fi ORNL-2061, Part |, 11, 111 z 133 -4 C-34 Reactors-Special Fuch.lres of Aircraft Reactors AEC RESEARCH AND DEVELOPMENT REPORT- TR ¢ 34 O =S E-.‘-'_.':f ;g‘ 3 4456 0251032 § o - i E!"E' ( "‘3}" g - ey J“"‘ rfi: AIRCRAFT NUCLEAR PROPULSION PROJECT e iy - R J QUARTERLY PROGRESS REPORT o™ g gjhfi FOR PERIOD ENDING MARCH 10, 1956 5 - 55| £ 5] 3 s 5 & & Iy | iopdl OAK RIDGE NATIONAL LABORATORY | OPERATED BY UNION CARBIDE NUCLEAR COMPANY A Division of Union Carbide and Carbon Corporation ucc) - POST OFFICE BOX P * OAK RIDGE, TENNESSEE 4o 150 5 ORNL-2061, Part |, 11, I} C-84 — Reactors-Special Features of Aircraft Reactors This document consists of 272 pages. Copyl 3301‘ 330 copies. Series A, Contract No. W-7405eng-26 AIRCRAFT NUCLE AR PROPULSION PROJECT QUARTERLY PROGRESS REPORT For Period Ending March 10, 1956 W. H. Jordan, Director S. J. Cromer, Co-Director A. J. Miller, Assistant Director DATE ISSUED mAY 23 1958 OAK RIDGE NATIONAL LABORATORY Operated by UNION CARBIDE NUCLEAR COMPANY A Division of Union Carbide and Carbon Corporation Post Office Box P ' : S dge, Tennessee MARTIN MARIETTA ENERGY SYSTEMS L1 i 3 4456 0251032 5 { i C-84 — Reactors-Special Features of Aircraft Reactors INTERNAL DISTRIBUTION 1. Rn G. Affel 2. C. R Baldock 3. C. J. Barton 4. M. BenderA 5. D. S. Billington 6. F. F. Blankenship 7. E. P.Blizard 8, C., J. Borkowski 9. W. F. Boudreau 10. G. E. Boyd 11. M. A. Bredig 12. W. E. Browning 13. F. R. Bruce 14, A, D. Callihan 15. D. W. Cardwell 16, C. E. Center {(K-25) 17. R. A. Charpie 18. G. H. Clewett 19. C. E. Clifford 20, J. H. Coobs 21. W. B. Cottrell 22. D. D. Cowen 23, S. Cromer 24, R. S. Crouse 25. F. L. Culler 26. J. H. DeVan 27. L. M. Doney 28. D. A. Douglas 29. E. R. Dytko 30. L. B. Emlet (K-25) 31. D. E. Ferguson 32, A. P. Fraas 33. J. H. Frye 34, W, T. Furgerson 35, H. C. Gray 36, W. R. Grimes 37. E. E. Hoffman 38. A. Hollaender 39. A.S. Householder 40, J. T. Howe 41, H. K. Jackson 42. W. H. Jordan 43. G. W. Keilholtz t 44, C. P. Keim ' Y 45, M. T. Kelley 46. F, Kertesz ORNL-2061, Part 1, 11, Il 47. 48-49. 50. 51. 52. 53. 54. 55. 56. 57. - 58. 59. 60. 61. 62, 63. 64. 65. 66. 67. 68. 69. 70. 71. 72. /3. 74. 75. 76. 77. 78. 79. 80, 81. 82. 83, 84. 85, 86. 87. 88. 89. 90, 91. 92. 93. R. Van Artsdalen " C. VonderLage M. Watson E. M. King J. A. Lape R. S. Livingston R. N. Lyon F. C. Maienschein W. D, Manly E. R, Mann L. A, Mann ¢ W, B, McDonald F. R, McQuilkin R. V. Meghreblian ? R. P. Milford A, J. Miller R. E. Moore J. G, Morgan K. Z. Morgan E. J. Murphy J. P. Murray (Y-12) G. J. Nessle R. B. Oliver L. G. Overholser . P. Patriarca R. W. Peelle A. M. Perry ' J. C. Pigg H. F. Poppendiek P. M. Reyling A. E. Richt M. T. Robinson H. W. Savage ’ -A, W, Savolainen ‘R. D. Schultheiss é D. Shipley L A Simon 0. §isman M. Ji. Skinner G. P, Smlth A, H. $ne|| C.D. %sono J. A, Swflrfout E. R. D. E. F. G. %. A. M. Weinberg 100. 9%, J. C. White 101-110. 963G, D. Whitman | 97. “E. P. Wigner (consultant) 111-131. 98. GAC. Williams 132. 99, J. 133-135. Wllson F Plant Representative, Seattle 55 . Air . Air . Air Red . . Air Tech’iycal Intelligence C*enfer 146. Aircraft L&bora’rory De5|gr,| Branch 147-149. ANP Projeck Office, Forf Worth 150. Argonne Non_‘jal Lobom‘rory 151. Armed Forces.! 152. Assistant Secref 153-159, Atomic Energy- 160. Battelle Memor 161, Bettis Plant 162. 163, 164. 165. 166, 167-168. ateriel Area ¥ Chicagg .percmons O Cthdgo ‘Patent Group © : Chlefd‘ff Naval Research EXTERNAL DISTRIBUTION C. E. Winters ORNL - Y-12 Technical Library, Document Reference Section Laboratory Records Department Laboratory Rec¢ords, ORNL R. C. Central Research Library . AF Plant Representative, Baltimore /. AF Plant Representative, Burbank AF PlantsRepresentative, Marietta “/AF Plant Representative, Santa Monica . AK Plant Representative, Wood- Rldge search and Development. Command (RDGN) earch and Developmenf 'Command (RDZPA) (WADC) ecml!Wedpons Project, Sandia ary of the Air Force, R&D Amission, Washington 169. Cor}vg ir-General Dynamics G ,rporcmon 170. 171. 172, 173-1754, Dl,rét'ror of Laboratories (WC‘”-:.( Dl’récfor of Requirements (AFD {3irector of Research and Develé‘_' Pirectorate of Systems Manageme “"‘f‘(RDZ ISN) 176-178./ Directorate of Systems Management {RDZ-1SS) 179 180-,1’ /}‘34. #1185, £ 186, 187. 188. 189. 190. 191. 192. 193. 194, Equipment Laboratory (WADC) General Electric Company (ANPD) Hartford Area Office ldaho Operations Office Knolls Atomic Power Laboratory Lockland Area Office Los Alamos Scientific Laboratory Mound Laboratory Naval Air Developmenf Cem‘er Headquarters, Air Force Special Wec:pons Can’rer 7\",‘:4?‘3 AN T . T o Materials Laboratory Plans Office (WADC) “\ National Advisory Committee for Aeronautics, Cieveiahd National Advisory Committee for Aeronautics, Wcshlngtol}& 195. 196, R 197. 198, 199-201. 202-205. 206, 207, 208. 209. 210, 211. 212-214, 215-329. 330. § e York Operations Office #F 7 American Aviation, |g. (Aerophysics Division) ion (Fox Project) Rhations Office f i"'ry of California Radia™® Laboratory, Livermore Air Development Center (WCOSI-3) nical Information Extension, Oak Ridge lvision of Research and Development, AEC, ORO Reports previously issued in this series are as follows: ORNL-528 ORNL-629 ORNL-768 ORNL-858 ORNL-919 ANP-60 ANP-65 ORNL-1154 ORNL-1170 ORNL-1227 ORNL-1294 ORNL-1375 ORNL-1439 ORNL-1515 ORNL-1556 ORNL-1609 ORNL-1649 ORNL-1692 ORNL-1729 ORNL-1771 ORNL-1816 ORNL-1864 ORNL-1896 ORNL-1947 ORNL-2012 Period Ending November 30, 1949 Period Ending February 28, 1950 Period Ending May 31, 1950 Period Ending August 31, 1950 Period Ending Decembker 10, 1950 Period Ending March 10, 1951 Period Ending June 10, 1951 Period Ending September 10, 1951 Period Ending December 10, 1951 Period Ending March 10, 1952 Period Ending June 10, 1952 Period Ending September 10, 1952 Period Ending December 10, 1952 Period Ending March 10, 1953 Period Ending June 10, 1953 Period Ending September 10, 1953 Period Ending December 10, 1953 Period Ending March 10, 1954 Period Ending June 10, 1954 Period Ending September 10, 1954 Period Ending December 10, 1954 Period Ending March 10, 1955 Period Ending June 10, 1955 Period Ending September 10, 1955 Period Ending December 10, 1955 FOREWORD This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL records the technical progress of the research on circulating-fuel reactors and other ANP research at the Laboratory under its Contract W-7405-eng-26, The report is divided into three major parts: |. Reactor Theory, Component Development, and Construction, Il. Materials Research, and Ill, Shielding Research. The ANP Project is comprised of about 550 technical and scientific personnel en- gaged in many phases of research directed toward the achievement of nuclear propulsion of aircraft. A considerable portion of this research is performed in support of the work of other organizations participating in the national ANP effort. However, the bulk of the ANP research at ORNL is directed toward the development of a circulating-fuel type of reactor, The design, construction, and operation of the Aircraft Reactor Test (ART), with the cooperation of the Pratt & Whitney Aircraft Division, are the specific objectives of the project. The ART is to be a power plant system that will include a 60 Mw circulating- fuel reflector-moderated reactor and adequate means for heat disposal., Operation of the system will be for the purpose of determining the feasibility, and the problems associated with the design, construction, and operation, of a high-power, circulating-fuel, reflector-moderated aircraft reactor system. vii $ A CONTENTS FOREWORD SUMMARY -------------------------------------------------------------------------------------------------------------------------------------------------------- ............................................................................................................................................................ PART I. REACTOR THEORY, COMPONENT DEVELOPMENT, AND CONSTRUCTION 1. REFLECTOR-MODERATED REACTOR ART Facility Design and Construction ... sieies s stiene e s er s e s esbesasesssnen e AR T DS igN ettt ete s e bbbt e e et e s s et et b b eanteeates SIS S ANGIYSES oo e e bbb et en e s ee e ne e st te et e taerestaebtenens SOdIUM SYSTOM ettt e etk et bR b e e nae et s et et ensenrens REACOr Shield ..ot et ba e s bbb b eab e ny b e e s e e et sranebe e b beeteetnas ------------------------------------------------------------------------------------------------ ART Component Development, Procurement, and Testing ..c.ccccoeoereininnineencne e Water Test of Aluminum Mockup of North Hedd........o.vveioii s Core Flow SHUIES .ottt e e eas bbbt s v e sae b aree e st e nnsbesraseasaetnsrerens Engineering Test Unit. ittt Procurement of Special Reactor Materials and Components ........cccoeveeiiiiniiicinceececicinreena, Inspection of Materials and Fabricated Parts ..o ART Instrumentation Gnd Comtrol oo eooeeeeeeereterssesessaeesssseesssesansssereessseesssserssasassssasassresssesesnsareess REACIOE PRYSICS oottt ettt v et e et e e ses it ek aeaeeaeesebbeseebenenseae e e ae e e enssane NaK Activation in the Fuel-to-NaK Heat Exchanger of the ART ..o, Activity of Mass-Transferred Material in the ART Radiators ..., Radiation Heating in the ART ..t sttt s Self-Absorption of the Decay Gamma Rays in the ART Fuel Dump Tank ..coocovviiicincnneinncennnn, Activation of the Sodium in the ART ..ot e sr s s aenen Shutdown Reactivity of Lithium in the ART Reflector Coolant ..o Forced-Circulation Corrosion and Mass-T ransfer Tests vttt ee e es e s sbae Fused Salts inm IReonel .o e et e s e s estabtsssase st assserasaeseaentasntasaes s eaasnsasnrsseeeesrrnnannees Liquid Metals in Inconel and in Stainless Steel ......cccocvccvvvrviienneen. e eeeiee et ee e as i nr ot aeteeteannenes PUMP DEVEIOPMENT .oiieiiieeeeeeee ettt s et b et st e ab s ee e ses b ae bbb e e e enr s eneenens Bearing-and-Seal Tests ...t e b Sodium-Pump Performance Tests with Water ..ottt NaK-Pump Performance Tests with Water ..ot aene e High-Temperature Tests of ART Fuel Pumps oo High-Temperature Pump-Performance Test Stands ..., Heat Exchanger Development ..o e Intermediate Heat Exchanger Tests ..ot e Small Heat Exchanger Tests ..ottt s e Heat-Transfer and Pressure-Drop Correlations ......ocoiiiienerriinieineisriece e vii 15 15 21 22 23 23 23 23 23 24 24 25 26 26 28 28 30 35 37 39 39 41 41 41 41 42 42 42 43 43 43 45 48 48 51 52 52 52 54 Structural Tests Outer-Core-Shell Thermal-Stability Test ..oocoviiiirec e e Inconel Strain-Cycling Tests .otk s e s Swagelok Tubing-Connector Tests ... s e Reactor Component Development ... DUMP Y alV@ et ettt e e e s e a e bR e e e e e ea et et eeae b e e e s bt Cold Trap and Plugging Indicator ..ottt e s Zirconium Fluoride Vapor Trap .ot ente e i ece et rear et sa b e ees e 3. CRITICAL EXPERIMENTS Lt e Compact-Core Reflector-Moderated-Reactor Critical Experiments......ccoooiiiiiinniiiniiinn N EUIEON=F LUX DiStribUIOM o viviuiireee ettt ee et e sr e sb e s b et s e b b rme e s men e s e emrnesas o neeenansseres seneeens Mass-Reactivity Coefficients of Reactor Component Materials .....cccoooiiivviiiiiii PART Il. MATERIALS RESEARCH 4, CHEMISTRY OF REACTOR MATERIALS ..ot Phase Equilibrium StUdies . ..ottt The System UF j-UO, ot The Systern NGF-UF4 ............................................................................................................................ The System KE-ZrF ;.o The System RBF-ZIrF 4. b s The System NaF-ZrF (-UF ;oo The System NaF-KF-ZrF ;o The System NaF-RBF-ZrF ; oo s The System RBF-BelF , o s The System NaF-BeF ,-UF ;s The System NaF-LiF-BeF , o The System NaF-LiF-BeF j-UF | oo The System NaF-LiF-CaF , oo The System NaF-MgF ,-CaF .o The Systems N(JF-CeF3 and RBF-CeF 5 oo The System LiF-NiF , oo Optical Properties and X-Ray Patterns for Compounds in Fluoride Systems ...ccocovviciiiiccnnnnenn. Chemical Reactions in Molten Salts.. ..ot Equilibrium Reduction of FeF , by H, in NaF-ZrF , ..., EMF Measurements in Fused SQIts .ot ev e et e e e ene s Activity of Chromium in ATOys ..o sttt en st seees Reduction of UF, by Structural Metals........coooiiiiiiii s Reaction of UF ; with NaF-KF-LiF Eutectic i Experimental Preparation of FIuorides ...ttt st s rasnees Color Studies of Alkali Metals in Equilibrium with Alkali Halides at High Temperatures .............................................................................................................................. Physical Properties of Molten Materials ..........ccceeiiiieiiciinec s e Vapor Pressures in the System KF-ZrF ;oo Vapor Pressure of FeCly, o) Surface Tensions of Molten Salts ..o e eab e Density and Surface Tension of Molten KCI at 800°C Physical Chemistry of Fused Salts ...................................................................... ........................................................................................................ 58 58 60 60 60 60 62 63 64 64 64 64 Production of Purified Fuel Mixtures .....oo.oiiieieeriicrrcec et 107 Use of Copper-Lined Equipment for Fuel Purification ..., 107 Preparation of ZrF , by Gas-Phase Hydrofluorination of ZrCl, .cccovmiceriiiiiiis 109 Laboratory-Scale Purification Operations ... s 109 Pilot-Scale Purification Operations ...t sss e 109 Production-Scale Operations .........c.oeieiineeeceeree st ens s arab bbb 110 Quality Control of Raw Materials and Products .......cccooiviiriciiiiniiiic 111 Batching and Dispensing Operations ... iciicnceicins st ss e m Filling, Draining, and Sampling OPErAtiONS .v.veivieierteeeteeeeeceietetete et st bet bbb bbbt enae e 112 Calibration of ART ERFICher ..ottt e st et s bbb s 112 . CORROSION RESEARCH ...ttt ettt ottt st e et sa b 114 Forced-Circulation STUIies ..ottt bbb 114 Fluoride Fuel Mixtures in INComel ..ot et eener e e 114 Liquid Metals in Inconel and Stainless Steel ... 117 Thermal-Convection StUdIs ...cciiiicirciencr sttt 118 Alkali-Meta!l Fluoride Fuel Mixtures in Hastelloy B ..cocoooeiiiii 118 Effect of Condition of Inner Surface of Hastelloy B Tubing on Depth of Attack .....c.cccoevrieneen. 119 Screening Tests of Special Fuel Mixtures ..., 120 General Corrosion STUdIES ..oiiiire ittt ettt ee e sttt bbb bbbt 122 Low-Cross-Section Brazing Alloys in Liquid Metals and in Fluoride Fuel Mixtures .................. 122 Cathalloy A-3T in SOAIUM ceiceeicicii e b 122 Rare-Earth Oxides in Sodium .ottt 124 Sodium in Type 316 Stainless Steel Thermal-Convection Loops ......cccccccovniiiiiiienierciecicce, 125 Boiling Sodium in Type 348 Stainless Steel Loops ..o 127 Compatibility of Hastelloy B and Beryllium in Sodium ... 128 Solid-Phase Bonding of Cermets oot ses e b e 129 Solubility of Lithium in NaK ..o 130 Mass Transfer and Corrosion in Sodium Hydroxide ......ccooviiniie 131 SHAtic CaPSUIE TeSES ittt bbb 131 Thermal-Convection Loop Tests .. ettt e et st 133 Chemical Studies of CorroSion ..o.iceiriieenreeiere s ereseseeee st sssise st st e r st s bbb s s snsbsnes 134 Physical Properties of Elastomers Exposed to Attack by Liquid Metals ...cocoevineiiiiincin, 134 Resistance of Possible Moderators to Fused Fluoride Mixtures ..o, 137 Diffusion of Chromium in AlIOYS «eveieeeieireice ittt snens 138 . METALLURGY AND CERAMICS .....oiiiitieieiere ettt ias s et et 139 Welding and Brazing STUdies ... 139 Examination of NaK-to-Air Radiators After Service at High Temperatures.....cooceoiiinniinii 139 Crack Susceptibility of Back-Brazed Tube-to-Header Joints ..o, 139 Measurement of Weld Shrinkage ...t 140 Dimensional Control During Fabrication of Pump Volutes ..o 142 Carmet-t0mMetal JOTNES troeoiii ittt st sttt et s s b r e s v e e st 143 Brazing Alloy Development ...t e 144 Mechanical Properties of InConel .....cc.cccviiiiiniiii s 146 Creep-Rupture Design Data ..ot 146 Low-5tress Creep Data .o iiieiccciii i s 146 Effect of Biaxial Stress on Creep Ductility at High Temperatures ....cccovroninniniiininnins 151 - Xi Special Materials StUies ... e 151 Neutron Shield Material for High-Temperature Use ... 151 Fabrication of Boron-Containing Materials ..ot e 152 Inconel-Boron Compatibility ..o e et e 154 Nickel-Molybdenum-Base Alloys .........cccccioniiiiii e 155 N O UM F OB O I ON e et a s e s rar e st sbany e seee e e raae s ben e e rmnase s baneanses 160 Seamless Tubular Fuel Elements ... cenie s cesee s st s eeaeseasane s e sbnanineaane 161 Control Rod FObriCtion ciieeeeieieivveeteeimeteise s eeer s nseesesaaestsentsteasssesesnesassrerssesasessmestesarasesssesssransesanases 163 Shield Plugs for ART PuUmps ..o ittt st 163 Lithium-Magnesium Aloys ..o s s 163 Fabrication of Ceramic Materials ......ocivireiiiiieiee e e e 164 NONAESTIUCTIVE T@STING oottt st s em e s s b e s s s s e s sn e et s nresaneneens 164 . HEAT TRANSFER AND PHYSICAL PROPERTIES ...ttt ssranees 171 Fused-Salt Heat Trams er. ettt ettt e e bbb s e bba b ee s e se e ba e be e e sr e e s bbabba st e e sabaesmaese e 171 ART Fuel-to-NaK Heat EXChanger .........cocoiiiiiiiiiiire it et 172 ART Core Hydrodynamics ..occeoveiii it e bbbt e e 174 Temperature Structure in Region Beyond the ART Reflector ..o 174 Temperature Structure in an ldealized ART Core...oooiiiii e 176 ART Core Heat-Transfer Experiment ..ottt e e 176 HEAE CAPACIHTY ooviviiieieere st ee s er et s s e e e s aeese s s en e ar s e na et e e st eat nreaneneessenreereranes 176 VESCOSITY Lovrreainniiieieietetes et s e c bt e om e s b s bbb bbb e b s bt des Sha b a4 b b eat et bR e b e S b e b e e R bR e et e b e n Rt en e 178 Performance Comparisons of Several Fluoride Fuels ..o, 179 RADIATION DAMAGE ...ttt ettt s eaesteare s re et at e e e esteebeemeeesennas 181 Review of Tests of Corrosion of Inconel and Stability of Fuel Under Irradiation.........cccceevvrennnn.e. 181 Examination of the Disassembled MTR 1n-Pile Loop No. 3.t 183 Effects of Radiation on the Mechanical Properties of Structural Materials .....cccooooivieneee 187 SHPES S COMTOSTON T@STS 1oviviiieieieeees et ee it e e e e e ettt e st e e eee e eee e e e e eeaeesarestneesten e snsebtsstesesseeseenssens seneornaeons 187 Alternate Stress-Corrosion APParatus ... i ce e ieie et es e e s sse s s s e s e e s sresessssase s smeeeesnes 188 MTR Tensile Creep TeStS . airiairieseee e e e erreess e e srrsbeesessesasmateseressasssseneessarsesssstessasarens 190 Preumatic SHressing DeVice ettt et st ene e ean 191 Ductility of Nickel ATTOYs .ottt et esne e st st aocarasr s e nnees 191 Experimental Studies of Reactor Materials and Components .......ccocccoieiiniiiiciinnreseee e 192 Effect of Radiation on Stotic Corrosion of Structural Materials by Fused Salts ... 192 Holdup of Fission Gases by Charcoal Traps......ccciioieieiicicceicrteee sttt e 193 LITR Vertical In-Pile Loop ottt e 196 Instrumentation for ART Off-Gas Analysis .ciiriiciiiiiiecrecenec et 196 Inconel as a Thermal-Neutron-Flux Monitor.......ccoe ittt e 197 Measurement of MTR Flux near Tip of In-Pile Loop No. 3 et 198 Fast-Flux Measurements in Hole 19 of ORNL Graphite Reactor ......cccoevmeevvevcviinicveicieceeiee, 198 Chemical Effects of Nuclear Reactions ... it eva st s ra e 199 Effects of Fission Products on Properties of Fluoride Fuels ..o 199 Use of Natural Lithium in Fluoride Fuels Circulated in In-Pile Loops .....coocoiiiciiciiiien 203 Radiation Damage to Boron Carbide......cccuuiieriinicriniieriii i ete e e saen et e e ceaeaees 204 10. 11. 12. 13. 14. ANALYTICAL CHEMISTRY OF REACTOR MATERIALS ..ottt ene e 207 Detection of Traces of NAK in Al et cr e et e e sae e e e 207 Detection of Microgram QUantities ........oociciiiiiiiiiiicciiece et et et s 207 Detection of Submicrogram QUAantities .....cccoiciieriieiiieiieeitcie et et 209 Spectrophotometric Determination of Titanium in Mixtures of Fluoride Salts with Tiron................ 210 Determination of Tantalum in Fused Mixtures of Fluoride Salts ...ccooveviieieiieccce 211 Determination of Oxygen in ZrF 4 by Bromination .......coeovmmieieniisciiiiicstr s 212 Determination of Micro Amounts of Boron in Fused Fluoride Salt Mixtures......c.ccccoeveccvivnmrineeene. 212 Spectrophotometric Determination of Bismuth in Fused Mixtures of Fluoride Salts ......ccoceenvennnnn. 213 Determination of Dissolved Oxygen in Lubricating Fluids....cccocoovveviriiriireece e, 213 Determination of Zirconium by the Compleximetric-Versene Method.........cccccocvierevieriniceeeiieceien. 214 Determination of Rare-Earth Elements in Stainless Steels and Inconel ....oooveviiniencicic, 214 Determination of Oxygen in Metallic Sodium ..ot s e 214 ANP Service Laboratory ...ttt sttt s a e 215 RECOVERY AND REPROCESSING OF REACTOR FUEL ..o 216 Pilot-Plant Design and Construction ......cciiereieeicicieie sttt esissss et s ns s smses s sesesesesssessssssnns 216 Engineering Developments .. ...t e 216 OMEACTOR .ttt ettt ettt e ettt e e be et e e s eaa b et ettt et esneae et s nenters e st ere et et et sesanreneseees 216 Freeze Valves .t ee saeee sae bbb ebb e b e eb et e beeseae e aeen s 216 Process Development ...ttt bbbttt e erae s 216 FIUOrination SUIes ... ee et bt et eb bbb e ee e sens 216 Vapor Pressure of the UF (-NaF Complex .......ccoiiiiimieiic st 219 Uranium Losses on Desorption of UF‘S From NaF oo eeene 219 PART Ill. SHIELDING RESEARCH SHIELDING ANA LY SIS ettt e e sttt et saea b sne bt st saesbe e 223 Energy Absorption Resulting from Gamma Radiation Incident on a Multiregion Shield With SIab GeOMEIIY ...ceeieie ettt e n bt st resanarebmesasebeaee s 223 REACTOR SHIELD DESIGN ..ottt ettt ettt s et s st b ase st e sesmnanesas s s s e e e enens 227 Gamma-Ray Heating in a 300-Mw Circulating Fuel Reactor ........ccoevveeicennivieceicicecescee e 227 Primary Gamma-Ray Heating ..ottt ettt s s ne st s ne s 228 Fission-Product Gamma-Ray Heating .....ccccvvimriiriiierseee et st sev e essean 233 Heating by Thermal-Neutron Captures in the Shield ... 234 Dose Rate Outside the ART Shield ..ottt eee st s s 236 LID TANK SHEELDING FACILITY oo iee e saeseesresasne s sersessasssasssss srsensssesessnnsseanens 237 Analysis of the Dynamic Source Tests on Mockups of the Reflector-Moderated Reactor oTaTe 25T T 1 I O TSROSO 237 BULK SHIELDING FACILITY ettt e saeb et s e 249 Gamma-Ray Streaming Through the NaK Pipes That Penetrate the ART Shield ... 