC BOEING AIRPLANE COMPANY — D, L. MANNING AC BALDWIN-LIvA-tAMILTON CORPORATION o e aove € CHEMISTRY DHVISION - ORNL R. 1, MECUTCHEN® AC g;a;fi;fi?fmfl%o%fi«?w 5. J. CROMER CT CHEMICAL TECHNOLOGY DIVISION - ORNL B. 1. MCDOWELL AC . K. ERGEN A. 5. MEYER 4 AC TERROTHERM ¥. . EREN CV CONSOLIDATED VULTEE AIRCRAFT CORPORATION . PEELE AC MASSACHUSETTS INSTITUTE YR GRIVES EM ENGINEERING AND MAINTENANCE DIVISION - ORNL . 1. ROSS AC oF TEC”"OLSOG"C W. D, MANLY W. . VAUGHAN AC METAL HYDRIDES, INC. GE GENERAL ELECTRIC COMPANY R WILLIAMS AC RENSSELAER POLYTECHNIC INSTITUTE : .Js ‘,‘Jééfé’umsx e {NSTRUMENTATION AND CONTROLS DIVISION - ORNL E. M. ZARZECKI, SEC. AC SUPERIOR TUBE COMPANY H. W, SAY M METALLURGY DIVISION — ORNL GLENN L, MARTIN COMPANY E.D. SHIPLEY TECHNICIANS MC MATERMLS CHEMISTRY DIVISION — ORNL 7. P. BACON AC JA- ?i. -‘;:g::‘;gg; MP MATHEMATICS PANEL - ORNL R. G. BRYANT AC CERAMIC RESEARCH NACA MATIONAL ADVISORY COMMITTEE FOR AERONAUTICS M. R. CHILDS AC L ) P PHYSICS DIVISION — ORNL t *;A ."5‘3& :2 J. M. WARDE M PW PRATT AND WHITNEY AIRCRAFT DIVISION — UAC M. A MARLER AC €. E. CURTIS™ M RD RESEARCH DMRECTUR'S DEPARTMENT - ORNL T.G. MILLFR AC L. M.DOREY® M REE REACTOR EXPERIMENTAL ENGINEERING DIVISION - ORNL ;7 3 ;FSB:ER :g > ?‘ mflii‘m" :: s1 STABLE ISOTGPES DIVISION - ORNL i M. P. HAYDON* P $S SOLIG STATE DIVISION = GRNC CONSULTANT & R. fir‘:NSON“ " USAF UN{TED STATES AR FORCE H. H. WILLARD, UNIVERSITY OF MICHIGAN o5 witer " SPECTROGRAPHIC ANALYSIS A. HOBBS,” SEC. "‘ . PART TIME iR, MCNALLY® o TEGHNICIAN . J. A, NORRIS® H J. A, GRIFFIN " AND OTHERS CONSULTANTS MASS SPECTROMETRY T. N. MCVAY, UNIVERSITY OF ALABAMA T. 5. SHEVLIN, OHIO STATE UNIVERSITY