249 Decay of Fission-Product Gamma Radiation.........c.ccocoiiiiciiiiciicii e 250 ; T vas xiii ANP PROJECT QUARTERLY PROGRESS REPORT SUMMARY PART |. REACTOR THEORY, COMPONENT DEVELOPMENT, AND CONSTRUCTION 1. Reflector-Moderated Reactor Construction work on the building additions, building alterations, and cell installation (package 1) for the Aircraft Reactor Test facility was on schedule and at the 50% completion point at the end of the quarter. A contract for $58,400 was awarded to Rentenbach Engineering Company, Knoxville, Tennessee, for package 2 construction work, which consists in the installation of the diesel-generators and facility, the electrical confrol centers, and the spectrometer-room elec- trical and air-conditioning equipment, Design was completed and contract negotiations are under way for package A work, which consists in auxiliary-services piping. Design work on package 3, which includes process equipment, process piping, etc., is currently under way, Detailed layouts have been completed on all major subassemblies of the ART, Typical weld joints are being fabricated to determine shrinkage allowances for the final weld designs. Also, extensive calculations of stress distributions in major structural components and in piping are under way, Recent calculations have indicated that the heat flux through the core sheils of the ART will be higher than previously estimated. The design conditions for the reflector-moderator cooling system were therefore modified accordingly. The design sodium-temperature rise was increased from 150 to 200°F and the design sodium-system operating-temperature range became 1050 to 1250°F to take the estimated heat load of 6.2 Mw. Minor changes were also made in the dimensions of the circuit, to assure that the pressure drop will not exceed 30 psi, and in the arrangement of the auxiliary radiators, to accommodate the increased heat load. The thickness of the lead portion of the reactor shield was reduced from 7 to 4.3 in. to obtain space for increased neutron shielding around the pumps, This decrease in lead thickness will in- crease the gamma-ray dose rate by a factor of 10 {from 7/8 r/hr at 50 ft to ~9 r/hr), but the dose rate will be substantially less than that proposed by most of the airframe companies for nuclear aircraft. A detailed one-sixth-scale model of the facility is being constructed to assist in layout, fabrication, and installation work., Also, the one-halfUF4+K° neither in NaF-KF-LiF eutectic indicated that UF, is quite unstable, possibly because of complexing of UF, by fluoride ions. This produces not only unexpected quantities of K® but also g high activity of uranium metal to satisfy the 4UF , —> 3UF, + U disproportionation reaction, Many special structural-metal fluorides were prepared for experimental studies, including CrF,, Fer, CeF,, LaF,, complexes of ZrF4 and metal fluorides, CuF,, NaCrF,, and NaNiF,. Color studies of alkali metals in equilibrium with alkali halides at high temperatures are under way. In general, alkali-metal phases in contact with alkali-metal salts are highly and strikingly colored, In preparation for thermodynamic studies of molten salts, vapor-pressure, density, and surface- tension measurements were made, Vapor-pressure equations for ZrF, obtained by various investi. gators were correlated, and measurements of the system KF-ZrF, were made., The vapor pressure of pure F:e(:l2 was measured in preparation for a study of ihe activity of FeC[2 in melts of alkali- metal chlorides, are being studied by the sessile-drop technique. Apparatus added to the conventional equipment for fuel purification is being used for the study of variations of density and surface tension of molten salts with changes in chemical composition, It was established that, in the solid, CaCl, forms 1:1 complexes with KCl, RbCl, and CsCI. Density and preliminary conductance data were obtained for several molten rare«earth halides, and fused alkaline-earth halides and their mixtures with some alkali halides are being investigated. Experimental studies were made with the copper- lined stainless steel reactor cans proposed for Surface tensions of molten salts use in large-scale fuel production to replace the unsatisfactory nickel reactor cans, It appears that the copper-lined stainless steel reactor cans will be satisfactory with ZrF ,-bearing mixtures. Tests with alkali-metal and BeF ,-bearing materials are yet to be made. The equipment for conversion of ZrCl, to ZtF, by direct hydroflucrination of the solid is being rebuilt to overcome mechanical difficulties, and an alternate process for the conversion is being studied on a small scale, Experiments carried out with the fuel-enrichment apparatus used during the high-temperature critical experiment on the ART showed that the equipment will be satisfactory if usage is restricted to addi- tions greater than about 500 g. 5. Corrosion Research The effects on corrosion of varying the ratio of hot-leg surface area to volume of fluoride fuel mixture in a forced-circulation Inconel loop were investigated. In a loop in which the ratio was decreased by a factor of 2 in comparison with the PERIOD ENDING MARCH 10, 1956 ratio for a standard loop, the maximum depth of attack in 1000 hr was 6 mils, which is comparable to the usual attack in a standard foop. The addi- tion of sufficient fuel mixture to decrease the ratio by a factor of 4 resulted in attack to a depth of 9 mils in 1000 hr, Further, a very thin gald- colored deposit was found in the cooler portions of the loop. The significance of the increase in depth of attack with increased volume cannot be judged until the source of the deposit is found, but, even if the increase in depth of attack can be attributed to the increase in volume, it appears that wvarying the system volume in the range studied does not have a significant effect on corrosive attack over a 1000-hr period. Examinations of two forced-circulation loops that had operated with the fuel mixture NaF-KF- LiF-(UF, + UF,) clearly demonstrated that the presence of U3* in such a mixture is very effective in reducing hot-leg attack. Continuous layers of metallic uranium, formed by the dispropertionation of UF3, were found, however, along the walls in the cooled sections of the loops. The maximum depth of attack in 1000 hr was 2 mils, in contrast to attack to a depth of 35 mils in a loop operated under similar conditions but which circulated an alkali-metal fluoride mixture with present only as UF the uranium Two Inconel forced-circulation loops in which NaK was circulated completed 1000 hr of operation with a temperature gradient of 300°F and a maxi- mum fluid temperature of 1500°F, These loops differed only in that one included a bypass cold trap. As in the case of Inconel-sodium loops, metallic deposits were found in the economizers and in the cooled sections of the loops. Analyses of the deposits showed them to be identical to those found in Inconel-sodium loops., The simi- larity of the results for the two loops indicated that the cold trap was not effective in reducing mass transfer. A stainless steel loop operated under the same conditions with sodium showed only very slight mass transfer in the economizer and in the cold leg. Examinations of two Hastelloy B thermal- convection loops operated for 500 hr with an alkali- metal fluoride mixture containing UF4 at a maximum temperature of 1500°F showed the maximum attack to be less than 2 mils. The attack appeared as heavy surface pitting, A similar loop that was operated for 2000 hr to determine the effect of time ANP PROJECT PROGRESS REPORT of operation showed no increase in depth of attack, but the surface pits were more concentrated. Thermal-convection loops constructed of Hastel- loy B tubing that had been reamed to ensure uni- formity of the inner surface were operated with sodium and with NaF-ZrF ,-UF , for 500, 1000, and 1500 hr to determine the effect of increasing the operating time. The depths of attack were similar to those found in loops constructed of as-received tubing, and the variations in aftack as a function of operating temperature were slight, Screening tests of special fuel mixtures are under way. In the preliminary Inconel thermal-convection loop tests, it was demonstrated that NaF.LiF- ZrF ,-UF , mixtures show corrosion behavior typical of alkaliemetal fluoride mixtures rather than zirconium-base mixtures. Six standard thermal-convection loops were operated as the first of a series of loops to study the effect on corrosion of various alkali-metal fluorides as components of the basic fluoride fuel mixture MF-ZrF ,.-UF, (50-46-4 mole %), where M stands for potassium, rubidium, or lithium. The attack with the lithium-bearing mixture was severe, but the attack with the potassium and rubidium mixtures was similar to that normally found with NaF-ZrF 4-UF“. Inconel tube-to-header joints brazed with nickel- chromium-germanium-silicon low-cross-section al- loy s were tested in sodium, NaK, and NaF-ZrF ,.UF in seesaw apparatus. Preliminary data indicate that alloys with high nickel contents are the most corrosion-resistant to NaK. In a study of the effect of brazing time on corrosion, the depths of attack on joints brazed slowly (4 hr) and rapidly (10 min) were found to be the same. Cathalloy A31 {4 wt % W-96 wt % Ni) was found to have good corrosion resistance to static sodium at 1500°F. One specimen of Sm203 and two specimens of a commercial rare-earth oxide mixture (63.8 wt % Sm,0;-26.9 wt % Gd,0;—balance primarily other rare-earth oxides) were tested in static sodium in Inconel containers at 1500°F, One of the com- mercial specimens was tested for 500 hr, and the other commercial specimen and the Sm O, speci- men were tested for 1000 hr. All the specimens changed color; however, powder x-ray diffraction comparisons of the untested and tested specimens did not show any reaction products, Two type 316 stainless steel thermal-convection loops were operated with sodium to study the effect of a diffusion cold trap on the amount of corrosion and mass transfer observed in such a system. There was less attack and fewer mass-transferred crystals in the loop which had the cold trap. Three type 348 stainless steel-boiling-sodium loops have been operated for various lengths of time to study the extent of mass transfer in a stainless steel—sodium system in which the oxygen content of the sodium is held to a very low level. No mass-transfer crystals were detected in the cold sections of these systems, Tests have been conducted at 1200°F for 1000 hr to study the extent of dissimilar-metal mass trans- fer of beryllium metal to Hastelloy B in contact with sodium as a function of the distance between the two materials. The results indicated that to avoid extensive alloying of beryllium with Hastel- loy B at 1200°F a minimum separation distance of 20 mils is required, A solid-phase-bonding screening test of K151A (70% TiC-10% NbTaTiC,-20% Ni) against K152B (64% TiC—6% NbTaTiC,-30% Ni) showed that these cermets solid-phase bond when in line con- tact at a pressure of approximately 280 Ib per linear inch in NaF-ZrF ,.UF, (53.5-40-6.5 mole %) at 1500°F for 1000 hr. A series of differential-thermal-analysis tests were performed to determine the solubility of lithium in NaK. The data obtained showed that the solubility is very low, 0.25 wt % Li being soluble at 75°C, Nine nickel- and iron-base alloys of special compositions were prepared and were subjected to static corrosion tests in sodium hydroxide for 100 hr at 1500°F. The most promising alloy of the group was a 90% Ni-10% Mo alloy, which was attacked to a depth of less than 0.5 mil. Further tests are to be conducted on specimens of this composition, Thermal-convection loop tests of fused sodium hydroxide in promising container metals are also under way. The preliminary tests with Inconel loops showed some discoloration of the hot zones and etching in the cold legs. There were no massive deposits in any of the loops. Preliminary tests of nickel are under way. Screening tests of elastomers for possible use as valve seat materials initiated, in NaK circuits were and several promising materials were selected for further testing. Three General Electric Company silicone specimens remained intact after exposure to NaK at 200°C for three days under a helium atmosphere, A search for a moderator that can be cooled by direct contact with the molten fivoride fuel mix- ture was initiated, Thus far, beryllium carbide, beryllium oxide, and boron carbide have been tested in NaF-ZrF .UF, in Inconel capsules at 800°C for 24 hr. The beryllium carbide specimen reacted completely with the fuel melt; the beryllium oxide specimen showed a 2% change in dimensions and a 6% gain in weight; and the boron carbide specimen showed a slight loss in weight. Experiments are under way in which activation analyses are being used to study the diffusion of chromium in Inconel. It is believed that succes- sively milling small amounts of metal from the inside of an irradiated tube will produce a satis- factary series of samples from increasing depths, The amount of the sample, the depth of each milling, and « measurement of the amount of Cr51 (278-day) radioactivity in each portion will be used to de- termine the degree of chromium diffusion as a result of corrosive attack by molten fluorides. 6. Metaliurgy and Ceramics Metallographic investigations are under way on NaK-to-air radiators that failed after various periods of service at high temperatures. A corre- lation of the degree of adherence of the braze material at the tube~to-fin joints and the degree of oxidation of the copper fins with heat transfer performance is expected to give a better under- standing of the relative importance of the fabri- cation features. An experiment was conducted to determine the influence of the stresses set up during welding on the crack susceptibility of back- brazed tube-to-header joints., Two core halves were brazed, one in the conventional upright position and one in the horizontal position, and dye-penetrant inspection indicated freedom from cracks both before and after welding. Tests of transverse shrinkage of welds of the type to be used in the fabrication of the ART were made. [t was found that weld shrinkage increased with plate thickness increases and that the Heliarc welding process resulted in a larger shrinkage for a given joint design in a plate of given thickness than did the metallic-arc welding process. PERIOD ENDING MARCH 10, 1956 Methods for controlling, during fabrication, the critical spacing between the two volutes which constitute the pump housing of the ART NaK pump are being studied, A combination of welding and annealing was developed that gave satisfactory spacing, and a brazing procedure is being in- vestigated. An experimental program is under way for de- veloping a reliable and consistent procedure for joining cermet valve components to metallic structural materials, such as Inconel. A highly promising procedure has been found in which an interfacial reaction between the cermet and the nickel transition layer at 1350°C results in a metallurgical bond. The nickel layer is then brazed to the Inconel component, Flow=-point determinations were made on several low-cross-section brazing alloys being considered for use in the fabrication of the ART. Four alloys with satisfactorily low flow points were submitted for corrosion festing. Satisfactory brazing of aluminum bronze fins to Inconel was obtained by wusing a thin (0.002-in.) electroplate of iron to facilitate welding, The addition of manganese to nickel-base high- tubes temperature alloys was also found to promote melting, The creep-rupture testing program for Inconel was nearly completed, The design curves for as- received (fine-grained) and annealed (coarse- grained) Inconel in argon and in the fuel mixture NaF-ZrF -UF, (50-46-4 mole %) at 1300, 1500, and 1650°F were completed. The effect of a biaxial-stress system, such as is found in pressurized tubing, on the ductility of Inconel was evaluated by comparing the ductility obtained in a tube-burst type of test to that ob- tained for uniaxially-stressed sheet-type speci- mens. The decrease in ductility shown by the tube-burst specimen may be attributed to the 2-to-] hoop-to-axial stress ratio set up in a closed-end pressurized tube, Metal-bonded boron bodies are being studied as possible alternates for hot-pressed B ,C as neutron shielding material because of uncertainties with regard to the effects of irradiation on hot-pressed B,C. The densities obtained have varied from about 0.30 to 1.5 g of boron per cubic centimeter of material. Metal matrices of copper, iron, and molybdenum were investigated. Rapid deteriora- tion of the metallic properties was found to occur ANP PROJECT PROGRESS REPORT as the boron content was increased. The most promising cermet was the copper-B,C mixture, A design scheme for the ART neutron shield was devised which will permit the radiation damage to be absorbed in a ductile matrix of copper and, also, permit the use of B C ceramic tiles to increase the boron density and, hence, reduce the NaK activation. This was achieved through the use of a layer of copper-B4C backed by canned B ,C ceramic tiles. The reactions which occur between Inconel and metal borides, oxides, nitrides, carbides, and carbonitrides were determined at temperatures be- tween 1100 and 2000°F. The reactions between lnconel and boron-containing materials would prohibit their use in contact with each other above 1500°F, Fabricability studies were carried out with fair results on heats of the nominal composition of Hastelloy B with carbon additions of up to 0.1%. The carbon additions were effective in reducing oxide-type inclusions, and the ingots were hot- rolled to sheet successfully., A similar series of melts of Cr-Mo-Ni compositions with carbon addi- tions is being prepared in an effort to improve fabricability of these alloys. Metallographic ex- amination of a cracked extrusion of Hastelloy W revealed eutectic melting at the base of a crack. Identification of the eutectic was not possible, Room-temperature tensile tests of pack-rolled clad-niobium sheet have been completed. The results compare favorably with those obtained with wrought stock. Studies were continued of diffusion barriers for cladding of niobium with Inconel. At present the duplex barrier of copper—stainless steel looks most promising. A technique has been developed by which completely clad creep speci- mens of niobium will be prepared and tested. Extensive compatibility tests of UO, and niobium, both powdered and wrought, were carried out by heating for 500 hr at 1000°C, All evidence indicates that the two materials are compatible, Niobium is being considered for use in solid fuel elements because of its good high-temperature strength, Encouraging results were obtained on the ex- trusion of simulated seamless tubular fuel ele- ments with mixtures of Al,O, and stainless steel as cores. lhree-ply extrusions at ratios of 9:1 and 21:1 showed fairly uniform cladding and core thicknesses. Sections have been sent to the Superior Tube Company for redrawing. A similar technique is to be used for the extrusion of ART control rods. Work is continuing on the development of a high- density barrier shield plug for the ART pumps. Fabrication procedures have been established for several suitable materials. A thermal conductivity apparatus has been constructed and tested for conductivity measurements on the possible ma- terials, A study of the fabricability of a light alloy con- taining 20% lithium, which would be useful as shielding material, was initiated. Several pro- tective coatings and cladding materials for a lithium-magnesium alloy are being investigated. Methods are being studied for preparing compacts of samarium and gadolinium oxides and cermets of iron and rare-earth oxides and nickel and rare- earth oxides for use in reactor control elements. The cyclograph flow-detection instrument is being used successfully for the inspection of small-digmeter tubing. The instrument is essen- tially a tuned oscillator in which the oscillator coil encircles the tubing, The variables that cause flow signals have been studied, and spurious signals have been defined. The wobble of the tubing in the coil, which was the worst offender in producing spurious flow indications, can be eliminated by a well-designed mechanical feed : mechanism, Optimum test frequencies have been determined as functions of tubing diameter and wall thickness, The results obtained from immersed ultrasound, radiography, and fluorescent penetrant inspections were compared. It appears that no one of the inspection methods alone is adequate and that if any one of them were eliminated a few defects would be overlooked. 7. Heat Transfer and Physical Properties Preliminary heat-transfer data were obtained for molten NaF-KF-LiF-UF, (11.2-41-45.3-2,5 mole %) flowing through Hastelloy B tubes in a pressurized forced-convection system. The data fall about 30% below the normal correlation for ordinary fluids. Heat-transfer and isothermal friction char- acteristics were determined in the transition re- gion for three degrees of spacer density for the ART heat exchanger. The limited data obtained thus far appear to indicate that for a given heat exchanger pressure drop the heat transfer coeffi- cient is almost independent of the spacer density. A 5/22-scale model of the ART sodium coolant annulus was fabricated and assembled for a study of the fluid flow distribution. Pressure-drop and flow-distribution measurements are being obtained for concentric-annulus conditions, for various radial displacements of the outer core shell, and for other conditions, ‘ Yisual studies were made of the flow through the 21-in. ART core model after the installation of vortex generators supplied by Pratt & Whitney Aircraft, Yane angles of 50, 55, 60, 65, and 70 deg were used. Although no large reversed flow regions existed, steady and unsteady flow were observed in various regions of the core., QOther means to obtain better flow distribution are being studied., The temperature structures in an idealized ART core and in the region beyond the beryilium re- flector were calculated. Upper and lower limits for the temperature rise of the sodium flow over the outer surface of the beryllium reflector were 25 and 33°F. The core-wall cooling by the sodium in the core annuli was found to be approximately 2.5 Mw. The half-scale model of the ART core to be used for volume-heat-source heat-transfer ex- periments was completed and was operated with water. Operating procedutes and schedules are being established, A study of the effect of composition on heat capacity for NaF-ZrF , salt mixtures was investi- gated; no significant differences were found over the composition range investigated, 53 to 65 mole % NaF and 35 to 47 mole % ZrF ,. A study has been initiated to determine the influence of fission products on the viscosities of fuel mix- tures. The viscosities of NaF-ZrF , (50-50 mole %), NaF-LiF-BeF, (49-36-15 mole %) and NaF-.LiF- BeF ,-UF, (56-21-20-3 mole %) were determined, Performance comparisons of several of the more important ANP fuels were made for heat and momentum transfer in the reactor core and in the heat exchanger. The alkali-metal base fuels were found to be superior to the zirconium-base fuels. The major differences between the fuels occur in the heat exchangers. The heat exchanger pressure drops were found to be only about half as great for the alkali-metal base fuels as for the zirconium-base fuels, and the radial fuel tempera- ture differences were from 20 to 40% lower. 8. Radiation Damage The effects of irradiation on the corrosion of Inconel exposed to a fluoride fuel mixture and on PERIOD ENDING MARCH 10, 1956 the physical and chemical stability of the fuel mixture have been investigated by irradiating Inconel capsules filled with static fuel in the MTR and by operating in-pile forced-circulation Inconel loops in the LITR and the MTR. In the many capsule tests and in the three in-pile loop tests made to date, no major changes that can be at- tributed to irradiation, other than the normal burnup of the uranium, have occurred in the fuel mixtures, The metallurgical examinations of the lnconel capsules and tubing have likewise shown no change in corrosion that can be the result of radia- tion damage. The low corrosion results obtained for the in-pile loops have been confirmed by chemical analyses for corrosion products in the fuel mixtures. The sectioning of MTR in-pile loop No. 3 was completed at NRTS, and, upon receipt of the sec- tions at ORNL, examinations of the disassembled loop were initiated, Metallographic examination of as-polished Inconel specimens from the nose coil revealed no evidence of corrosive attack. Etched specimens appeared to have moderate inter- granular attack to a depth of about 1 mil, but this may be an etching effect rather than true corrosive attack. Cells are being readied for obtaining specimens from the pump and the remaining lengths of fuel tubing. Chemical analyses of both the fuel and fuel tubing are under way. Provisions for the disassembly of the next loop are being made, Eight Inconel, helium-pressurized, tube-burst, stress-corrosion specimens were expased to radia- tion in hole HB-3 of the LITR in a helium atmos- phere, with a circumferential stress on each speci- men of 2000 psi, Similar specimens are being tested out-of-pile to obtain confrol data. A plot of the time-to-rupture vs the operating temperature during irradiation showed an unusual degree of scatter in the data and led to an investigation of the quality of the tubing. All but four of 28 speci- mens of tubing stock examined by nondestructive inspection were rejected because of pinholes and spongy regions. Metallographic inspection of 25 sections from the 24 rejected tubes, however, resulted in the location of only one appreciable defect. Better quality tubing will be used for future experimentation, An alternate apparatus for stress-corrosion test- ing of structural materials in contact with a fuel mixture has been designed for use in the LITR, The thermal-neutron-flux data needed for the ANP PROJECT PROGRESS REPORT design calculations were obtained by irradiating a mockup capsule at a position about 2 in, from the inner end of hole HB-3 of the LITR. Measure- ments on the mockup indicated that a flux of 9 x 1072 neutrons/cm?:isec will exist at the outer surface of the fuel in the annular stress-corrosion apparatus and that the flux at the inner surface of the annulus will be 5 x 1012 neutrons/cm2sec, Calculations based on these measurements indicate an unperturbed thermal-neutron flux at this position of 2 x 1013 neutrons/cm? sec. Tubular Inconel creep specimens subjected to a tensile stress of 1500 psi at 1500°F in a helium atmosphere and irradiated for 33 days in hole HB-3 of the MTR were found to have fractured. Post- irradiation examination has suggested that localized high temperatures or the unknown contaminant that was responsible for the film and the intergranular attack on the specimen caused premature fracture, Elevated-temperature tensile tests on monel have shown that the existence of a ductility mini- mum at about 1000°F is strain-rate dependent. The ductility minimum was found at a strain rate of 2 in./min, but it was not observable at a strain rate of 0.002 in./min, Seven Inconel capsules containing NaF-ZrF base fuels that were irradiated in the MTR for periods of six to nine weeks are being sampled. Facilities have been put into service for irradiating two or three capsules simultaneously in a hitherto unused high-flux position in the MTR, The holdup of radickrypton in charceal traps saturated with nitrogen or with helium was investi. gated over the temperature range -110°C to +16°C, A theoretical expression was developed and fitted to the results; this expression may be used for predictions of holdup under other conditions or for optimum design of the charcoal traps. A second vertical in-pile logp is being prepared for insertion in the LITR. The loop is essentially the same as the one operated previously, but the tip of the loop will extend more than 4‘/2 in. below the position of maximum flux in the reactor to obtain a greater temperature differential than that obtained previously, Design conditions for the gamma spectrometer for the off-gas system of the ART were established. A flux-monitoring system in which Inconel is used as the monitor has been developed for use at high temperatures, The Cr3! activity is fol- lowed for exposure times of up to one or two 10 months, and the Co%0 in the Inconel is used for longer exposures. F lux measurements have been obtained from co- balt foils that were attached to the nose coil of MTR in-pile loop No. 3. Fast-flux profile measure- ments were obtained for hole 19 in the ORNL Graphite Reactor, and similar measurements were started in hole HB-3 of the LITR. Although theoretical considerations predict a substantial increase in the corrosion of Inconel by fused fluoride fuels under irradiation, this has never been observed experimentally, The excess oxidizing power of the irradiated fuel, which re- sults from the imbalance of fluorine between UF and the fission-product fluoride mixture, might be expected to increase corrosion. The deposits of the fission-product metals, ruthenium, niobium, and (presumably) molybdenum appear, however, to inhibit attack of such fuel on the container surface. The tritium produced by the Li%(n,¢)He4 reaction is expected to cause a large increase in the cor- rosion of metal capsules by fuels in the NaF-KF- LiF-UF, system. In order to prevent this, any lithium used in MTR capsule irradiations must contain less than 0.2% Li%, In in-pile loops, up to 1% Li% can be used without depressing the neuiron flux unduly and without causing serious increases in corrosion. A study of radiation effects on B4C and related thermal-neutron shielding materials is under way. The first irradiations were made on uncoated hot- pressed B,C (Norbide) and slip-cast B,C bonded with SiC (The Carborundum Company) at low tem- peratures (~200°C)., The hot-pressed samples retained their physical dimensions and bulk struc- ture with no helium evolution. The slip-cast bonded material showed inadequate radiation sta- bility, In order to determine whether the hot- pressed samples will retain helium at elevated temperatures, samples will be irradiated at tem- peratures up to 2150°F. 9. Analytical Chemistry of Reactor Materials Instruments are being developed for detecting leaks of NaK into the exhaust air stream from the NaK-to-air radiators of the ART and from radiators installed in test stands. The instrument to be used on the ART is capable of detecting an incremental increase of about 0.01 ppm of alkali metals in the air while it is crossing the bank of radiators, whereas the instrument for monitoring the exit gases from the test stands is to have a sensitivity of about 10 ppm. The detection method for use with the test-stand radiators is based on the measurement of the hydroxyl ions which are formed when the alkali metal oxides and hydroxides are absorbed from a sample of the contaminated exhaust air in an aqueous solution containing the color indicator bromcresol green. The absorbance of this indicator is increased by a factor of 2 by the addition of 1 to 10 ppm of NaK. The instrument will be designed so that an alarm will be activated if the coulometric current required to hold the pH of the absorber solution at a constant value ex- ceeds a predetermined rate. A more sensitive in- strument for use with the ART radiators is being studied that is based on the absorption of light of the frequency of the sodium doublet resonance line at 5890 A, Calculations show that 0,01 ppm of sodium in air would produce a 50% attenuation in a beam of sodium-doublet radiation which traversed a 50-cm optical path. A spectrophotometric method for the determina- tion of trace amounts of titanium in mixtures of fluoride salts was developed. The method is based on the yellow color of the titanium-Tiron complex, and uranyl, zirconium, and ferric ions are masked to prevent their interference. The method for the determination of trace amounts of tantalum in NaF-KF-LiF mixtures without uranium was modified and was applied to the de- termination of tantalum in uranium-bearing mix- tures. The tantalum was separated from the inter- fering uranyl ions with cupferron, Studies were continued on the bromination method for the determination of combined oxygen in fluo- ride salts. Attempts to improve the recovery of oxygen as CO, from the bromination of ZrO, by the addition of NiF2 were unsuccessful, However, the solid products resulting from the bromination of the mixture of Zr02, NiF,, and graphite were large crystalline masses rather than the finely divided solids obtained previously from the bromination of the mixture of ZrO,, FeF,, and graphite. Analyses of the crystalline product are expected to provide a basis for a better understanding of the reaction mechanism, A method for the determination of small amounts of boron in fluoride salt mixtures was developed in which the salts are dissolved, at room tempera- ture, in a solution of AiCl;+6H,0 and 2 M HCI, The boron remains in solution and is extracted by PERIOD ENDING MARCH 10, 1956 ethyl ether and determined by the carminic acid method. The extraction coefficient was found to be approximately 0.45 + 0.02. By this method, boron can be determined in concentrations as low as 10 ppm. A spectrophotometric method for the determina- tion of bismuth as the tetraiodobismuthate(lll) complex was applied to the determination of bis- muth in NaF-ZrF -UF,. The effect of uranium on the absorbance of the complex was negligible for uranium-to-bismuth weight ratios of less than 30:1. A method for the determination of dissolved oxy- gen in lubricating fluids was developed that is based on the displacement of the dissolved oxygen by carbon dioxide. The oxygen content of a number of lubricants ranged from a maximum of 42 ug per milliliter of fluid to @ minimum of 15 ng/ml, Modifications were made to the method for the compleximetric determination of zirconium in zirconium-base fuels. The zirconium-VYersene com- plex was formed at pH 6, and the solution was digested on the steam bath before tifration, Zir- conium was determined in a total of 21 samples by the modified procedure. The average difference between the results and those obtained by the gravimetric procedure was 0.6%. A significant decrease in the time required for the determination of microgram quantities of ele- ments of the rare-earth group in stainless steels and in Inconel was obtained by removing the chromium, before electrolysis, as the volatile chromyl chloride, The optimum conditions for the distillation of sodium metal from Na,O in the distillation method for the determination of oxygen in metallic sodium were established. The optimum temperature range for distillation is 800 to 850°F. Distillation periods of as much as 4 hr can be tolerated at these temperatures without loss of Na,0. Dis- tillations at higher temperatures confirmed earlier observations that NQZO is volatilized at tempera- tures above 900°F. 10. Recovery and Reprocessing of Reactor Fuel The completion of the pilot plant for recovering fused salt fuels is now scheduled for May 15, 1956. Tests of the percolator type of contactor planned for use in the fluorination vessel showed it to be satisfactory, A freeze valve designed for use in the pilot plant held satisfactorily against a pressure of 20 psig in 50 freezing and melting cycles, T Continued studies of the fluorination process for the recovery of uranium from fused salts have indicated that the presence of salt impurities has a much greater effect on the rate of UF , volatiliza- tion than other factors, such as use of nitrogen with the fluorine or variations of the method of introducing the fluorine into the molten salt. A fluorine efficiency of 70% was achieved with >99% UF , volatilization when the impurity content in the initial salt was kept to a minimum, The vapor pressure of UF , over the complex UF ;:3NaF was determined at various points in the tempera- ture range 80 to 320°C, The data were success- fully fitted by the linear equation log P = 10.88 - (5.09 x 103/1) . mm Hg An enthalpy change of +23.2 kcal per mole of UF was calculated for the reaction 3NaF-UF , —> 3NaF + UF, Uranium losses in NaF beds under process con- ditions did not exceed 0.05%, even with repeated use of the NaF over three cycles, PART IlIl, SHIELDING RESEARCH 11. Shielding Analysis The code of a Monte Carlo calculation of energy penetration and deposition resulting from transport gamma-ray radigtion in a shield of slab geometry was used in a parametric study of a two-region lead-water shieid, The radiation was 1-Mev gamma rays incident on a slab at 0 deg, 60 deg, 70 deg 32 min, and 75 deg 31 min. The first region of the slab was composed of water 1.5 mean free paths (mfp) thick at the initial gamma-ray energy, and the second region was composed of lead 0.5 mfp thick. The coding was also extended to calculate energy flux and tissue dose rate, The results indicate that lead is more effective in stratified slab shields when it is placed behind a good scatterer, such as water. 12. Reactor Shield Design The heating in the lead and alkylbenzene shield of a 300-Mw circulating-fuel reflector-moderated was calculated for the following com- (1) primary gamma rays originating in or near the reactor core, (2) fission-product gamma reactor ponents: 12 rays from the heat exchanger, and (3) thermal neutron captures in the lead and borated (2% boron) alkylbenzene. = The third component was sub- divided to take into account the secondary gamma lead, hydrogen, and boron capture. particles resulting from thermal-neutron rays from Alpha captures by boron were also considered. The gamma-ray and fast-neutron dose rates at a distance of 50 ft from the ART were calculated as a function of the thickness of the lead shield. Contributions to the total gamma-ray dose rate by primary gamma rays from the core, secondary gamma rays originating in the shield, and heat exchanger gamma rays were determined, 13. Lid Tank Shielding Facility Analysis of the dynamic source tests on mockups of a reflector-moderated reactor and shield was completed, In the analysis the experimental dose rate resulting from fission-product gamma rays emitted in the heat exchanger mockup was sepa- rated out and compared with dose rates calculated by two different methods. In one calculation it was assumed that all the fission-product gamma rays were of a single energy (2.7 Mev/photon). The other calculation was based on the spectrum of gamma rays from the fuel belt, previously meas- ured at the LLTSF. The two calculated values agreed, but they differed by about 30% from the measured value, |t was found that the difference could be attributed to the dose buildup factor for water having been used in the calculation (that is, the buildup factor was chosen as if the lead were an equivalent thickness, in mean free paths, of water), Substitution of a new buildup factor for the total mean free paths (lead and water), based on Monte Carlo studies of laminated shields, re- sulted in agreement between the measured and calculated dose rates, 14. Bulk Shielding Facility A mockup experiment at the ORNL Graphite Reactor thermal column was initiated to measure the gamma rays streaming through the NaK pipes of the ART, The measurements of the decay rates of fission- product gamma rays as a function of time after fission were extended to include six energy groups between 0.28 and 5 Mev for times up to 1600 sec after fission. Part | REACTOR THEORY, COMPONENT DEVELOPMENT, AND CONSTRUCTION 1. REFLECTOR-MODERATED REACTOR W. F. Boudreau A. P. Fraas H. W. Savage Aircraft Reactor Engineering Division A. M, Perry Electronuclear Research Division E. R. Mann Instrumentation and Controls Division ART FACILITY DESIGN AND CONSTRUCTION F. R. McQuilkin Aircraft Reactor Engineering Division Design and construction work on the Aircraft Reactor Test (ART) facility in Building 7503 was continued. Package 1 construction work on the building additions, building alterations, and cell installation was on schedule and at the 50% com- pletion point. The principal work accomplished during the quarter included completion of the building-addition structure, completion of the spec- trometer room and tunnel, erection of the 30-ton crane, modification of the 10-ton crane, con- struction of the blower house and switch house, placement of reinforced concrete for the stack base and for the walls of the special-equipment room, and partial erection of the cell tanks and surrounding structure, Some of this work can be seen in Fig. 1.1, which is a view from a location southwest of the altered 7503 building, The dark-sided portion of the main building is the addition which will house the reactor cell, heat- dump systems, and spectrometer room and tunnel, In the foreground, from left to right, can be seen the switch house, blower house, and stack foun- dation, A closeup view of the blower-house interior, prior to erection of the exterior walls, is shown in Fig. 1.2. In the left foreground can be seen the walls of the ramp access way from an elevation of 840 ft to the radiator pit at an elevation of 820 ft in the main building. In the center foreground are the forms and the reinforcing steel for the upstream end of the main air duct. An interior view across the building addition looking toward the southeast corner of the building is shown in Fig. 1.3. In the bottom center can be seen the spectrometer tubes, which extend to the spectrometer tunnel beneath the low bay and which will penetrate the cell water tank. At the bottom right can be seen portions of the special- equipment-room walls, An interior view across the building addition [ooking toward the southwest comer is shown in Fig. 1.4. The lower sections of the cell tanks can be seen in the foreground, and the opening into the blower house for the main heat-dump air duct is shown in the center left, The duct will enter the building and pass hori- zontally across the corner to the stack, The main and auxiliary radiatoers will be mounted in the duct on a diagonal line approximately as repre- sented by the left edge of the picture. The bottom sections of the cell tanks are shown in Fig. 1.5. The inner tank is the bottom hemispherical head of the pressure vessel, and the outer tank is the water tank. The incomplete cylindrical portion of the pressure vessel is shown in Fig. 1.6. At the top center can be seen the 4-in.-thick panel which contains sleeves for the NaK and the off-gas piping. In the left and right foreground are the 4-in ~thick panels which will receive the junction panels for auxiliary-services lines, heater leads, and the instrumentation and control lines. The large 2-in. panel at the upper left contains the small nozzles for beam penetrations into the spec- trometer tubes. In the bottom of the tank, under the scaffolding, can be seen the structural framing for the floor, equipment support, and reactor support, Bids were received for package 2 construction work, and a contract for $58,400 was awarded on January 20, 1956, to the Rentenbach Engineering Company, Knoxville, Tennessee. This unit of work consists in the installation of the diese! generators and facility, the electrical control centers, and the spectrometer-room electrical and air-conditioning equipment. At the end of the quarter, this work was in the material procurement stage, with no site work having been started. Approximate completion dates for package 2 are 15 91 Fig. 1.1. View of Altered 7503 Building from Southwest. Photograph taken Feb. 17, 1956. UNCLASSIFIED PHOTO 16503 LA0dIY SSTYO0Yd LDIF0¥d dNY Ll Fig. 1.2. Interior of Blower House Prior to Erection of Exterior Walls, Photograph taken Feb, 10, 1956. "UMCL ASSIFIED 9661 ‘0L HOYVYW ONIAGN3T Q01334 ANP PROJECT PROGRESS REPORT UNCLASSIFIED PHOTO 16456 o o " oy e Fig. 1.3. Interior View Across Building Addition Looking Toward Southeast. Photograph taken Feb. 10, 1956. | 18 PERIOD ENDING MARCH 10, 1956 . UNCLASSIFIED 'PHOTO 16455 Fig. 1.4, Interior View Across Building Addition Looking Toward Southwest. Photograph taken Feb. 10, 1956. 19 ANP PROJECT PROGRESS REPORT £$991 0LOHd AF4ISSYIONN 9661 .m 'qa4 uajp} _._aEmoho_._n_ ‘SHuUD) ||oD) j0 suoldeg wojjog S l nm_m 20 PERIOD ENDING MARCH 10, 1956 UNCLASSIFIED § PHOTO 16506 Fig. 1.6. Incomplete Cylindrical Portion of Pressure VYessel, Photograph taken Feb. 17, 1956. June 24, 1956, and July 15, 1956, respectively, for installation of contractor-supplied and Govern- ment-supplied items, Design was completed and negotiations for a proposal are in progress with the package 1 contractor, V. L. Nicholson Company, for the installation of package A work, which consists in auxiliary-services piping. Two proposals are being requested, one based on the package 1 completion schedule, which currently is June 8, 1956, and the other based on the contractor’'s proposed schedule in the event the first schedule presents unreasonable situations, The proposals are to be submitted March 7, 1956. Design work on package 3, which includes the process equipment, the process piping, etc., is currently under way, and it is expected to be complete early in June 1956. Package 3 work will be performed by ORNL forces, ART DESIGN A. P. Fraas Aircraft Reactor Engineering Division H. C. Gray Pratt & Whitney Aircraft Detailed layouts have been completed on all major subassemblies of the ART., Detailing of the north head and the main heat exchangers has been completed, and the detailing of the control rod, reflector-moderator, island, and pressure-shell subassemblies and the reactor support structure is under way. Typical weld joints are being fabricated to determine weld shrinkage allowances (see Sec. 6, ‘‘Metallurgy and Ceramics'') for the final weld designs. Detailed layouts of the reactor shield and the piping inside the reactor cell are being prepared. 21 ANP PROJECT PROGRESS REPORT Stress Analyses R. V. Meghreblian Aircraft Reactor Engineering Division An extensive series of calculations has been under way for several months in order to determine the stress distribution in major structural com- ponents at the top of the reactor assembly, referred to here as the ‘‘north head.’”” The design loads being used in these calculations are based on full-power {or ‘‘design-point'’) operation of the reactor. Two components have been studied in detail: the double-ring structure, which transmits the up-load from the reflector-moderator to the north head (and thence to the pressure shell), and the composite double-deck structure of the north head. The proposed design of these members is based on these calculations. It is believed that these analyses provide a reasonably accurate picture of the gross features of the stress distribution and, therefore, are adequate for determining basic dimensions and proportions of structural members. However, the structural configuration of the north head is so complex that analytical studies must be based on relatively simple geometric models which cannot include all the details of the actual design. Thus the finer details of the stress distribution must be determined by other techniques. experimental A program of stress analyses was prepared for this purpose, and the work is presently under way at the University of Tennessee. The program includes stress studies of the two north-head components mentioned above, as well as studies of the structural integrity of the pump-barrel and NaK pipe attachments to the pressure shell and full-scale tests on portions of the NaK piping outside the pressure cell. The NaK circuit selected for this experiment is that portion of the system which serves the NaK pump-and-radiator assembly on the reactor side of the main air duct (Fig. 1.7). This circuit is the shortest portion of the entire piping system, and it is believed to be the least flexible; thus stresses from thermal expansion of the piping are expected to be most severe in this circuit. At the conclusion of the thermal-stress study, the facility will be used to determine the vibrational characteristics of the complete NaK pump-radiator-piping assembly. The program of detailed stress analysis of the ART is being supplemented by a parallel program 22 ORNL-LR-DWG 13200 aft s¥in, 7 ft Qin, A e | RADIATORS - /, e Fig. 1.7. Assembly on Reactor Side of Main Air Duct. Nak Circuit for Radiator-and-Pump of material tests on the high-temperature character- istics of Inconel and beryllium. The purpose of these tests is to establish a better basis than that presently available for the design analysis of structural components that are exposed to cyclic load conditions. A research contract has been awarded to the University of Alabama for a program of tests on the high-temperature strain and thermal-cycling properties of Inconel. Con- tracts are being negotiated with the Universities of Michigan and Syracuse for a series of tests to determine the relaxation and creep properties of Inconel in the temperature range 1200 to 1600°F. The OSyracuse program is to include a joint analytical-experimental study of the creep-buckling of thin hemispherical shells. The data to be obtained through these contracts are needed for the design analysis of the core and reflector shells, which will contain the circulating-fuel system. An analytical and experimental study of the reflector shells is being undertaken by the Lewis Laboratory of the National Advisory Com- mittee for Aeronautics. Sodium System R. I. Gray Aircraft Reactor Engineering Division Recent calculations of the effects of the non- uniform power distribution in the reactor core indicate that the heat flux through the core shells of the ART will be higher than previously ex- pected. The combined effects of heat transfer and radiation heating yield a total estimated heat load to the sodium system of 6.2 Mw. The increase in the heat load required either an increase in the design sodium flow rate or an increase in the design temperature rise of the sodium system, Since an increased sodium flow rate would result in excessive pressure drops in the system, the sodium temperature rise was increased from 150 to 200°F, and the sodium-system operating temper- ature range became 1050 to 1250°F. Minor changes in the dimensions of the reflector-moderator cooling circuit were also made to stay within the pressure- drop limitation of 30 psi and the maximum sodium temperature limitation of 1250°F. These changes included the removal of the staple-spacer support lands of the island and the reflector annuli, the increase of the island and the reflector-annuli thicknesses to 0.188 in., the shortening of the reflector-return-annulus spiral spacers to 3.1 ft (6, = 45 deg), and the increase of the reflector- retum-annulus thickness to 0.125 in. The flow of sodium in the island and reflector cooling holes is to be adjusted by variation of the entrance conditions of each tube row., In this manner a balanced flow condition can be obtained with a minimum pressure drop. Detailed relaxation calculations have determined an island cooling-hole distribution which limits the maximum beryllium temperature difference to 50°F, while maintaining @ minimum number of cooling holes. This design includes a total of 120 holes. The previously determined analog solution for the reflector cooling-hole distribution PERIOD ENDING MARCH 10, 1956 specified 288 holes, and that solution remains unchanged. The increased sodium-system heat load alse necessitated modification of the auxiliary radi- ators. The four radiators are to be mounted to form a V in order to obtain low NaK and air pressure drops. Reactor Shield W. B. Cottrell Aircraft Reactor Engineering Division The thickness of the lead portion of the ART reactor shield was reduced from 7 to 4.3 in, to obtain space for increased neutron shielding around the pumps. This decrease in lead thickness will increase the gamma-ray dose rate by a factor of 10 (from 7/8 r/hr at 50 ft to ~ 9 r/hr), but the dose rate will still be substantially less than that proposed by most of the airframe companies for nuclear aircraft. Collars of gamma-ray shielding material are being designed to relieve the problem of gamma-ray streaming through the various ducts in the lead shield. Reactor Study Models W. L. Scott Aircraft Reactor Engineering Division A detailed one-sixth-scale model of the ART is being fabricated to assist in the layout of piping and in the study of fabrication and installation procedures. Tentative mockups of the oil, helium, electrical, and instrumentation lines are now being made. All the basic cell structure, the reactor- and-shield support structure, and the NaK lines have been installed. The existing one-half-scale plastic model of the north head is being modified to reflect the current design. This model is to be used for the determi- nation of the locations of the pressure taps required for flow studies and for layout design of the shielding of the north head. ART COMPONENT DEVELOPMENT, PROCUREMENT, AND TESTING Water Test of Aluminum Mockup of North Head | D. R. Ward Aircraft Reactor Engineering Division The fabrication and assembly of the fuel-system components of the aluminum north-head mockups are nearly completed, and tests of the fuel system 23 ANP PROJECT PROGRESS REPORT The sodium-te-NaK heat exchanger mockups and other parts of the sodium system will be fabricated while the fuel-system tests are being performed. No unexpected problems are to begin soon. have yet been encountered in the fabrication and assembly of this mockup. Warping of aluminum parts during welding has somewhat reduced the precision of the assembly, and similar warping may be encountered in welding the Inconel reactor parts. Fuel-pump performance, studied under both normal and unbalanced con- ditions. Several view ports have been provided for observation of fluid ingassing and degassing. The interchange of liquid between fuel pumps and the fuel swirl chamber will be studied thoroughly. Special instrumentation is being provided for de- termining the level to which the liquid will rise around a pump shaft when one pump is stopped while the other pump is running. with water, will be Core Flow Studies G. D. Whitman W. J. Stelzman W. T. Furgerson Aircraft Reactor Engineering Division Studies of the flow in a full-scale plastic model of the ART core were continued through the use of the techniques described previously.! Several guide-vane configurations that had been previously tested in the full-scale aluminum core model were retested in the plastic model, and photegraphic, conductivity, and pressure-drop data were obtained for each inlet configuration. Attempts were made to alter the flow patterns obtained, by adding baffle plates and turbulators to the trailing edges of the vanes, but flow-reversal regions still continued to exist in the core annulus gbove the equator, An inlet-guide-vane system designed on the basis of one-quarter-scale-model axial-flow data? provided the lowest core-entrance pressure drop to date, but it generated a flow pattern similar to that obtained with the axial-flow type of header that had no means of auxiliary flow guidance. Further modification and testing of this guide-vane system by means of systematically attaching and relocating a conical baffle plate on the trailing ]G. D. Whitman, W. J. Stelzman, and W. T. Furgerson, ANP Quar. Prog, Rep. Dec. 10, 1955, ORNL-2012; p 23. 2k, E. Lynch et al.,, ANP Quar. Prog, Rep. Dec. 10, 1955, ORNL-2012, p 171. 24 edges of the guide vanes yielded the arrangement shown in Fig. 1.8, which generated a core flow pattern containing no flow reversal throughout the core annulus and, ot the same time, provided good surface scrubbing, good midstream mixing, and fairly high velocities throughout the upper half of the core. Below the equator the velocities were high, but the fluid adjacent to the inner and outer sutfaces tended to hug these surfaces as it approached the core outlet; thus the movement of the fluid from these surfaces was slight, and, in general, the midstream mixing was poor. Further modifications to this design are being made so that the flow in the lower half of the core will be improved. Engineering Test Unit M. Bender W. C. Tunnell G. W, Peach G. D. Whitman Aircraft Reactor Engineering Division Flow diagrams and instrumentation lists have been prepared for the Engineering Test Unit (ETU), which is a nonnuclear mockup of the ART, It will differ from the ART only in heat-input and heat-removal arrangements. For the ETU, two gas furnaces will supply 1.5 Mw of heat to the fuel system through two adjacent NaK systems. The remaining two NaK systems will operate isothermally., The 1.5 Mw of heat input will be removed from the fuel through two sodium-to-NaK reflector-coolant systems. The ART auxiliary systems, such as the fuel drain tank, the off-gas system, the fuel recovery system, the fuel sampling system, the drain-tank cooling system, and the various gas systems, will not be mocked up as part of the ETU. Work has been started on a detailed study of the test program for the ETU, and a manual covering all phases of the program is to be issued. The building modifications required to accommo- date the ETU are being made. Contracts have been let for the core shells, the heat exchangers, and the radigtors. The design of the boron- containing tiles was modified to conform to lower temperature limits and new radiation-damage infor- mation, and therefore procurement of the boron- containing materials has been delayed. Present design and shop schedules indicate that the ETU will be ready for operation in January 1957. 4 PERIOD ENDING MARCH 10, 1956 PHOTO 25581 Fig. 1.8, Guide Corrections for variations of the boron content in the first boron curtain were made by using a plot of sodium activity vs boron layer 3J. B. Dee et al.,, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 209, Table 12,1. 4w, Frank, Nuclear Development Corporation of America (NDA), private communications to H. Bertini. 51, B. Dee, op. cit., p 204, Fig. 12.1. PERIOD ENDING MARCH 10, 1956 thickness® and an extrapolation for the thicknesses expressed as 0.587 and 0.681 g of B'0 per square centimeter, No corrections were made for the variations in heat exchanger thicknesses or for the variations in the second boron curtain thickness. The activity was assumed to be directly propor- tional to the mass of the sodium (in the NaK) and to the design power. An additional correction was made to the LTSF data to account for the incorrect assumption that, at the time the LTSF results were reported, the power of the LTSF source plate was 3.6 w. A recent measurement’ indicates that the source-plate power was more nearly 2.1 w. The delayed-neutron activity calculations by NDA are nof included in Table 1.2, because there is no simple correction factor that will take into account the large heat exchanger thickness used. Only the L TSF-source-power and sodium-mass corrections were applied to the LLTSF delayed- neutron results, and only the sodium-mass correc- tion was applied to the author’s delayed-neutron results, 6G. T. Chapman, J. B. Dee, and H. C. Woodsum, ANP Quar. Prog. Rep. Jume 10, 1955, ORNL-1896, p 199, Fig. 12.14 7D. Otis, private communication to H. Bertini. TABLE 1.1. CONFIGURATIONS USED FOR CALCULATIONS OF THE ACTIVITY OF THE NoK IN THE ART FUEL-TO-NaK HEAT EXCHANGERS Configuration For LTSF For NDA For For or ! Author’s Normalization Calculation Calculation . Calculation Calculation Core radius, in. 10.5 13.19 10.63 10.5 Beryllium-reflector thickness, in. 1 11.81 11.81 1 First boron curtain, grams of g1l per square 0.325* 0.587 0.681 0.216** centimeter Heat exchanger thickness, in. _ 2.1 6.5 1.97 3.25 Second boron curtain, grams of g0 per square 0.325 0.162 0.341 0.216** centimeter Sodium content of NaK in heat exchanger, kg 42 200.67 13.4 35 Reactor design power, Mw 60 60 60 *A ‘,lf'a-in. layer of pure B'0 has 0.731 g of pl0 per square centimeter. el 3/8-in. layer with a natural-boron density of 1.3 g/cm3. 29 ANP PROJECT PROGRESS REPORT TABLE 1,2. RESULTS OF NORMALIZATION TO THE ART CONFIGURATION OF CALCULATIONS OF THE ACTIVITY OF THE NaK IN THE FUEL.TO-NaK RADIATORS OF CIRCULATING-FUEL REACTORS LTSF Calculation NDA Calculation Author’s Calculation Correction Factors Used Reflector thickness First boron curtain 1.45 Sodium mass 0.833 LTSF source power 1.71 NDA to ART normalization factor 1.45 1.45 2.5 2.9 2.61 426 x 107 kg of sodium per core neufron per curie per activation Previous and Normalized Values Previous values of NaK 470 curies activity Normalized NaK activity 970 curies from core neutrons Normalized NaK activity 87 curies from delayed neutrons Total normalized NakK 1057 curies activity 0.3 x 'Il')_7 activations 55 curies per core neutron per kg of sodium (in NaK) 463 curies 600 curies 50 curies 650 curies Activity of Mass-Transferred Material in the ART Radiators A, M, Perry Electronuclear Research Division A potentially troublesome aspect of the wrap- around heat exchanger design of the ART, when applied to an aircraft power plant, is the degree of radioactivity that will be induced in the secondary heat-transfer circuit, Activation of the primary coolant, NaK, has been extensively discussed (see above). Estimates of Na?4 activity in the NaK have varied, depending upon the particular reactor design being studied, from a few hundred to o few thousand curies. While the activity would not be sufficient, in general, to affect shield design appreciably, it would certainly interfere with routine maintenance of the aircraft engines, At least two solutions have been proposed for dealing with this problem. One solution entails the installation of an intermediate heat-transfer circuit to isolate the coolant for the fuel circuit from the engine radiator circuit; the other consists 30 in draining the radioactive NaK from the radiator circuit and replacing it with fresh NaK before the aircraft is approached for maintenance. While the latter solution is much to be preferred from the standpoint of aircraft weight and performance, there still remains the possibility that constituents of Inconel, which have become activated in the heat exchanger and carried in the fluid stream to the radiators, will be deposited on the radiator tube or header walls and will create a residual radiation field too high to permit unshielded access to the engines. An estimate of the activity of the mass- transferred material that will be deposited in the radiators of the ART has been made, and extrapo- lations to higher power reactors would ke reasonably straightforward, Specific Activity of Mass-Transferred Material, = It is assumed, for the present, that all materials in the heat exchanger region are activated primarily by thermal neutfrons. Activation rates are then proportional to the thermal-neutron absorption cross sections of the several materials present, The Na?4 activity in the NaK is known, and, since it serves as a normalizing factor, the thermal- neutron flux need not be known, {f a material is being activated in a constant neutron flux, the saturated decay rate is equal to the rate of activation, Thus the saturated gamma- ray activity per gram of an element resulting from activation of an isotope of the element is given by A 0 1, = - S0l (in gamma rays per gram-second), where A, = Avogadro’s number, 6,023 x 1023, < A = atomic weight of the element, S; = abundance of isotope j in the element, o; = capture cross section of isotope 7, 1’]- = gamma rays per disintegration of the product isotope, activating flux, ¢ In particular, the sodium activity is given by A 0 Ive = BTy (0.49 barns) (2 gamma rays per disintegration) (¢} , and, then, using the sodium activity as the normal- izing factor, S.a./. 177 - @) oJ For the ART the saturated sodium activity is 1000 curies and the sodium mass is 35,000 g, Therefore I, = 0.029 curies/g, and Si94 5 l,; = 0.66 After operation of the reactor at constant power for a time ¢, the activity of isotope j is less than I . by the factor (1 - e"t/Tf'), where 7. is the mean life of the product isotope. The activity of isotope j (or, rather, of the product isotope re- sulting from neutron capture by isotope j) after operation for a time T is therefore PERIOD ENDING MARCH 10, 1956 T —-t/T. =(T=t)/T; - j i A]. = fo mW Io].('l - e e dt ) =T/T. =T/, - N . I = mW, Ioj [Tj (1 —e Y- Te ] T, ] ~T/7. ~-T/T. MW L iy - 7 fiMWkIOj. [T (1 ~e e } . where m = rate of mass transfer, grams per unit time, W, = weight fraction of the element in the mass- transferred material, M = mT = the total amount of mass-transferred material. Spectrographic analyses of Inconel and of mass- transferred material taken from the cold leg of a sodium-Inconel forced-circulation loop are shown in Table 1.3. The contributions of each of the materials to the activity in the ART radiators, after 1000 hr of operation, are shown in Table 1.4, Only those isotopes are listed that lead to gamma-emitting isotopes. Isotopes that are stable, that emit beta particles only, or that decay by electron capture with no nuclear gamma rays are omitted from the table. The over-all activity of the mass-transferred material is about 1.2 x 10™% curies/g. (By con- trast, the saturated activities of pure sodium, pure cobalt, and pure manganese in a thermal-neutron flux of 10" neutrons/cm?.sec are 0.035, 1.0, and 0.4 curies/q, respectively., The saturated fission- product activity of the fuel in the ART is about 100 curies/g.) Since chromium, manganese, and cobalt are the principal contributors to the activity in the mass- transferred material, it is well to re-examine the assumption that their activation rates are pro- portional to their thermal-neutron cross sections, especially since cobalt and manganese are known to have pronounced resonances in the same general energy range as that of sodium. The neutron flux as a function of energy was obtained from multi- group calculations performed by the Curtiss-Wright Corp. Absorption cross sections as a function of energy were calculated from the Breit-Wigner 31 ANP PROJECT PROGRESS REPORT TABLE 1.3. COMPOSITIONS OF INCONEL AND OF MASS-TRANSFERRED DEPOSITS IN AN INCONEL FORCED-CIRCULATION LOOP THAT CIRCULATED SODIUM Abundance in Mass=Transferred Element Abundance in Inconel (%)* Deposits (%)* Nickel 72.0 10 79.0 N to 95 Chromium 14.0 to 17.0 5 te 10 Iron 6.0 to010.0 0.2t0 0.5 Mangoanese 0.25 te 0.50 1 to?2 Carbon 0.02 to 0.07 Copper 0.15 to 0.50 <0.1** Silicon 0.15 to 0.50 <0.1** Sulfur 0.007 1o 0.015 Titanium 0.15 to 0.50 was also approximated by an ana- fytical expression., This expression was modified to include the fraction of the long-lived decay gamma-ray emitters, which were not measured ex- perimentally, and to omit the fission-product gases, which are expected to escape from the reactor. The resulting spectrum is 10e=*33F photons/Mev- fission. This expression was integrated between the same energy limits as those used for the prompt-spectrum calculation, and the integration yielded information of the same nature. Almost all the capture-gamma-ray data were taken from papers by Kinsey, Bartholomew, and Walker, who give capture-gamma-ray spectra for gamma- ray energies above about 3 Mev. The references to these papers, and some of the data used, are given by Mittleman and Liedtke.!® Because of 12y, Goldsteih and J. E. Wilkins, Jr., Calculations of the Penetrations of Gamma Rays. Final Report, NYO- 3075 (June 30, 1954). 13,, Motetf, Miscellaneous Data for Shielding Calcu- lations, APEX-176 (Dec. 1, 1954). g, Gamble, Prompt Fission Gamma Rays from Uranium 235, unpublished Ph.D. thesis (June 1955). ]5R. W. Peele et al., ANP Quar. Prog, Rep. Dec. 10, 1955, ORNL-2012, p 226, Fig. 13.22, ]6P. S. Mittelman and R. A. Liedtke, Nucleonics 13(5), 50 (1955). 35 ANP PROJECT PROGRESS REPORT the lack of information on the spectrum below 3 Mev, a spectrum for these gamma rays was chosen rather arbitrarily in such a way that the integral under the total spectrum gave the neutron binding energy of the compound nucleus. Most of the capture-gamma-ray energy is given off above 3 Mev; so errors in the assumptions should not contribute significantly to the total error. Partial integrations were performed numerically over the same energy range as that used for the prompt- and decay- spectra calculations. The results of a fairly extensive literature search on the inelastic scattering cross sections and the (n,n",y) gamma-ray spectrum of the materials in the ART are being assimilated. There is insufficient experimental information available, however, and it has been necessary to make approximations and extrapolations over large energy regions. The calculated heat deposition rate, as a result, will have to be used rather cautiously., However, pre- liminary calculations indicate that, in general, radiation resulting from inelastic scattering is not the major contributor to the heating, and therefore the error introduced into the total heat deposition rate is relatively small, Preliminary Calculations. — Some preliminary calculations, as mentioned above, have been made to justify calculational techniques and to deter- mine the gamma-ray sources that can be neglected in calculating the heat deposition rate. Calcu- lations were made, by using slab geometry, of the heat deposition rate at a peint in the pressure shell resulting from decay gamma rays produced in the heat exchanger. In one case the buildup factors for the heat exchanger were used, and in another case the buildup factor for Inconel was used, but the results were essentially the same. The negligible differences in the results indicate that, for gamma rays passing through boundaries of materials of similar Z, the heat deposition rate is nearly independent of the choice of buildup factor, so long as it is restricted to those for similar Z. In another case, it was assumed that the decay- gamma-ray energy was monoenergetic, and the results were compared with the results obtained by integrating over the gamma-ray energy spectrum, Monoenergetic gamma rays of 0.5, 0.75, and 1.0 Mev were used, and differences from the spectrum calculations of as much as 20% were obtained, This indicates that it will be necessary to inte- 36 grate over the spectrum of gamma-ray energies in order to achieve the desired accuracy, The !{‘-in.-thick Inconel liner on the outside of the beryllium reflector becomes a fairly hot source of capture gamma rays, Test calculations of the heat deposition, in slab geometry, made by assuming the Inconel liner to be a plane surface source of gamma rays were compared with calculations made by assuming a plane distributed source of gamma rays. For regions of small penetrations (0.05 mean free paths) there was a difference of 10% between the results for the two calculations, and for regions of deeper penetrations (2.5 mean free paths) the difference became as high as 30%. Therefore care must be taken before it is assumed in the calcula- tions that apparently thin sources of gamma rays may be approximated by plane surface sources. Test calculations were also made to determine the relative source strengths of gamma rays re- sulting from inelastic scattering in the fuel, in the core shells, in the island, and in the first 5 em of the beryllium reflector, By using the fluxes and the absorption rates determined by multigroup calculations in spherical geometry,!? it was found that the energy of the gamma rays produced by inelastic scattering in the fuel con- stitutes about 20% of the total gamma-ray energy released from the fuel. This is a surprisingly large contribution to the total heating, and it cannot be neglected. The contribution to the total heating by the gamma rays resulting from inelastic scattering in the core shells was found to be negligible. How- ever, the gamma rays arising from inelastic col- lisions in the island and in the beryllium reflector near the fuel are significant. They are comparable to the capture gamma rays in intensity, and, al- though their source strength per unit volume is small, their contribution to the heating cannot be neglected because of the large volumes involved. Calculations were also made of the source strength of the capture gamma rays in the pressure shell. The strength was found to be extremely small in comparison with the strengths of other sources, Calculations of the heating in the heat-exchanger— pressure-shell region have been nearly completed, Heat deposition rates for all regions amenable to 17H. Reese, Jr., 8. Strauch, and J. T. Mihalczo, Geometry Study for an ANP Circulating Fuel Reactor, WAD-1901 (Sept. 1, 1954). the analyticodexd numerical approach will be published in a summary report when all the cal- culations have been completed. Application of Methods to a Test Case. — The energy deposition from gamma rays was determined for a test case by the Monte Carlo method and by an analytical approach, for which an exponential form of the buildup factor was used, in order to make a comparison between the two methods and also to gain some insight into the validity of the analytical buildup-factor method for nonhomogene- ous media. The results of these calculations are shown in Fig. 1.9. The source was a uniform plane source of 1-Mev gamma rays adjacent to the fuel region shown in Fig. 1.9. The Monte Carlo calculations were per- formed by using a code developed by Auslender for slab geometry (see Sec, 11, ‘‘Shielding Analy- sis''), The following equation was used for the analytical calculations: S,ue (1) H = {AE1 [(1 v ) Z (pTX)i] + (1 = AE, [ ,- (,,,TX)Z.J} , where H = energy deposition, S = gamma-ray source strength, {t, = linear energy absorption coefficient, gy = linear total absorption coefficient, {(pyX); = total thickness traveled through each slab in mean free paths, A and a are the energy-absorption buildup-factor constants, which depend on the material and on the energy of the incident gamma rays, and E, is the usval exponential integral E,[)/] = J;m e~ Yy~ Vdu The energy deposition was determined in each region by using the buildup factor for each of the materials present. When it is considered that the statistics of the Monte Carlo calculation are not too good because of the relatively small number of histories computed and that the buildup factors are derived on the assumption of a homogeneous medium, the results shown in Fig. 1.9 are guite encouraging. They seem to indicate that the ana- Iytical buildup-factor approach will give reason- PERIOD ENDING MARCH 10, 1956 ably accurate values of the energy deposition for multilayer regions if the buildup factor appropriate to each material is used in computing the depo- sition in that material, The greatest discrepancies between the two methods occur near the boundaries of different materials, where the buildup-factor approach should be incorrect. However, since the equivalent Z of the fuel is very close to the Z of the Inconel, the ratio of the energy deposition on either side of the first boundary (fuel-Inconel interface) should be approximately equal to the ratio of the energy absorption coefficients, which is 0.221/0.094 = 2.35. If the values of the energy deposition, as given by Eq. 1, are used with the fuel buildup factor, the ratio is 19.4/8.2 = 2.36, while from the Monte Carlo calculation the ratio is 13.6/9.7 = 1.40. Therefore it is felt that the results predicted by the Monte Carlo calculation for the first one- half of a mean free path in the first Inconel layer must have a large statistical error and that the analytical buildup-factor approach probably gives the more nearly correct energy deposition at this point. This reasoning cannot be applied to the Inconel- sodium boundaries, since the Z's for these ma- terials are quite different. However, it is felt that the increase in the energy deposition in the last one-quarter of a mean free path in sodium (as pre- dicted by the Monte Carlo calculation) is not o real effect and is caused by statistical uncertain- ties in the calculation. More test cases are being planned in order to get a better estimate on the validity of the buildup- factor approach and to improve the statistics of the Monte Carlo calculation. However, from this one test case, it seems that the buildup-factor approach will give reasonably good answers for the gamma-ray energy deposition in multilayered regions, even though the buildup factors are derived for homogeneous media, Self-Absorption of the Decay Gamma Rays in the ART Fuel Dump Tank H. W, Bertini Aircraft Reactor Engineering Division The self-absorption of the decay gamma rays in the ART fuel dump tank has been calculated,'® By, w. Bertini, An Estimate of the Self Absorption of the Decay Gammas in the ART Fuel Dump Tank, ORNL CF-55-12-47 (Dec. 9, 1955). 37 ANP PROJECT PROGRESS REPORT o ORNL=-LR-DWG 13201 0.20 ’ | \ o | org ‘ | —— MONTE CARLO METHOD | W —— — — FUEL BUILDUP FACTOR USED ‘\‘\ —— - —— INCONEL BUILDUP FACTOR USED | —— \\ [—=--— SODIUM BUILDUP FACTOR USED ‘ A i A \\ 0.16 \ ! \ \\\‘. \ 1 ) o \ \ ] = \ \ S o4 \ \ = ' \ v\ 5 \ \\\ K & \ A L\ : \ \ 2 \ \\ e 012 \ X o \ \ : § \ \ \ ' sopiumM | S \‘ ‘ (MULTIPLY SCALE BY 10) | INCONEL \ @ 10 W N ; < 3\ X > \\ P\ \ ; i ‘ \ W\ W \ W\ > \ W o \ \\\ - \ \ = 008 \\} \\ - — : [T e | ‘ fi \ =z ! re N 5 | AN S | NI\ < 0.06 I\ ] [T \ . i \\l‘ ‘ x\ ‘ | N 0.04 | | | ‘\\ : [ I \ \ NS FUEL INCONEL \\\\\ \ — by | SO NI . < O 002 N N ’ \\\\\.\ 0 [ 0 i 2 3 DISTANCE (MEAN FREE PATHS) Fig. 1.9. Gamma-Ray Heating as a Function of Distance from o Plane Source of 1-Mev Gamma Rays Adjacent to the Fuel Region. 38 The results indicate that 90% of the decay gamma rays originating in the fuel dump tank will be absorbed there, and therefore cooling facilities must be provided in the dump tank to remove almost all the decay heat in the fuel. It was assumed for the calculations that the dump tank contained only fuel and that it was spherical rather than cylindrical. Spheres of two different radii were used, so that one sphere had the same volume as the cylinder and the other had the same surface-area-to-volume ratio as the cylinder. The results obtained for the two spheres differed by only 2%. The expression for the fraction of gamma rays escaping from a sphere was taken from the work of Storm, Hurwitz, and Roe, ! ? for which the straight- ahead approximation was made; that is, it was assumed that only those gamma rays that made first-flight absorption collisions in the sphere were absorbed there, Since multiple-scattering effects were neglected, the expression yields an underestimate of the self-absorption. If the totai gamma-ray cross section is used, rather than the energy-absorption cross section, the expression will yield an overestimate, since every collision is counted as an absorption collision. The energy absorbed in each sphere was calculated by using both the total and the energy-absorption cross sections, The maximum difference in the results obtained was about 10%, A numerical integration was made over the decay- gamma-ray spectrum?® to check its effects, and the value obtained was about 3.5% lower than that obtained by assuming that all the gamma rays were monoenergetic, The estimate that 90% of the decay gamma rays originating in the fuel dump tank will be absorbed there appears therefore to be con- servative. Activation of the Sodium in the ART H. W. Bertini Aircraft Reactor Engineering Division The activation of the sodium to be used as the reflector coolant in the ART was recalculated?’ to correspond to the mid-December design of the ]9M. L. Storm, H. Hurwitz, Jr., and G. M. Roe, Gamma Ray Absorption Distributions for Plane, Spherical, and Cylindrical Geometries, KAPL-783, p 67, Eq. (87) (July 24, 1952). 2OR. W. Peele et al., ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 226, Fig. 13.22. PERIOD ENDING MARCH 10, 1956 reactor., The starting point for the calculations was the Curtiss-Wright multigroup calculations on spherically symmetric systems, specifically their reactor No. 675,22 Attempts were made to modify the multigroup results to account for the asym- metry of the ART and to account for the changed hole distribution in the beryllium island and reflector, The results of this work indicate that the ac- tivity of the sodium will be 5.3 x 101 d/sec or 1.45 x 10% curies. For a total sodium volume of 2.45 13 this represents a decay-gamma-ray source strength of 0.51 w/cm®. An increase of 20% in the volume of the sodium passages in the island and reflector would increase the source strength only about 10%. Shutdown Reactivity of Lithium in the ART Reflector Coolant A. M. Perry Electronuclear Research Division In the event that it becomes impossible to dump the fuel out of the ART, it might be necessary to augment the reactivity effect of the control rod by some poison in order to shut down the reactor to room temperature, The change in reactivity in going from iscothermal operation at 1200°F to room temperature involves the separate effects of di- mensional changes, o thermal-base change, and a fuel-density change, The slow net temperature coefficient of reactivity calculated by Curtiss-Wright,?® when corrected for the fuel expansion coefficient, is 1k ~2.34 x 10~5/°F k dt The temperature coefficient measured in the high-temperature critical experiment was =-2.3 x 10~5/°F. The close agreement of the calculated and the experimental values was considered to justify the use of the Curtiss-Wright calculations of the separate effects involved in going from 24, w. Bertini, Activity of the Na Coolant in the ART, ORNL CF-55-12-78 (Dec. 16, 1955). 22, Reese, Jr., S. Strauch, and J. T. Mihalczo, Geometry Study for an ANP Circulating Fuel Reactor, WAD-1901 (Sept. 1, 1954). 2crRE Temperature Coefficients for 1/6 in. and 1/8 in. Thick Inconel Core Shell, WNE 55-6-2 (May 18, 1955). 39 ANP PROJECT PROGRESS REPORT 1200°F to room temperature. Therefore, for the dimensional changes, 1 4k — — = =077 x10~5/°F , E dt AR = +0.92% ; for the thermal-base change, 1 4k — — = +1.99 x 10~5/°F |, E dt Ak = —2.4% . The results of the high-temperature critical ex- periment?4 showed the specific mass reactivity coefficient of the uranium in the fuel, M dE YR o to be about 0.14, and the Curtiss-Wright calcula- tions2% gave y = 0.18 for the uranium and 0.22 for 24A, D. Callihan et al., ANP Quar. Prog. Rep. Sepl. 10, 1955, ORNL-1947, p 60. 25¢, B. Mills and H. Reese, Jr., Desz'gn Study of an ANP Circulating Fuel Reactor, WAD-1930 (Nov. 30, 1954). 40 the whole fuel salt. By using y = 0.22 and o fuel density of 3.3 g/cm? at 1200°F and 4.13 g/em3 at room temperature, the change in reactivity as a result of the fuel-density change is Ak = +5,0% . Thus the over-all change in reactivity is +3.5%. The poisoning effect of the sodium in the ART moderator (island and reflector) has been esti- mated to be about 0.8% in reactivity, The required reactivity effect is therefore about 4.5 times as great as that from sodium alone. The absorption cross section of natural lithium is 140 times that of sodium, and therefore the required shutdown reactivity could be achieved by adding 4.5/140 = 3.2 at. % of lithium to the sodium, The sodium volume in the moderator cooling system is 2.45 ft3, and, since lithium is less dense than sodium (0.47 vs 0.8 g/cm®), the volume of lithium required would be 2.45 x 0.032 x 0.8/0.47 = 0.13 f+3 , Shutdown from 1400°F would involve a reactivity change of about +4.1% and would require about 0.15 13 of lithium. Even if some credit is taken for the control rod, which could be as little as 1% at 1200°F, 0.1 £t3 of lithium would be required. PERIOD ENDING MARCH 10, 1956 2. EXPERIMENTAL REACTOR ENGINEERING H. W. Savage Aircraft Reactor Engineering Division IN-PILE LOOP DEVELOPMENT AND TESTS D. B. Trauger Aircraft Reactor Engineering Division Loop No. 3 C. W. Cunningham J. A. Conlin Aircraft Reactor Engineering Division In-pile loop No. 3, which was operated in the MTR,! was sectioned in the General Electric Company's hot cell at the National Reactor Testing Station (NRTS) for convenience in shipping to ORNL. The loop was sectioned inte 14-in. lengths that could be accommodated by the available carriers. The sectioning sequence proceeded from the front, or nose, end of the loop to the rear of the pump drive unit, with one additional cut at the intermediate bulkhead. It was possible to return the nose section and the entire pump assembly intact. An unsuccessful attempt was made during the sectioning to determine the cause for plugging of the fission-gas lines, The lines extending through the concrete shield were observed to be open, but it was not feasible to check the passage into the pump or the lines out of the pump in the region where the plugging probably occurred. These lines will be examined in a hot cell at ORNL (see Sec. 8, ‘*Radiation Damage’’), After the sectioned loop was received at ORNL, it was placed in a hot cell, and the fuel was removed. The loop was then sectioned further, and samples were prepared for metallographic examination. This loop, as previously reported, operated with a maximum fuel temperature of about 1500°F. There was a temperature differential in the fuel system of 155°F for 103 hr and 100°F for 168 hr. The metallographic examination revealed corrosion penetration to a maximum depth of 1 mil. Comparison with extrapolations of results ob- tained with forced-circulation loops operated out-of-pile indicates that the 1 mil of corrosion is about what would be expected out-of-pile under similar conditions, and hence irradiation appears L. A. Carpenter et al., ANP Quar. Prog., Rep. Dec. 10, 1955, ORNL-2012, p 28. to have little, or no, effect. However, the oper- ating time, buildup of fission products, and temper- ature differential in the fuel were less than are anticipated in the ART, and the results can only be treated as preliminary, The pump has now been disassembled, except for the impeller housing. Parts in the bearing- housing region appeared, in general, to be quite clean, Less oil has been observed than was expected from the lubrication rate for the seals and bearings. The excess oil may have been absorbed in the Fiberglas insulation used on thermocouple and electrical leads, but this was not fully evident, No direct cause for the stoppage of the purge-gas flow through the bearing housing has been found. The seals and bearings were found to be in good condition, Considerable quantities of decomposed or partially decomposed materials were found forward of the front seal, The seal bellows contained a thick, reddish«black, waxlike substance, and there was a similar deposit on surfaces in this vicinity, The fuel in the pump sump appeared to be mixed with a carbonaceous material, which increased the sump level to above the point for normal operation, It is possible that the presence of these materials caused a friction drag on the pump shaft and produced the erratic-speed behavior observed during operation. Similar materials were found in the off-gas line for the sump purge, but it seems doubtful that these materials could have com- pletely plugged the lines in the regions where they have been observed, No ZrF, vapor (**snow”) deposits were noted. Loop No, 4 C. C. Bolta R. A. Dreisbach D. M. Haines Pratt & Whitney Aircraft J. A, Conlin C. W. Cunningham Aircraft Reactor Engineering Division Assembly of in-pile loop No. 4 was completed on January 20, and it was shipped to NRTS for checkout and operation. Loop No, 4 was the same as the previous loops, except for minor modifi- cations, which included larger diameter tubing for 4] ANP PROJECT PROGRESS REPORT the gas purge system and the substitution of nitrogen for helium in the nose purge system, along with a separate cold trap external to the reactor cubicle, The liquid-nitrogen heat-transfer system for cooling the fission-gas adsorption traps was also revised to provide for more automatic control, and vacuum-jacketed lines were added to minimize nitrogen losses, The loop was found to be satisfactory and was inserted in the MTR HB-3 beam hole on February 6. Isothermal operation started on February 8, and the reactor was brought to full power on February 9, The operating conditions were approximately the same as those for loop No. 3, that is, a temper- ature differential of about 150°F and a maximum fuel temperature of about 1500°F. It is apparent, however, from several observations, that both loop No. 4 and loop No. 3 operated at a somewhat higher power level than was estimated for loop No. 3. The temperature differential observed from the thermocouples on loop No., 4 was between 175 and 200°F and that estimated from heat removal in the air was approximately 200°F, The higher power level indicated by the temperature difference has been substantiated by analyses of the activity in the cobalt-foil flux monitors in loop No. 3. The flux is now estimated to have been 50% greater than that originally calculated from operating conditions, It was not possible to use the nitrogen purge system for the nose region, because the cold traps plugged. The plugging was apparently caused by moisture in the gas. Extensive checks had been made at ORNL to determine the reliability of this purge system, and no trouble had been encountered. As many as 60 bottles of nitrogen, prepared to the same specification as that for the nitrogen normally purchased at the MTR, were passed through a similar cold trap without noticeable plugging. Less than two bottles resulted in plugging of traps at the MTR. Helium was therefore utilized again as the nose purge gas, but no difficulty occurred from heater short circuits, such as was experienced previously,?2 Both the bearing-housing purge system and the pump-sump purge system plugged after ] I/2 weeks of operation, and the helium inlet lines to these systems were clamped to prevent diffusion of radioactive fission gases to the operating area. 2p, A. Gnadt and A, A, Abbatiello, ANP Quar, Prog. Rep. Dec, 10, 1955, ORNL-2012, p 29. 42 Activity appeared in the nose purge gas at one time, but it was contained in the nose-purge-gas traps. This activity apparently came from a leak in the pump bulkhead and did not indicate trouble in the main circuit, Operation was terminated during the scheduled MTR shutdown of February 27, and the loop was removed from hole HB-3 on February 29, High radiation levels were observed in the cubicle during removal of the loop and were apparently due to plated material on the walls of the gas lines. In general, the activity of the lines and of the traps appeared to be higher than had been originally estimated or extrapolated from experience with foop No, 3. The loop circu- lated the fuel for approximately 500 hr and operated with a temperature differential for approximately 390 hr. Loop No, 5 J. A, Conlin P. A, Gnadt Aircraft Reactor Engineering Division Loop No. 5 has been assembled except for thermocouple attachment, insulation, and instal- lation of the water jackets., It is leaktight and pump operation has been checked. Completion of assembly has been delayed until the operation of ioop No. 4 can be analyzed sufficiently to indicate any required changes. The schedule for utilization of the reactor beam hole provides adequate time for this delay. FORCED-CIRCULATION CORROSION AND MASS5-TRANSFER TESTS Fused Salts in Inconel C. P. Coughlen G. E. Mills P. G. Smith Aircraft Reactor Engineering Division R. A. Dreisbach Pratt & Whitney Aircraft Ten electrical-resistance-heated and two gas- heated forced-circulation Inconel loops were operated with fused salts as the circulated fluids during this quarter, The operational data for these loops are summarized in Table 2.1. The results of metallurgical examinations of the loops are presented in Sec. 5, ““Corrosion Research.” A development study of the electrical-resistance- heated loops has revealed that temperature differ- ences of up to 270°F exist between opposite sides e . - — —— TABLE 2.1. SUMMARY OF OPERATING CONDIT PERIOD ENDING MARCH 10, 1956 IONS FOR INCONEL FORCED-CIRCULATION LOOPS THAT CIRCULATED Nc:F-ZrF“-UF4 (50-4 6-4 mole %) Maximum Maximum . Reynolds Temperaf:.lre Fluid Tube-Wall Operating Loop No. Differential Time Comments Number (OF) Temperature Temperature b (°F) CF) () Loops with Electrical-Resistance-Heated Straight Sections 7425-6 10,000 200 1500 1565 1000 Terminated on schedule <7A 10,000 200 1500 1610 1000 Terminated on schedule -8 10,000 200 1500 1570 1000 Area of cooled surface increased _ to five times that of standard loop; tetrminated on schedule -9 4,500 200 1500 1670 Life test Accumulated time in test: 1991 hr . on Feb. 10, 1956 -10 10,000 200 1500 1600 1000 Terminated on schedule -11 16,000 200 1500 In test Heated volume increased fourfold -42 Variable 200 1300 to 1500 1700 Variable Development loop for studying heater design -43 6,500 200 1500 1700 1000 Part of series to determine effect of wall temperature; terminated on schedule . -44 Variable Variable Variable Variable Variable Development loop for studying heater design . =45 6,500 200 1300 1700 320 Terminated because of a failure in auxiliary equipment Loops with Gas-MHeated Coiled Sections <71 Variable Variable Yariable Variable Variable Development loop for studying thermocouples 4935-6 8,000 200 1500 1575 Life test Accumulated time in test: 6637 hr on Feb., 10, 1956 of the heated tubing, top and bottom, when the Reynolds number of the circulated fluid is less than about 5000. A lower limit has therefore been set for the Reynolds number, and the problem is being studied further, Liquid Metals in Inconel and in Stainless Steel C. P. Coughlen A, G, Smith . Aircraft Reactor Engineering Division R. S, Dreisbach Pratt & Whitney Aircraft Six forced-circulation loops were operated with sodium and noneutectic NaK in Inconel and in stainless steel systems during the quarter, A summary of the conditions of operation of these loops is given in Table 2.2, PUMP DEVELOPMENT E. R. Dytko, Pratt & Whitney Aircraft A. G. Grindell Aircraft Reactor Engineering Division Bearing-and-Seal Tests W. L. Snapp, Pratt & Whitney Aircraft W. K. Stair, University of Tennessee The fluid used in initial tests of ART pump bearings and seals for lubrication and heat removal 43 ANP PROJECT PROGRESS REPORT TABLE 2.2, SUMMARY OF OPERATING CONDITIONS FOR INCONEL FORCED-CIRCULATION LOOPS THAT CIRCULATED SODIUM OR NaK T ; Maximum 0 . Cold Reynolds %mpera ‘ure Fluid Operating perating Loop Ne. Differential . Time Remarks Trap Number o Temperature Fluid ( F) OF (hr) ("F) 74266~ Yes 15,000 300 1500 Sadium In test Cold trap at 250°F test (cleaned) -6 Yes Variable Variable Variable Sodium Variable Development loop for studying service plug indicators -7 No 15,000 300 1250 Sodium In test Beryllium insert in hot leg; massive beryllium in stream 7439-3 Yes 15,000 300 1500 Noneutectic 1000 Cold trap at 100°F; terminated NaK on schedule -51 Yes 15,000 Maximum 1500 Noneutectic Life test Started February 13, 1956 possible NaK -52 Yes 15,000 Maximum 1600 Noneutectic Life test Started February 13, 1956 possible NaK was a light-weight, refined, petroleum-base mineral oil, having a viscosity of 65 SSU at 100°F, When it was determined, however, that the temperature of the pump lubricating fluid would be 200 to 240°F in the operating ART, an investigation of suitable synthetic [ubricants was begun, since most low-viscosity petroleum oils have a tendency to “‘coke’ at temperatures above 180°F. Mechani- cally, coking would be of no consequence, but it was feared that the reduction, with time, in the amount of heat removed by the lubricant might be hcrm{ul. The UCON LB series of synthetic lubricants was tested first, These are water-insoluble polyalkylene glycol-base fluids that are available with or without various inhibitors in o wide range of viscosities. Other properties being equal, they are superior to petroleum products in that their specific heat is higher, and, more important, they do not coke, even when used at temperatures well above those specified for ART pump operation, These two advantages are offset, however, by operational limitations. The loosely held hydroxyl ions in the UCON fluids have a tendency to pick up copper from system components and to pla’re it out elsewhere. Unfortunately, this copper plating occurs primarily on the mechanical-seal interfaces. Although this problem could be circumvented 44 by eliminating all copper-brass parts, equally difficult design problems would be introduced, It has also been found to be impossible to es- tablish and maintain good seal performance with the UCON fluids. Early successful operation with the UCON fluids could not be reproduced. One out of ten seals tested with the UCON fluids might be acceptable, whereas further testing with mineral oil showed six to eight out of ten seals to be acceptable, ' The lack of success with UCON fluids does not preclude the eventual use of a synthetic fiuid. Testing and evaluation of other possible fluids ore continuing, However, to permit further pump studies, all pump rotary assemblies are now being tested with the originally rejected petroleum-base mineral oil, but the oil operating temperature is being limited to 180°F. If no single fluid is found to be serviceable, it is possible that a bifluid system will be required, that is, one fluid for cooling and another fluid for lubrication and seal contact., As an aid in the solution of the fluid problem, tests are being conducted to determine the severity of the effect of petroleum-oil coking on heat removal, Since proper control and removal of the lower seal oil leakage are essential to pump operation, the seal design has been modified to include a baffle that extends upward beyond the seal interface to direct any oil leakage downward into a catch basin, from which it can easily be removed by means of a dip tube and continuous gas purging. The Durametallic seals discussed previously3 cannot be used with the modified design, and it now appears that the seal will be a bellows type with a stationary Graphitar 39 nose piece running against a hardened-steel ring (tool steel or AISI-8620 hardened to Rockwell C 50-55) clamped to the rotating shaft. The original leakage-rate specification has been changed from 2 cm3/day or less to 4 to 5 cm3/day. Sodium-Pump Performance Tests with Water H. C. Young Pratt & Whitney Aircraft M. E. Lackey Aircraft Reactor Engineering Division Additional tests of the performance of the ART sodium pump, model MN, with water were performed Sw. L. Snapp and W. K, Stair, ANP Quar. Prog., Rep. Dec, 10, 1955, ORNL-2012, p 33. PERIOD ENDING MARCH 10, 1956 to determine the effect of inlet conditions on performance and cavitation. The volute inlet conditions for the tests described previously4 did not mock up the final design, and the additional tests were therefore made to determine the effect of the modified volute inlet on the pump character- istics. The results of these tests are compared with those of previous tests 5 and 6 in Table 2.3, and performance curves for tests 5 and 7 are presented in Figs. 2.1 and 2.2. The impeller used in test 5 was fabricated according to the final design, and the volute inlet conditions of test 7 were those to be used in the ART pump. A test of @ pump that incorporated all the final design conditions was started, but the inlet shroud vane separated from the impeller because of the failure of a soldered joint, and the test could not be completed. Since all data obtained had shown good agreement, it was decided to consider the tests as having been completed, A typical cavitation curve is presented in Fig. 2.3. Cavitation tests were conducted with 4M. E. Lackey and H. C. Young, ANP Quar. Prog, Rep, Dec. 10, 1955, ORNL-2012, p 35. UNCLASSIFIED CRNL-LR-DWG 13082 140 - ‘ ‘ EFFICIENGY (%) ‘ |( _ DESIGN POINT ‘ 120 | | | TN | ‘ / | |- VOLUTE /1/0 L ’ © \ \/,/ BALANCE LINE / ' | 100 M A o Lo e _ . | >/ m //‘;, // ’ /J ) - . 2900 e 7 / \ — 80 "7/—/ _'.00 rpm - /’l _% /71 \;i T +— 2 / / -~ +, / , - ///, .7\/\ // e 4+ [ L F it 400 R8T 7~ ' / / I 60 240 —-60-62—65 , T | s 7 / / & h’\ / 60 Y~ 4 o 55 "~ 0 A ‘ | 20 e | 0 . 1 0 50 100 150 200 250 300 350 400 450 500 550 600 650 FLOW (gpm) Fig. 2.1. Sodium-Pump Performance Test 5. 45 TABLE 2.3, RESULTS OF WATER PERFORMANCE TESTS OF ART SODIUM PUMP Design point: 430 gpm, 90-ft head Impeller Clearance , T f Inlet Effici Test No, Inlet Conditions (in.) ype ob Inle retency Remarks Shroud (%) Axial Radial 5* Volute inlet plate 3162 in. G.050 0.0625 Shroud 0.258 in. shorter 64 thick; opening 3.5 in. in than in previous test; diameter with 3]{32-in. shroud vanes removed radius at edge; 0.37Q~in. spacetr between plate and bottom of impeller & Same as in test 5 0.050 0.0625 Shroud saome as in test 5; 63 shroud vanes reinstalled and extended 3/32 in. 7** Volute inlet plate 1/2 in. 0.050 0.0625 Same impeller conditions 63 thick; opening 3.5 in. as fest § in diameter with 1,/z-in. rod at edge; no spacer 8 Same as in test 7 0.050 0.0625 Same as in fest 5§ Inlet shroud broke away from impeller; test incomplete Q Same as in test 5 0.050 0.0625 Same as in test 5 Test made to check volute unbalance *Tests 5 and 6 run during previous qucn'fel‘4 with final impeller configuration. **Final volute inlet conditions used. L¥0d3Y $§3¥J0¥d LIO2Ir0dd ANV PERIOD ENDING MARCH 10, 1956 UNCLASSIFIED ORNL-LR-DWG 13083 140 — ‘ ’ I T T ] } T \ T L [/DES\GN POINT | 120 - { - { // | 100 | =L /l . - _ ! -~ ® -~ -y ‘\\ \ \‘:\ i | s 17 _ _[ /. — 80 66 ' | 2 = o6 / \ 2 . DA A Lfi ‘ ‘ > 60 .'\ // * 60 ‘ 55 /X‘Q | voLuTE 50 BALANCE LINE ; 40 fl ’\ J %0 | l T | ) 0 L ] _ 0 50 100 150 200 250 300 350 400 450 500 580 600 650 FLOW {gpm} Fig. 2.2. Sodium-Pump Performance Test 7. UNCLASSIFIED ORNL—LR~DWG 13084 VALUE OF CAVITATION PARAMETER, o 02 03 0.4 05 06 CHANGE N MAXIMUM EFFICIENCY (%) 24 — .2900rpm 90°F, FIXED THROTTLEl 102900 rpm, 92°F, 430-gpm FLOW [0 2400rpm 90°F 380 gpm FLOW 28 , 32— . TOTAL INLET HEAD LESS VAPOR PRESSURE PUMP HEAD 36 | | Fig. 2.3. Sodium-Pump Cavitation Characteristics. two different inlet conditions and with and without vanes on the impeller inlet shroud. The variations in the data were slight, but the data obtained under the conditions that most nearly simulated the design conditions appeared to give the best cavitation characteristics. Estimates of performance with sodium, based on the water test data, show that to prevent cavitation at a sodium temperature of about 1500°F a total inlet head of 6.26 psig will be required; at a sodium temperature of 1285°F a total inlet head of 0.96 psig will be required. The ART sodium pumps are to operate at an inlet pressure of 15 psig, and therefore the cavitation character- istics of the pump appear to be satisfactory, Cavitation data will also be obtained in high- temperature tests with sodium. Four static-pressure taps were equaily spaced on the pump volute, at the periphery of the impeller, to obtain measurements of the radial hydraulic forces on the impeller. It was found, however, that one of the taps had been improperly located in a corner of the volute near the tongue, and it measured part of the velocity head in addition to the static pressure, The other three pressure taps indicated that the balance was satisfactory. |In order to investigate the balance more thoroughly, 47 ANP PROJECT PROGRESS REPORT the number of taps was then increased to nine. The data indicated that the unbalance at design head and flow was somewhat [ess than that measured with the improperly located tap and somewhat greater than that measured by the other three taps. The volute balance lines are shown in Figs. 2.1 and 2.2. At design point the maximum unbalance is less than 7 psi, aond therefore the balance is considered to be satisfactory. It appears from the water test data that the sodium pump will produce the required design head of 90 ft and flow rote of 430 gpm at an efficiency of 63% at 2860 rpm and that it will require 15.6 bhp. To obtain the high head required with a minimum speed, an impeller vane angle of 90 deg is used. As a result, the head-vs-flow curve is rather flat for parallel pump operation. The design point lies on the negative slope of the curve, however, and it is believed that satisfactory, stable, parallel pump operation will be achieved, NoK-Pump Performance Tests with Water H. C. Young Pratt & Whitney Aircraft M. E. Lackey Aircraft Reactor Engineering Division The test stand used for water tests of the ART NaK pump consists of a 6-in, closed-pipe loop; a 200-hp wound-rotor motor, with speed control by variable resistance; a flow-metering orifice; a heat exchanger to remove the heat added to the water by the pumping power; a throttling valve; a test volute; a test impeller; and the necessary instru- mentation, The test volute was a precision aluminum casting, and the test impeller was fabricated of bronze., The preliminary tests indicated that the primary design heod and flow rate could be obtained at speeds below the design speed. lLeckage flow around the impeller hub was excessive, and serious ingassing occurred at design point, with the pump pot acting as a surge tank. For the first test the pump was operated with the pump pot full of water, and a separate surge tank was connected to the loop. Disassembly of the pump after the first test disclosed that the volute tongue had been damaged by local cavitation and that the impeller had rubbed against the side of the volute, For further testing, the volute tongue was filed to @ smooth contour and re-used. Vanes were installed on the back of the impeller hub to 48 decrease bypass flow and thus to decrease the ingassing that resulted from splashing, Also, a sleeve was installed in the volute to decrease the cross-sectional area at the point of highest hy- draulic unbalance. As the tests progressed, changes were made to hub vanes and the volute as indicated by the tests. The performance curve obtained in test 7 is shown in Fig. 2.4, The curve shows that the design point can be reached at considerably below the maximum motor speed of 3550 rpm. The efficiency is approximately 75% at design speed, and the volute unbalance is within satisfactory limits, Additional tests indicated that the impeller will cavitate at an inlet pressure of 5 psig, which is below the design inlet pressure of 10 psig under ART conditions. This condition is considered to be acceptable, but an alternate design is being prepared. Further tests are to be conducted to investigate more thoroughly the degassing character- istics, to determine the optimum volute tongue angle, and to check further the performance and cavitation after minor revisions have been made to the pump assembly. High-Temperature Tests of ART Fuel Pumps S. M, DeCamp, Jr, Aircraft Reactor Engineering Division The startup of an ART fuel pump (model MF-2) on a short-circuit test stand (No. 1) was described previously, After the initial difficulties had been overcome, the pump operated continuously at ¢ maximum fluid (NaK) temperature of 1400°F for 2056 hr. No further operating difficulties were encountered until about 4 hr before termination of the test, when the power trace of the drive motor increased sharply to about 1Y times normal, or 7 kw. This power increase lasted several minutes, and then the power level returned te normal, Approximately 4 hr later the power consumption again increased to about 'l‘/z times normal and continved to vary rapidly over a wide range, When the power consumption failed to return fto normal, the test was terminated to prevent possible damage to the pump. Upon disassembly of the pump it was discovered that the lower seal oil-leakage removal system had not been functioning properly and that some oil had gone down the shaft, The low seal leakage rate 35, M. DeCamp, ANP Quar. Prog, Rep. Dec, 10, 1955, ORNL-2012, p 36. previously reported® was, of course, invalidated. The seal was not damaged. Since this test was run without continuous purge-gas flow through the system, hot NaK vapor apparently diffused up the annulus around the shaft, contacted the lower seal leakage oil, and formed the greasy deposits shown in Fig. 2.5. Close examination has revealed no damage to the pump, The pump is designed so that oil leaking past the seal will flow across the seal runner to the PERIOD ENDING MARCH 10, 1956 shaft and down the shaft into the pumped fluid or will flow down the side of the seal bellows and into the oil catch basin, Qil is forced from the catch basin by applying gas pressure to the surface of the oil and forcing it up an annulus around one of the tubes that carries coolant to the shield plug. In this test it was possible, apparently, for the gas to pass through the annulus without carrying the oil over with it. Efforts are being made to improve the catch-basin drainage system UNCLASSIFIED ORNL-LR-DWG 13085 450 ! 400 |- o ‘5660 " ‘5600 © 350 ‘5“00 (et " 3309 o0 ‘ m 300 ‘ L 30— o , cne / DESIGN POINT — / \ | . _ \ ’ EFFICIENCY (%) ' | VOLUTE BALANCE LINE /“/ ‘ 250 _ £ | 2 | w i 200 i ] 150 100 - 50 - — 0 l i I 0 200 400 600 800 1000 1200 1400 1600 1800 FLOW (gpm} Fig. 2.4. NaK-Pump Performance Test 7. 49 ANP PROJECT PROGRESS REPORT UNCLASSIFIED PHOTC 25418 UNCLASSIFIED PHCTO 17449 Fig. 2.5. Disassembled ART Fuel Pump (Model MF-2), Showing Greasy Material That Formed When Leakage Oil Contacted the Vapors from the NaK Being Circulated in the Pump at 1400°F. This pump operated continuously for 2056 hr before disassembly, 50 to ensure the removal of all seal oil leakage and to prevent oil from being slung from the seal onto the shaft, Short-circuit test stand No. 2 was ready for use in November 1955, and another fuel pump was installed, This system was filled with a fluoride fuel mixture and brought to temperature for seal tests, However, the lubrication oil pump was not functioning properly, and the system was shut down. An examination indicated that the sparging impeller located just below the top pump bearing was causing ingassing of the oil system, Also, it was found that the UCON oil used in this system retained the trapped gas to a greater extent than did the Gulfspin 60 oil used in stand No. 1. In an attempt to reduce the ingassing, the oil was changed from UCON LB-300X to UCON LB-65 (the number of the oil indicates the SSU viscosity). The stand was then again prepared for high- temperature operation. After three days of oper- ation, the lower seal developed a gross leak, and operation was therefore stopped. Examination showed that the UCON oil had removed copper from the system and deposited it on steel and graphite surfaces, especially in the seal region, Particles of the copper had entered the space between the graphite seal and the seal runner and caused the leak. Studies are now in progress to find a more satisfactory oil. It appears that Guifspin 60 oil will be acceptable if its heat- stability and heat-transfer properties prove to be PERIOD ENDING MARCH 10, 1956 satisfactory., The ingassing problem is also being studied further, It may be possible to circumvent the problem by using a larger oil reservoir or by removing the sparging impeiler, The tests conducted to date on the short-circuit test stands are described in Table 2.4. High-Temperature Pump-Performance Test Stands J. J. W, Simon R. Curry H. C. Young Pratt & Whitney Aircraft S. M. DeCamp Aircraft Reactor Engineering Division The design layouts and details and the fabrication of components for two test stands for high-temper- ature performance and endurance tests of ART fuel pumps were completed. These loops, which were described previously,® will be used to obtain performance data on head, flow rate, cavitation and vibration characteristics, and the functioning of the xenon-removal system and to test endurance, A similar stand for testing sodium pumps is being fabricated, and stands for testing NaK pumps and various special pumps have been designed. 6R. Curry and H. Young, ANP Quar. Prog. Rep, Sept- 10, 1955' 0RNL']947' p 44. TABLE 2.4, ART FUEL PUMP SHORT-CIRCUIT LOOP TESTS Test Stand Hours Speed Temperature . . R n for Test Date Started No. No. Type of Test Run (fpm) P Fluid Type of Gil e'?::minourfio:s Troubles Discovered 8-31-55 1 1 Endurance 168 2600 1400 NaK Gulfspin-60 Bad shield plug Bad heater and bad O-ring seal 9-13.55 2 1 Endurance 26 2600 1400 NaK Gulfspin-60 Bad shield plug 9.20.55 3 1 Endurance 2056 2600 1400 NaK Gulfspin-60 Increased power Qil leak to system consumption 12-1.56 4 2 Seal 16 2600 1200 He UCON-300 Oil system in- gassing 12-16-56 5 2 Seal 96 2600 1200 He UCON-65 Seal failure Copper plated out on seal face 2.8-56 6 2 Seal 12 2600 1000 He Gulfspin-60 Seal failure 51 ANP PROJECT PROGRESS REPORT HEAT EXCHANGER DEYELOPMENT E. R. Dytko Pratt & Whitney Aircraft R. E. MacPherson J. C. Amos Aircraft Reactor Engineering Division Intermediate Heat Exchanger Tests J. W. Cooke H. C. Hopkins L. R. Enstice Pratt & Whitney Aircraft Test operations were continued on intermediate heat exchanger test stands A and B. A summary of the tests conducted is presented in Table 2.5. York radiator No. 3 and Pratt & Whitney (PWA) radiator No. 2 failed in stands A and B, re- spectively, during the quarter. Both radiators failed on the air-upstream side at or near the base plate, The side plates of PWA radiators Nos. 1 and 2 were slit prior to initial operation (modi- fication 1). The York radiator No. 3 side plates were removed, and the support and base plates were split (modification 2) prior to operation. Eighteen thermocouples were installed on the tubes of York radiater No. 3 to obtain data on individual tube temperatures. The radiator is shown in Fig. 2.6, with the thermocouples installed. Temperature readings provided by these thermo- couples indicated that, when NaK flow was stopped and restarted, transient temperature conditions occurred because of unequal NaK temperatures throughout the loop. Typical transient temperatures are indicated in Fig. 2.7. The tube temperatures measured by thermocouples 1, 2, 5, and 7 varied much more rapidly and through greater extremes than did the bulk NaK inlet and outlet temperatures measured by thermocouples 8 and 9. York radiator No. 9, which is being installed in stand A, is the first radiator fabricated according to the revised design. It has no side plates, support plates, or base plate., Nickel plates, 10 mils thick, with oversize holes provide top and bottom air seals. It is hoped that these design revisions will alleviate thermal-stress concen- trations in the fin-tube matrix, Cambridge radiators Nos. 1 and 2, modified to be identical to York radiator No, 3 except for two additional slits cut in the base plate parallel to the air flow (modification 3), were installed in stand B, and test operations were started. This test was interrupted, however, by failure of the 52 Fig. 2.6. York Radiator No. 3, Showing Thermo- couple Installations. intermediate heat exchanger, ORNL-1 (type IHE-3), which had operated for 1825 hr. Construction of test stand C was delayed be- cause of difficulties in procurement of the heat exchanger, and therefore the Cambridge radiators Nos. 1 and 2 originally procured for stand C were installed in stand B. The decision was then made to convert stand C to a test facility for prototype ART radiators, Each radiator to be tested will be one half of a complete ART radiator unit, The modified test facility will be capable of testing these radiators at approximately 85% performance, The thermal stresses which will be encountered by the ART radiators will be substantially duplicated in these tests, Small Heat Exchanger Tests L. H. Devlin J. G. Turner Pratt & Whitney Aircraft Test operations continued on small heat ex- changer test stand B, and stand C was placed in £9 TABLE 2.5, SUMMARY OF OPERATION OF INTERMEDIATE HEAT EXCHANGER TEST STAND Test Unit Hours of Nonisothermal Total Hours of Number of . R o . . Operation Operation* Thermal Cycles eason for Termination ORNL heat exchangers Nos. 1 and 2 {type IHE-2) ORNL radiators Nes, 1 and 2 York radiators Nos, 1 and 2 York radiator No, 2 York radiator No. 3 (medification 2) Circulating cold trap (modified diffusion) Circulating cold trap (80-gal system) NaK screen filter Plugging indicator (5 hole, 0,030 mil) Plugging indicator (7 hele, 0.030 mil) ORML heat exchangers Nos, 1 and 2 PWA radiators Nos. 1 and 2 {modification 1) Cambridge radiators Nos. 1 and 2 (modification 3) Circulating cold trap (with precooler) Circulating cold trap (80-gal system) Plugging indicator Test Stand A 15 358 73 Heat exchanger failed 70 621 8 ORNL radiator No. 1 failed 45 150 20 York radiator No. 1 failed 105 437 39 York radiator No. 2 failed 1 361 20 York radiator No, 3 failed 437 Replaced by 80-gal+system cold trap 361 Test continuing 120 Test continuing 360 Test continuing 648 Test continuing Test Stand B 993 1825 21 Heat exchanger failed 585 1195 20 PWA radiator No, 2 failed 408 603 1 Test continuing 1195 Replaced by 80-gal-system cold trap 630 Test continuing 1825 Test continuing *For tests in progress, the total operating time is shown as of February 15, 1956. 9561 ‘0L HOIVYW ONIONI Q0I¥3d ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL-LR-DOWG 130488 1400 T Nak PUMP RESTARTED ! g | i 1000 2 \\ 800 —— \ - 600 £ V w 5 |q_: 400 m r w a Z | ul | 200 : - THERMOCOUPLE LOCATIONS O o o TIME (min) Fig. 2.7. Typical Transient Temperotures That Resulted from Stopping and Restarting NaK Flow Through York Radiator No. 3. initial operation during the quarter. A summary of operation of these stands, which were described previously,” is presented in Table 2.6. ORNL radiator No. 3 failed in stand B after 716 hr of operation. The failure occurred on the air-upstream face of the radiator at or near the center support plate., The design of this radiator was identical to the design of ORNL radiator Ne. 1 and York radiators Nos. 1 and 2, which failed in intermediate heat exchanger test stand A, except that the side plates were not welded together (modification 1). This radiator is shown before and after failure in Figs. 2.8 and 2.9. York radiators Nos. 4 and 5 were modified by removing the side plates and slitting the support 7). C. Amos, L. H. Devlin, and J. G. Turner, ANP Quar. Prog. Rep, Dec. 10, 1955, ORNL-2012, p 49. 54 UNCLASSIFIED Y6655 SUPPORT PLATES NOT WELDED = Fig. 2.8. ORNL Radiator No. 3 Prior to Installa- tion in Small Heat Exchanger Test Stand B. plates and base plate (modification 2) and are now in operation in test stands B and C, respectively. The installation of radiator No. 4 is shown in Fig. 2.10. Heot-Transfer and Pressure-Drop Correlations J. C. Amos Aircraft Reactor Engineering Division Radiator air pressure-drop and heat-transfer data obtained on the heat exchanger test stands operated during the quarter have been added to previous data® and are presented in Figs. 2.11 and 2.12, The additional heat exchanger heat-transfer data 8R. D. Peak and J. C, Amos, ANP Quar, Prog. Rep. Dec. 10, 1955, ORNL-2012, p 49. Fig. 2.9. PERIOD ENDING MARCH 10, 1956 Photo 25131 ' f‘\ ’;P’fl“ ‘ UNCL ASSIFIED ORNL Radiator No. 3 After Failure in Small Heat Exchanger Test Stand B. 55 TABLE 2.6. SUMMARY OF OPERATION OF SMALL HEAT EXCHANGER TEST STAND Hours of Tota! Hours Number of i Nonisoth ] R for Terminati Test Unit oniso .erma of Operation* Thermal Cycles eason tor lermination Cperation Test Stand A** ORNL heat exchanger No. 1 1540 1645 12 Test completed (type SHE-1) Test Stand B ORNL heat exchanger No. 1 816 1710 27 Test continuing (type SHE-2) ORNL radiator Na. 3 295 716 5 Radiator failed (modification 1) Y otk radiator No. 4 520 1000 22 Test continuing (modification 2) Circulating cold trap (80-gal 1710 Test continuing system) Plugging indicator 1710 Test continuing Test Stand C ORNL heat exchanger No. 2 0 400 0 Test continuing (type SHE-2) Yotk radiator No, 5 0 400 0 Test continuing {modification 2) Circulating cold trap (80-gal system) Plugging indicator Test continuing Test continuing *For tests in progress, the total operating time is shown as of February 15, 1956. **Rebuilt for core-shell stability tests upon completion of one heat exchanger test. and fluoride fuel pressure-drop data are in good agreement with the previous data, Radiator NaK pressure-drop data are presented in Fig. 2.13. The NaK pressure drops measured during nearly isothermal operation were observed to increase when oxide plugging temperatures were above 900°F (370 ppm O,). Subsequent reduction of contamination (plugging temperatures below 600°F, 100 ppm O,) by cold-trap operation reduced the NaK pressure drops to the initial levels, In every instance the NaK pressure drop gradually increased when the system was operated with a continuous temperature differential, For example, after 430 hr of nonisothermal operation, the NaK pressure drop across Cambridge radiators Nos. 1 and 2 increased over 30%, while the pressure drop 56 across the heat exchangers in the same stand increased only 6%. All heat exchanger test stands are currently equipped with 80-gal-system circulating cold traps of the type previously described.? These cold traps are capable of maintaining a contami- nation level of 100 ppm 0,, as represented by a plugging temperature of approximately 600°F, in the 80-gal NaK system in the intermediate heat exchanger test stands. On the small heat ex- changer test stands, with approximately 20-gal NaK systems, the cold traps are capable of maintaining a contamination level that is below %F. A. Anderson and J, J. Milich, ANP Quar. Prog. Rep, Sept. 10, 1955, ORNL-1947, p 54. Fig. 2.10. PERIOD ENDING MARCH 10, 1956 UNCLASSIFIED Photo 25333 SUPPORT PLATES AND BASE PLATES SLIT. York Radiator No. 4 as Installed in Small Heat Exchanger Test Stand B. 57 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL—LR--DWG 13092 Of I CAMBRIDGE—1,-2 ~- ! [ AIR PRESSURE DROP PER TUBE ROW (in. OF Hs0 {CORRECTED FROM AVERAGE AIR TEMPERATURE TO 60°F) PWA—1,—2- | | i 1 | i ‘ | L] 0.05 : J 0.2 04 086 10 2.0 40 60 100 AIR MASS FLOW RATE THROUGH MINIMUM RADIATOR FLOW AREA (Ib/secflz) Fig. 2.11. Correlation of Radiator Air-Pressure- Drop Data Obtained from the Operation of Heat E xchanger Test Stands. UNCLASSIFIED ORNL - LR-DWG 13083 o] T T ’7%— CAMBIRJID(ESEE-]:(Z o l 1 . - ORNL-3 T \/ | - | i 3 Z 2 3 : B i o e J—— =<1 /—-——————‘ ° 0.2 ‘ 0 0 i 2 3 4 5 6 7 8 9 10 1 DISTANCE FROM MIDPLANE (in.) Fig. 3.3. Neutron-Flux Distribution Along the Fuel-Reflector Interface of the Compact-Core Reflector- Moderated-Reactor Critical Assembly. 66 PERIOD ENDING MARCH 10, 1956 ORNL-LR-DWG 13437 « \ , ref————|SLAND ———®="~mt— FUEL REGION —— g -—BERYLLIUM REFLECTOR o 30 - ; - - cem - - e A INCONEL, 140 g, 2% x 275 x 043 in. O IRON, 103.8 g, 2% x 2% x 040 in. A MOLYBDENUM, 148.5 g, 27 x 2% x 040 in. a 0 o NIOBIUM, 94 g, 2% x 2% x 0.082 in. HASTELLOY B, 410.6 g, 27 x 2% x 0090 in. NICKEL, 122 g, 2% x 275 x 0104 LOSS IN REACTIVITY PER 100 g CF SAMPLE (cents) . ‘it : \[1 ‘ | . | o - _ : v ’% ‘ '\j\\o\\ 0 ’ ‘ T%—‘D ‘ i \. 0 5 10 15 20 25 SAMPLE POSITION; RADIAL DISTANCE FROM AXIS (in.) Fig. 3.4. Reactivity Change Effected by the Addition of Yarious Materials in the Compact-Core Reflector- Moderated-Reactor Critical Assembly. 67 LOSS IN REACTIVITY PER 100 g OF SAMPLE (cents) 30 25 n o o o ORNL—LR—DWG 13136 > » i J-—MULTIPLY SCALE BY {00 FUEL BERYLL!IUM J_REGION_| : REFLECTOR | | TYPE 316 STAINLESS STEEL, 49.06 q, 2% x 1% x 0.086 in. TYPE 316 STAINLESS STEEL, 525 g, 4%g x 10Y5 x 0.096in. SODIUM, 12¢73 g, 2T % 2% x 1in. BORAL (165 mg OF BORON /em?), 4.4 g, 27/8 X '-016 X '/8 in, i e — | | | 4 | 1 25 ISLAND _| O L O 5 SAMPLE POSITION; RADIAL DISTANGE FROM AXIS (in) Fig. 3.5, Reactivity Change Effected by the Addition of Various Materials in the Compact-Core Reflector-Moderated-Reactor Critical Assembly. 68 ORNL—LR—OWG 13138 25 I : ' i | - | © URANIUM, 27.47 g OF uese, S 4%¢ x 2% x 0.008in. £ 20 5 1 o ® BERYLLIUM, 25tq, a 2 x 2% x lin, S B }v(I g x1in < i : « ! i 5 15 Lo o BERYLLIUM | o 'REFLECTOR! Q T I o« i b i L ; . 10 ———————— —| t i > o i = . | 2 i‘. | Ll \ i ‘e \ Sl N = “-DIVIDE SCALE BY 10 | F-ZrF4 2 508 10 Optical data: 2.386 12 Cubic, N = 1,420 2.350 75 2,281 1n X-ray data: o 2.275 12 Simple cubic, a = 3.288 A 2,260 12 0 d(A) /1, 2.53 12 3.290 100 2.245 8 2.326 60 1.900 100 18T, N. McVay and G. D. White, The Optical Properties of Some Inorganic Fluoride and Chloride Compounds, 1.646 20 ORNL-1712 (May 5, 1954). 79 ANP PROJECT PROGRESS REPORT 80 O d(A) 2.226 2.152 2.067 2.027 1.959 1.932 1.897 1.868 1.833 1.770 1.763 1.679 d(;‘\) 5.350 4.835 3.590 3.080 2.414 2.338 2,199 1.861 1.789 1.569 1.422 Vi 12 45 12 12 30 30 15 20 20 30 12 2RbF'ZrF4 Optical dota: Uniaxial negative O =1.440 E =1.426 X-ray data: I/[‘ 10 100 34 40 21 17 17 B-RbF-ZrF , Optical data: d(A) 7.37 6.66 6.32 5.53 517 5.01 4.67 4.54 4.46 4,35 3.99 3.88 3.77 3.72 3.69 3.61 3.42 3.34 3.325 3.090 3.060 2.960 2.900 2.822 2.788 2.767 2.678 Biaxial negative; 2V ="75 deg N, = 1.490 to 1,502 Ny = 1,502 to 1.516 X-ray data: I/1 12 42 12 12 30 36 42 42 100 100 100 14 38 45 PERIOD ENDING MARCH 10, 1956 2.614 6 1716 17 2.564 6 1,709 17 2.514 8 1.699 10 2.501 8 1.694 10 2.482 6 1.679 18 2.455 6 1.662 6 2.433 6 1.654 18 2.423 6 2,398 . RbF-NaF-ZrF 2.380 12 Optical data: 2.355 12 Biaxial negative; 2,344 12 2V = 60 deg 2.309 6 Ng =71.385 2.292 8 N, ="~1.395 2.253 6 Often fibrous or platy; polysynthetic and rectangular twinning common 2.240 12 2.211 14 X-ray data: 2,206 15 d(A) 1, 2.191 32 5.21 10 2.116 42 482 s 2.074 10 468 38 2.034 6 4.21 45 1.997 30 375 5 1.989 48 2,65 5 1.935 48 3.59 10 1.929 48 339 65 1.926 55 3.34 10 1.903 30 299 20 1.894 6 3.24 35 1.872 12 2,571 10 1.857 10 2.514 8 1.836 10 2.350 15 1.773 6 2.332 28 1.757 6 2.321 28 1.751 6 2.152 100 1.742 6 2.094 100 1.731 15 1.886 10 1.728 14 1.763 18 1.719 24 1.614 8 81 ANP PROJECT PROGRESS REPORT 82 RbF-NaF-ZZrF4 Optical data: Uniaxial negative 0 =1.446 E =1,435 Often shows platy habit 7.43° 6.71 6.54 5.91 4.55 4.41 4.13 3.90 3.71 3.52 3.40 3.35 3.29 3.23 3.13 2.826 2,481 2,430 2,392 2.380 2.344 2.286 2.237 2.194 2.159 2,142 2.083 2.060 1.983 1.963 X-ray data: /1 40 22 50 60 45 100 12 M—l NN o o W n h " n — -— N e o o O O 12 1.943 1.935 1.920 1.854 1.836 1.757 1.754 1.694 1.676 1.670 d(/?\) 7.80 6.61 6.19 6.03 5.61 5.47 5.21 4.98 4.92 4.80 4.65 4.33 4.09 4.04 3.91 3.80 3.69 3.60 3.51 14 25 40 45 38 30 NaF.2U F4 Optical data: Biaxial negative; 2V =20 to 30 deg N, =1.520 N), =1.584 X-ray data: /1, 21 21 100 PERIOD ENDING MARCH 10, 1956 3.43 21 1.840 7 3.42 21 1.817 27 3.41 21 1.792 7 3.30 43 1.790 7 3.26 70 1.754 11 3.15 21 : 1.742 11 3.12 7 1.730 21 3.08 65 1.720 14 2.98 7 1.643 14 2.90 7 1.638 1 2.87 9 1.586 14 2.813 45 1.584 n 2.747 7 1.576 13 2.731 7 1.572 13 2.675 7 1.554 18 2.622 7 2.600 7 Na,UF o 2.449 8 Biaxial negative; 2.411 7 2V = 60 deg 2.355 1 N, = 1.520 2.290 1 N, =1.600 2.250 7 2,211 7 KU,Fy 2.184 7 Biaxial negative; 2,152 7 2V = 10 deg 2,144 7 N, = 1.520 2.137 7 N,, =1.584 2.128 24 2.090 7 “aFeCly 2.065 28 Uniaxial positive 2.034 14 0 =1.600 2.010 9 E =1.636 1.983 13 I(Fe:CI3 1.947 13 1,939 21 Biaxial positive; 1.931 32 2V =20 deg 1.912 35 Ny =1.700 1.865 10 ' Ny =1.740 ANP PROJECT PROGRESS REPORT ZnF2 Uniaxial positive 0 =1.501 E =1,526 szBeF4 Biaxial; 2V =90 deg Refractive indices near 1.394 Low birefringence RbgBer Uniaxial positive Refractive indices near 1,402 RbBeF3 Biaxial positive Refractive indices near 1.390 Low birefringence CHEMICAL REACTIONS IN MOLTEN SALTS L. G. Overholser F. F. Blankenship G. M. Watson Matertals Chemistry Division R. F. Newton Director's Division A diversified program of chemical studies of molten fluorides at high temperature was continued during the quarter. Evaluations of activities of materials such as FeF, or NiF, in molten NaZrF, are being made by measuring emf’s and by de- termining equilibrium conditions for reactions such as H2 + FeF2 = Fe® + 2HF In addition, the reversible electrodes M%/MF , were shown to be valuable for estimation of solubilities of the metal fluorides in selected fluoride solvents. Measurements of electrode potentials were aiso shown to be applicable to the estimation of the activity of chromium in nickel-chromium alloys. Preliminary studies of the kinetics of oxidation of UF, by KF were conducted. It is not possible 84 to separate the oxidation reaction from the simul- taneous disproportionation of UF,, but, by using some reasonable assumptions, a number of con- clusions regarding the Redox process can be drawn, Determinations of equilibrium concentrations for the reaction of UF, with pure metals in various solvents of reactor interest were continued. The data reveal that Nb° is quite resistant to UF, in NoaZrF . mixtures but that neither W° nor Ta® is of value in NaF-KF-LiF solvents, The list of experimental preparations of fVlI:x-ZrF4 compounds was extended to include Zan-Zer and FeF,.ZrF,. Solutions of alkali halides in molten alkali metals, with the sole exception of lithium halides exhibit intense, brilliant green-to-blue colors in the temperature range 450 to 800°C, in molten Li° were shown to Equilibrium Reduction of FeF, by H, in NaF-ZrF C. M. Blood G. M. Watson Materials Chemistry Division The study of the equilibrium constants for the reaction FeF, (d) + H, (¢ ==Fe (s) + 2HF (g) with a molten mixture of NaF-ZrF, (53-47 mole %) as the solvent, previously described,'? was brought to a successful conclusion. Additional measure- ments were made at 700 and at 600°C. In order to determine whether the extreme length of the period of operation at high temperatures had changed the nature of the solvent, the study was concluded with a series of measurements at 800°C under conditions that duplicated those used during the very early stages of the experiment. Within ex- perimental variations, essential duplication of the previous results was obtained after the four-month period of continuous operation. The experimental results obtained during this quarter are summarized in Table 4.2 and presented graphically in Fig. 4.4. It may be observed that, within the precision of the experiment, no changes occurred in the values of the equilibrium constants over relatively large ranges of concentration of iron. After the experiment was concluded, the remaining portion of melt was sampled and analyzed. The ¢, M. Blood, ANP Quar. Prog. Rep, Dec. 10, 1955, ORNL-2012, p 85. PERIOD ENDING MARCH 10, 1956 TABLE 4,2, APPARENT EQUILIBRIUM CONSTANTS FOR REDUCTION OF Fer BY H2 IN NuF-ZrF4 (53-47 mole %) Reaction: Fe:l'-'2 (d) + H2 (g) == Fe (s) + 2HF (g} Average total pressure: 740 mm Hg Initial charge: 6.0 kg of NaF-ZrF4 and 30.0 g of FeF, Container wall: mild steel (98% Fe) Partial Pressure of HF in Effluent Gas K * {atm X ]02) Fe in Melt Sampling Time {ppm) (days at temperature) Measurements Made at 700°C 760 38 2.81 0.62 725 39 2.67 0.58 725 40 2.76 0.62 775 41 2.84 0.61 675 42 2.72 0.65 Average of 15 measurements reported previously 0.64 Av 0.63 £0.04 Measurements Made at 600°C 1715 6 1.60 0.087 1525 7 1.64 0.107 1290 10 1.39 0.087 1315 11 1.52 0.102 1320 12 1.54 0.104 1245 13 1.40 0.091 1165 14 1.33 0.089 150 21 0.516 0.099 350 28 0.779 0.100 260 33 0.634 0.090 385 34 0.839 0.106 475 35 0.890 0.097 Av 0.097 £ 0,007 Measurements Made at 800°C 790 4 6.18 3.0 715 5 6.00 3.1 810 6 6.17 2.9 880 7 6.25 2.7 940 10 6.30 2,6 915 11 6.39 2.7 900 12 6.14 2.6 1010 13 6.34 2.4 Average of 13 measurements reported previously 2,5 Av 2.06 £0.2 *Kx = Plz-iF/XFeFZPHz’ where x is mole fraction and P is pressure in atmospheres. 85 ANP PROJECT PROGRESS REPORT UMCLASSIFIED ORNL-LR-DWG 12784 3.0 T T & O T . T e AR e e e e Kxao %4% v ! f /fil L~ [:/ "‘flI" e 300 400 500 600 700 8O0 900 1000 1400 4200 4300 0.80 K 060 0.40 600 800 1000 1200 1400 {600 4800 2000 2200 2400 2600 2800 3000 3200 3400 02 600°C J K, 010 [Tty zLL = '!”"”“Trr’ = ] ‘ 0 200 400 600 800 {000 {200 1400 {600 1800 Fe IN MELT {ppm) Fig. 4.4. Apporent Equilibrium Constants for the Reduction of FeF, by H, in NaF-Z:F, (53-47 mole %). results of the analyses of the initial and final mixtures are summarized in Table 4,3. The analy- ses indicate that the composition of the solvent substantially constant throughout the experiment. In order to determine the effect of the solvent on the equilibrium, it is pertinent to compare the experimental values with those that may be calcy- remained lated from the values of the free energies of formation of seolid FeF, and gaseous HF that are available in the literature.2? Since the FeF, used for the experiment was not in solid form but, rather, was dissolved in molten NaF-ZrF ,, another useful basis of comparison is the corresponding equilibrium constant that would be obtained by assuming the FeF2 to be a supercooled liquid. A comparison of the temperature effect on the calculated and ex- perimental constants is also of interest. From these calculations, activity coefficients for the dissolved FeF_ were obtained easily., Theactivity coefficients are useful for describing, in a concise form, the behavior of the equilibrium in solution, with reference to a given standard state. The equations of interest in the calculations are: FeF, (s) + H, (g) = Fe(s) + 2HF (g) il and FeF, (1) + H, (g) = Fe (s) + 2HF (g) . 20L. Brewer et al.,, pp 65, 110 in The Chemistry and Metallurgy of Miscellaneous Materials, Thermodynamics, NNES IV-19B, ed. by L. L. Quill, McGraw-Hill, New York, 1950, 86 TABLE 4.3, NOMINAL NaF-ZrF4 (53-47 mole %) COMPOSITION BEFORE AND AFTER EQUILIBRATION MEASUREMENTS Chemical Analyses Days of Major Constituents Minor Constituents Operation (wt %) (ppm) Na Zr F Ni Cr 1 12,0 42.4 453 55 110 121 12.2 42,8 45.1 30 90 Theoretical 12.1 42.6 45.3 The values of free energies of formation needed are those of FeF, (s), FeF, (1}, and HF (g). These values?? are listed in Table 4.4. |t is realized that the number of significant figures used in the tabulation is greater than is warranted by the precision indicated for the published values, but, since the calculations involve differences, all the significant figures were employed, except for the end results, Thus the significance of future com- parisons with other systems will be limited by experimental accuracy and not by the inherent variations introduced by employing small differences of rounded numbers. The values of the free energies of formation of liquid FeF, were obtained by calculation of the free energies of fusion. A heat of fusion, inde- pendent of temperature, of 8000 cal/mole was TABLE 4.4, FREE ENERGIES OF FORMATION OF FeF, (s), FeF, (1), AND HF (g) Free Energy of Formation {kcal/mole) Temperature 0 FeF, (s) FeF,(l) HF (g) FeF,(d) 800 —-130.45 -128.69 —65.84 -—129.64 700 ~133.95 ~131.61 ~65.76 -—132.41 600 ~-137.45 -134.53 =65.67 ~135.39 used,20 and integration of the equation A(AF/T) AH oT T2 between the temperature in question and the melting point of FeF, (1375°K) gave the free energies of fusion employed. The values listed in the last column of Table 4.4 are the free energies of formation of FeF, (<) obtained from the experimental data by using the free-energy changes listed in Table 4,5, The calculated free energies of formation should pre- sumably incorporate the free-energy changes resulting from solution, dilution, and solvation. At any rate, the calculated values of AF° for FeF, (d) will precisely describe the behavior of the equilibrium over the experimental range studied, subject to the precision of the experimental measure- ments, if the experimental K_ is treated as a true thermodynamic equilibrium constant. TABLE 4,5. PERIOD ENDING MARCH 10, 1956 The free-energy changes of the pertinent reac- tions, as well as the corresponding thermodynamic equilibrium constants, are given in Table 4.5. For purposes of comparison, the experimentally de- termined equilibrium constants are also given, along with the calculated free-energy change for reaction 3 of Table 4.5. The activities of HF at equilibrium with certain ranges of activities of FeF, (s) and FeF, (/) have been calculated and are presented graphically in Fig. 4.5, along with the experimentally de- termined values. The activity of FeF, was taken to be equal to its mole fraction. The mole fractions of FeF , were calculated as described previously. 19 The activities of HF are expressed in atmospheres. It may be observed that the experimental points are, in every case, between the calculated curves corresponding to the two standard states used. The temperature dependences of the calculated and experimental equilibrium constants are com- pared graphically in Fig. 4.6. An inspection shows that the slope of the experimental curve is the same as the slope of the curve calculated by taking supercooled liquid FeF, as the standard state. The slope of the curve calculated by taking solid FeF, as the standord state is steeper. The heats of reaction indicated by the slopes and the entropy changes of the reactions are summarized in Table 4.6. The activity coefficients of FeF were also calculated relative to the standard states that have been considered. in solution different The FREE-ENERGY CHANGES AND EQUILIBRIUM CONSTANTS FOR THE REDUCTION OF Fer BY HYDROGEN Reactions: FeF2 (s) + H2 (2) FeF, (1) + H, (g) = Fe(s) + 2HF (g) FeF, (d) + Hy (g) = Fe (s) + 2HF (g) (1) (2) = Fe (s) + 2HF (g) (3) Free-Energy Changes, AF° Equilibrium Constants Temperature (kcal/mole of Fer) (°C) Reaction 1 Reaction 2 Reaction 3 Ka for Reaction 1 Ka for Reaction 2 Kx for Reaction 3 800 -1.23 ~2.99 —-2.04 £0.16 1.78 4,06 2.6 £0.2 700 +2.43 +0.09 +0.89 £0.12 0.285 0.955 0.63 £0.04 600 +6.11 +3.19 +4,05 £0.12 0.0294 0.159 0.097 £ 0.007 87 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL—LR—DWG 12782 10~ ACTIVITY, OR PARTIAL PRESSURE, OF HF {atm) 3 o= 2 5 103 2 5 o2 ACTOATY, OR MOLE FRACTION, OF FeFy \N NoF—Zrfy (53—47 mole %) 10 Fig. 4.5. Comparison of the Activities of HF at Equilibrium with a Range of Activities of Liquid and Solid FeF, in NaF-Z¢F, (53-47 mole %). UNCLASSIFIED ORNL-LR-DWG '2783 TEMPERATURE (°C) EQUILIBRIUM CONSTANT, A 12.00 (x107%) 9.00 "~ 10.00 .00 Y (2K) Fig. 4.6. Temperature Dependence of Calculated and Experimental Equilibrium Constonts for the Reaction FeF, + H,==Fe + 2HF. 88 TABLE 4.6, HEATS OF REACTION AND ENTROPY CHANGES FOR THE REDUCTION OF FeF, BY HYDROGEN Heats of Entropy . Fer Standard State Reaction, Chonges, o AH® (cal} AS® (eu) Solid {calculated) 37,500 36.0 Liquid (calculated) 29,600 30.3 Dissolved (experimental) 29,600 29.4 equilibrium constants for the reduction reactions are given by the expression: 2 IHF “%Fe (s) . o a, a H2 FeF2 If the activity of solid iron is taken to be unity and the activities are converted to partial pressures or mole fractions, the expression becomes 2 2 Phe YHF Yy H2VFeF2 2 YHE x YH, YFeF, High-temperature extrapolation of the available data?1=23 on association factors of HF (g) indi- cates that, at the temperatures used, HF is monom- eric, If ideal gas behavior is assumed for both HF and H,, it appears quite reasonable to believe that YHE = YH, = 1.0. Thus the activity coefficient for FeF,, as defined by the previous equation, is given by . K X VFer = E; The activity coefficients of FeF, for the different 2lp, L, Jarry and W. Davis, Jr., J. Pbys. Chem. 57, . 600 (1953). 22R, W, Long, J. H. Hildebrand, and W. E. Morrell, J. Am. Chem. Soc. 65, 182 (1943). 23, Simons and J. H. Hildebrand, J. Am. Chem. Soc. 46, 2183 (1924). standard states, calculated with this equation by using the experimental equilibrium constants, are summarized in Table 4.7, Within the precision of the experimental measure- ments, the activity coefficients listed in Table 4,7 are independent of the FeF2 concentration over the ranges investigated. |t is apparent that, as far as this investigation is concerned, the most convenient choice of standard state is FeF, (d). M{ MFQ(Q]): However, since it is believed to be unlikely that the activity coefficient is independent of the concentration for the general case involving other solvents and solutes, it would appear to be more desirable to adopt the hypothetical pure super- cooled liquid as the standard state. TABLE 4.7. ACTIVITY COEFFICIENTS OF FeF, IN MOLTEN NaF-ZrF , (53-47 mole %) FeF2 Activity Coefficient Standard State Ay 800°C At 700°C At 600°C Sofid 1.46 2.20 3.28 Liquid 0.64 0.66 0.66 Dissolved 1.00 1.00 1.00 It is hoped that the equilibrium-reduction reactions of other metallic fluorides will prove to be amenable to investigation by this method and that, by com- parison of similar thermodynamic correlations, information may be obtained on the condition of the solute in the molten fluoride solvent. A study of the reduction of FeF, by H, in other solvents in the absence or presence of UF, is plonned in the hope of obtaining a better understanding of the effect of composition on corrosion. EMF Measurements in Fused Salts L. E. Topol Materials Chemistry Division Determinations of activities and activity co- efficients of metal fluorides from the emf’'s of PERIOD ENDING MARCH 10, 1956 concentration cells in molten NaF-ZrF, (53-47 mole %) were continued.?4 The cells were measured in the temperature range 550 to 700°C in a helium atmosphere, The half cells were contained in recrystallized alumina (morganite} crucibles, and electrical contact between them was achieved with a porous ZrO, bridge impregnated with the NaF- ZrF , solvent. The cells measured were of the type NaF(a,), ZrF ,(a,) [} NaF(aj), ZrF (aj), MF,(a]){M , where M = Fe, Cr, or Ni and (a) denotes the activity of a species. If it is assumed that Na* and F~ are the sole carriers of current, the cell reaction is MF,(af) + 2iNaF(a,) = MF,(a;) + 2:NaF(a}), where ¢ is the transference number of the Na*. If the amounts of added MF, are small, the activities of NaF are approximately equal in both half cells, and the cell potential should be given by RT 4 RT 1%y E = M—E II’I = —2_1—'«‘ n ’ n ’ r 4 V1%, where a,a’, x,x", and y,y” are the activities, mole fractions, and activity coefficients, respectively. Duplicate cells were measured, each for a period of two days, by using iron electrodes and FeF electrolytes. Three different concentration cells were used in the study. Agreement between dupli- cate cells was, in general, better than 13% of the emf observed. The values obtained on the second day for each cell were lower by 2 to 8% than those obtained on the first day. Typical data for the first day are presented in Table 4.8. From these data it is obvious that, for all the unsaturated cells, the change in activity-coefficient ratio with temperature is very small. A comparisen of the activity-coefficient ratios of the various concen- trations studied with the ratio of the most concen- trated mixture studied (x = 0.015) is shown in Table 4.9. The data show that the activity coefficient increases with increased dilution. It should be noted that the decrease in activity coefficient with concentration of FeF, is greatest 24y | E, Topol, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 97. 89 ANP PROJECT PROGRESS REPORT TABLE 4.8. ACTIVITY RATIOS FOR FeF, IN NaF-ZrF , (53-47 mole %) FROM THE CELL Fe| FeF(x,), NaF(x,), ZrF ,(x;) || ZrF (x3), NaF(xJ), FeF ,(x7)| Fe Emf (v) Temp;erufure a]/a{ V]/V; Q) Measured ldeal For x, = 0,000607; x| = 0.002309* 700 0.0505 0.0560 0.300 1.14 650 0.0472 0.0531 0.305 1.16 600 0.0450 0.0502 0.302 1.15 550 0.0429 0.0473 0.298 1.13 For Xy = 0.000607; x; =0,00714* 700 0.0830 0.1034 0.138 1.62 650 0.0790 0.0980 0.137 1.61 600 0.0750 0.0926 0.135 1.60 550 0.0502 0.0874 0.242** 2.85 For x, =0.00776; x; =0,01484* 700 0.0252 0.0268 0.548 1.05 650 0.0245 0.0254 0.540 1.03 600 0.0155 0.0240 0.662** 1.26 550 0.0008 0.0226 * & *|In mole fractions, **Solubility of FeF, in at least one half cell was less than the added concentration of FeF,, TABLE 4.9. VARIATION OF ACTIVITY COEFFICIENTS OF FeF, IN NaF-ZrF , (53-47 mole %) AT 650°C WITH CONCENTRATION xq xy Y17, Y1/Y0.015 0.00778 0.01485 1.05 1.05 0.00270 0.00744 1.17 1.24 0.000607 0.00714 1.63 1.72 0.000607 0.00231 1.18 1.46 in the dilute melts. Unfortunately, it is in the very dilute cells that any slight error in concen- tration (from weighing or from oxidation of the FeF,, etc.) leads to a fairly large error in the emf reading. Further experiments along this line are planned. 90 In the course of these measurements it was noted that a sharp change in emf occurred when the solubility of FeF, was exceeded. Thus, by adding different amounts of the same metal fluoride to the NaF-ZrF, solvent in two half cells and plotting the measured emf as a function of tempera- ture, the temperature at which the solute dissolves completely on heating {or precipitates out on cooling) can be detected by a break or discontinuity in the plot of emf against temperature. In the actual experiments, less than equimolar amounts of metal fluoride and ZrF , were added to NaF-ZrF , (53-47 mole %), since it is believed that the saturating phase in this solvent has the general formula MF,-ZrF,. Thus the solubilities were determined along a quasi binary between 7NaF- 6ZrF, and MF ,-ZrF , so that the composition of the saturated solution would be independent of the total amount of MF, present. The solubilities of - NiF,, FeF,, and CrF, contained in vessels of Ni, Fe, and Pt, respectively, were measured from 550 to 750°C. In all runs the melts were first heated to the temperature at which complete solubility in both half cells was attained. Emf readings were then taken at decreasing temperatures, with ap- proximately 30 min being allowed at each point for equilibrium to be reached. (Constant emf’s at each temperature were usually approached in 15 min.) Each cell was run in this way several times, and both heating and cooling curves were plotted. With both cells saturated, the emf should be zero, and it should be constant during heating until dissolution of the MF,-ZrF, in the dilute melt is complete. At this point the change in emf with temperature is abrupt and continues to be rapid until all the MF_.ZrF, in the concentrated melt dissolves. Subsequent increases of temperature result in fairly small changes in emf. The ZrF, concentrations in the half cells become increasingly different between the solubility temperature of the dilute melt and that of the concentrated one. A typical emf-vs-temperature plot for an FeF. cell is shown in Fig. 4.7. In all cases the data obtained at the lower temperature could be reproduced much better than the data obtained at the higher temperatures. The average deviation of the lower temperature, T, was about 3°C, whereas the higher temperature, T,, varied by about 20°C. Therefore, in practically all cases, the saturation temperature of the more concentrated half cell in each run (T,) was re- determined by measuring this half cell against another still more concentrated half cell, The UNCLASSIFIED CORNL—LR—-DWG 12784 30 o ON COOLING ) o ON HEATING | 20 - e ] E i - ____..-——.-——-.—'_ ’ | 10 —— li,,,./ , —_— — ,___-——-#/ ¢, = 6.07 mole % ! T¢)= 1103 male % 0 ! ' 650 700 750 800 TEMPERATURE (°C) Fig, 4,7. Temperatute Dependence of the EMF of the Cell Fe|FeF,, ZrF (c,) || FeF,, ZrF (c,"}|Fe in the Solvent NaF-ZrF , (53-47 mole %), PERIOD ENDING MARCH 10, 1956 solubilities given in Fig. 4.8 were found in this manner, The log of the solubility in mole % MF -ZeF, is plotted against the reciprocal of the absolute temperature in Fig. 4.9. As may be seen, the solubilities for the dilute melts obey a straight- line relation with respect to 1/T, but in the more UNCLASSIFIED ORNL-LR-DWG 12785 800 - — ‘ — TEMPERATURE (°C) 0 4 8 {2 16 20 MF-ZrF, (mole 7o) Fig. 4.8. Temperature Dependence of Solubility of MF:_,-ZrF4 in NaF-Z(F . UNCLASSIFIED ORNL-LR—-DWG 12786 NiFp ZrFs- 3 0.90 1.00 140 1.20 130 {(x10” Y1 (oK) ) Fig. 4.9. Solubility of MF,.ZrF, in NaF.Z(F, vs Reciprocal of Absolute Temperature. ANP PROJECT PROGRESS REPORT concentrated mixtures there is a deviation from linearity, This deviation is in the direction op- posite to that expected. It appears likely that at the higher temperatures the soaturating phase is no longer MFQ-ZrF‘; it may well prove to be the simple MF, compound, The slopes of the lines at lower temperatures and concentrations in Fig. 4.9 yield heats of solution of 37.5, 35.7, and 23.3 kcal/mole MF ,.ZrF, for the CrF,, FeF,, and Nil:2 compounds, respectively. By extrapolating these lines to unit mole fraction of MF,.Z¢F , “ideal’’ melting points of 775, 857, and 1155°C are found for CrF,-ZrF,, FeF,.ZrF,, and NiF,. ZrF ,. These ideal melting points have no physical meaning, but they may serve as a rough measure of the relative stability of the three compounds. The unexpected difference in both the ideal melt- ing point and the heat of solution of the NiF, complex from that of either the CrF, or the FeF, leads to the supposition that different types ot species may be involved. |t may be that NiF . ZrF , breaks down at temperatures lower than any used for these tests. Activity of Chromium in Alioys M. B, Panish Materials Chemistry Division The activity of chromium in alloys of chromium and nickel is being examined because of the great importance of such alloys as container materials for molten salts, Grube and Flad?25 studied the reaction Cr2O3 + 3H,=— 3H20 + 2Cr and tabulated the pressures of H,0 and H, in equilibrium with pure chromium and nickel-chromium alloys at 1100 and 1200°C, An activity diagram, not previously published, which was calculated from these data is presented in Fig. 4.10. This diagram must be considered to be surprising, since considerable negative deviation from Raoult’s law is exhibited, in spite of a region of immiscibility on the chromium-rich portion of the diagram, [t has been suggested that the experimental work was in error. 26 25G. Grube and M. Flad, Z. Elektrochem. 48, 377 (1942). 26K, w, Wagner, Thermodynamics of Alloys, Addison- Wesley Press, Cambridge, 1952. 92 UNCLASSIFIED ORNL-LR—DWG 12787 . 100 —— N i . | IDEAL— / ' i ‘ y I 0.75 |——— - : b ;- \ 3 / i I i c | / ! = ! = | ¢ 5 f - / : ; < J ; a w 050 |——- o / — | / 2 ' | : 2 X ‘ < | REGION OF & = IMMISCIBILITY—- s | oasl I S N e | | 50 75 i Cr IN Ni {mole %) Fig. 4.10. Apparent Activities of Nickel- Chromium Alloys at 1100°C. The activity of chromium in nickel-chromium alioys is being investigated here by emf measure- ments of electrolytic cells of the type Cr®|NaCl, RbCI, CrCl, | Cr(Ni) alloy A schematic diagram of the cell, which utilizes a central reservoir and four electrode compartments, is shown in Fig. 4.11. This assembly is of fused quartz with standard glass taper joints in the lower temperature regions. The nickel-chromium elec- trodes were prepared for this study by using powder metallurgical techniques and high-purity chromijum and nickel. The electrode connections are nickel wires that are joined to tungsten wires sealed through the glass. The alkali halide mixtures are prepared from clear crystals selected from slowly cooled melts of reagent-grade materials. Curves plotted from emf measurements at various temperatures for two alloys containing different proportions of chromium {22.1 and 27.6 mole %) are shown in Fig. 4.12, For these runs the electrolyte was the NaCl-RbCl eutectic containing 0.1 to 0.5 wt % CrCl,. An activity coefficient of about 0.9 at 800°C was determined from these curves, in good agreement with the 1100°C data of Grube and Flad.23 Attempts to utilize electrodes containing 10 mole % or less of chromium showed that the emf obtained was quite unreproducible, - fl « , UNCLASSIFIED ORNL-LR-DWG 12788 TO VACUUM /"/ | | | | TUNGSTEN 41" i e | I /k/—NlCKEL WIRE Ul I 0 I aUARTZ I ’ 6““' ‘ 1 Il LIQUID LEVEL L ~ELECTRODE Y%, TO 1-in. LONG LIQUID LEVEL o /-J Ya-in. DIA {-mm CAPILLARY -~ Fig. 4.11. Cell Vessel with Central Reservoir and Compartments for Four Electrodes. PERIOD ENDING MARCH 10, 1956 The voltages obtained for several cells in which the electrolyte was the LiCl-KCl| eutectic with 1 wt % CrCl, and in which the electrodes were chromium cmd2 Inconel are shown in Fig. 4.13. The line drawn through the points of Fig. 4.13 was calculated for an ideal chromium-nickel alloy containing 15 wt % chromium. Although Inconel nominally contains 5 wt % iron and the points scatter rather badly, the data seem to be represented reasonably well by the ideal case. The electromotive forces obtained should be reproducible after raising and lowering of the temperature and after brief shorting or electrolysis. Electrodes containing less than 10% chromium and the Inconel electrodes, however, responded poorly to the temperature test, All the electrodes re- sponded well to the shorting test. This indicates that a diffusive equilibrium between chromium on the surface and chromium within the body of the electrode is established at the temperatures at which these tests are performed. !t was found that mechanical agitation of the electrodes within the cell had no effect upon the electromotive force, This indicates that there is no sharp concentration gradient in the electrolyte in the immediate vicinity of the electrodes, Reduction of UF, by Structural Metals J. D. Redman Materials Chemistry Division Data obtained from filtration studies of the reduction of UF, by Cr® or Fe® with NaF-ZrF, (50-50 mole %),27 NaF-LiF-KF (11.5-46.5-42 mole %),28 NaF-ZrF , (53-47 mole %),2% or NaF- LiF-ZrF , (22-55-23 mole %)30 as reaction mediums were given in previous reports. More recently, studies have been made on the reduction of UF by Fe® or Cr® with molten NaF-ZrF , (59-41 mole %) 27), D. Redman and C. F. Weaver, ANP Quar. Prog. Rep. June 10, 1954, ORNL-1729, p 50; ANP Quar. Prog. Rep. Sept. 10, 1954, ORNL-1771, p 60. 28) D. Redman and C. F. Weaver, ANP Quar. Prog. Rep. March 10, 1955, ORNL-1864, p 56. 295, D. Redman and C. F. Weaver, ANP Quar. Prog. Rep. June 10, 1955, ORNL-1896, p 60. 30), D. Redman and C. F. Weaver, ANP Quar. Prog. REP. Septo 10, 1955' ORNL"]947, p 74- ANP PROJECT PROGRESS REPORT as the solvent. The results obtained in these studies show that Cr® is not stable with respect to UF, dissolved in these different fluoride mixtures, This observed instability of chromium, either as pure metal or in alloys, has led to studies of the stability of UI:4 toward other structural metals which might serve as substitutes for chromium in alloys. Earlier work31 showed that Mo® was more stable toward UF4 than Cr° was in both NaF-ZrF (50-50 mole %) and NaF-LiF-KF {11.5-46.5-42 mole %). The systems UF -V and UF ,-Nb® also have been studied with the NaF-LiF-KF mixture as 315, D. Redman and C. F. Weaver, ANP Quar. Prog. Rep. March 10, 1955, ORNL-1864, p 61. UNCLASSIFIED ORNL—~LR—DWG 12789 0.10 r . 0.08 A AT T (o) (o) 0 g . o ® 0.06 * veT t ¥ © ) O » 3 Z W Q.04 L J o } ALLOY ELECTRODE CONTAINING 27.6 mole % Cr o o0z A ALLOY ELECTRODE CONTAINING 22.1 mole % Cr (®) DOUBTFUL VALUE O ! 600 700 800 200 1000 TEMPERATURE (°C) Fig. 4.12. EMF Measurements Obtained with a Concentration Cell at 600 to 1000°C for Chromium-Nickel Alloys vs Pure Chromium in Molten NaCl-RbCl Eutectic Containing CiCl,. UNCILASSIFIED ORNL—-LR—DWG 12790 012 4 » 0.08 ¥ / —t— :_u? ¢ IDEAL * 2 Wy 0.04 o L 640 660 680 TO0 720 740 760 780 80C 820 840 TEMPERATURE (°C) Fig. 4.13. EMF Measurements Obtained with a Concentration Cell at 600 to 1000°C for Inconel vs Pure Chromium in Molten LiCI-KC| Eutectic Containing CCl,. 94 the reaction medium. Data for these systems, reported previously,32 show that Nb° is unstable toward UF, in the NoF-LiF-KF solvent. During the quarter, studies have been made on the systems UF ,-Ta® and UF ,-W® with the NaF-LiF-KF eutectic as the reaction medium. Studies also have been carried out on the UF ,-V® and UF ,-Nb°® systems with NoF-ZrF4 (50-50 mole %) and NaF-LiF-Zer (22-55-23 mole %) as the reaction mediums, The results of the studies of the reduction of UF, by Cr°® at 600 and 800°C with NaF-ZrF, 32 p, Redman, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 86. PERIOD ENDING MARCH 10, 1956 (59-41 mole %) as the reaction medium are given in Table 4.10. In these runs approximately 2 g of Cr® was reacted with UF (11.4 wt %, 3.7 mole %) dissolved in opproximcte1y 40 g of the Nc:F-ZrF’1 mixture contained in nickel. The data given in Table 4.10 show considerable scatter; however, it appears that the chromium concentrations are essentially the same at the two temperatures studied. Data in Table 4,11 present a comparison of equilibrium Cr** values for the several mixtures studied. As may be seen, the concentration of chromium fluoride at equilibrium becomes pro- gressively smaller as the ratio of alkali-metal fluoride to ZrF4 is increased, This is in general TABLE 4.10. DATA FOR THE REACTION OF UF, WITH C:° IN MOLTEN NaF-ZrF, (59-41 mole %) AT 600 AND 800°C Conditions of Equilibration Present in Filtrate Temperature Time Total U Total Ce* Total Ni (°C) (hr) (wt %) (Ppm) (pPm) 600 3 8.9 710 65 3 8.7 1010 95 5 9.4 1060 70 5 8.8 970 95 S 8.6 1030 35 800 3 9.9 1000 20 3 8.7 950 45 5 9.6 1190 60 5 10.4 1080 70 *Blank of 270 ppm of Cr at 800°C. TABLE 4.11. EQUILIBRIUM CONCENTRATIONS OF Cr** FOR THE REACTION Ccr® + 2UF, -r-_—‘ZUF3 + CrF, IN VARIOUS SOLVENTS Sclvent Composition Equilibrium Concentration of cett (ppm) (mole %) At 600°C At 800°C NaF-ZrF ,; 50-50 2250 2550 NaF-Z¢F ; 53-47 1700 2100 NaF-ZrF ,; 59-41 975 1050 NaF-LiF-Z¢F ;; 22-55-23 500 650 95 ANP PROJECT PROGRESS REPORT agreement with the concept that UF, is stabilized by the formation of a UF_= or UF,~= complex ion which must compete with ZrF , for the free fluoride ion, As the data in Table 4.12 indicate, however, the concentrations of Fe*t in equilibrium when pure Fe®is used as the reducing agent are approximately the same as those found when either NaF-ZrF (53-47 mole %) or NaF-ZrF, (50-50 mole %) is used as the reaction medium., Temperature ap- pears to have little effect on the reaction, although a slight decrease in iron concentration with increased temperature is indicated. A similar temperature effect occurs when the other two NaF-Zer mixtures are used as solvents, Data for the reaction of UF , with Ta® and with WS at 600 and 800°C with NaF-LiF-KF (11.5-46.5- 42 mole %) as the reaction medium are given in Table 4.13. Approximately 2 g of the metal was equilibrated with UF, (15 wt %, 2.3 mole %) dissolved in approximately 20 g of the NaF-LiF-KF mixture contained in nickel. The data presented in Table 4.13 indicate that neither Ta® nor W° is stable with respect to UF under the conditions studied, and therefore they afford little or no advantage over Cr° The data collected for the UF ,-W® system show considerable TABLE 4,12, DATA FOR THE REACT NUF'Z'F4 (59+41 mole %) scatter, the cause of which is unknown. Since there was some indication that equilibrium was not being attained during the short heating periods, a few experiments were run with 12-hr equilibration periods. The results obtained for these runs were not precise, but they seem to demonstrate that no significant increase in concentration of tungsten occurs after 5 hr of equilibration. Data are given in Table 4.14 for the reaction of UF, with Nb® at 600 and 800°C with NaF-ZrF, (50-50 mole %) and with Nc;:F-LiF--ZrF:4 (22-55-23 mole %) as the reaction mediums, The runs were made in nickel equipment with 11.4 wt % UF dissolved in approximately 40 g of the fluoride mixtures and 2 g of Nb°, These data show that no significant interaction between Nb®and UF, occurs in either of the ZrF ~containing fluoride mixtures studied. This behavior is in marked contrast to that observed when NaF-LiF-KF (11.5-46.5-42 mole %) is used as the reaction medium; with this medium, niobium concentrations of 700 and 1500 ppm were found at 600 and 800°C, respectively, The markedly different niobium concentrations found for the solvents are surprising and must be associated with the differences in free alkali-metal fluoride concentrations present and the resulting fON OF UF, WITH Fe® IN MOLTEN AT 600 AND 800°C Conditions of Equilibration Present in Filirate Temperature Time Total U Total Fe* Total Ni (°C) (hr) (wt %) (ppm) (ppm) 600 3 8.5 510 35 3 8.6 €30 40 5 8.6 370 60 5 8.5 600 70 800 3 8.9 680 70 3 8.6 290 65 3 9.2 460 70 3 9.2 440 75 5 9.3 465 100 5 8.8 280 35 *Blank of 140 ppm of Fe at 800°C. 96 PERIOD ENDING MARCH 10, 1956 TABLE 4.13. DATA FOR THE REACTION OF UF, WITH To® AND WITH W° IN MOLTEN NaF-LiF-KF (11.5-46.5-42 mole %) AT 600 AND 800°C Conditions of Equilibration Present in Filtrate Temperature Time Total U Total Ta* Total Wx=* Total Ni (°C) (hr) (wt %) (ppm) (ppm) (Ppm) For Reaction with Ta® 600 3 11.2 1320 140 3 1.6 1220 20 5 12.1 1150 110 5 11.8 1150 240 800 3 11.2 3220 170 5 11.9 2800 80 5 11.8 3020 180 For Reaction with W° 600 3 11.2 950 215 3 11.2 920 435 5 11.6 1410 250 5 11.2 1350 230 5 10.5 900 5 11.0 1125 800 3 11.3 980 260 3 11.2 920 205 5 11.2 1480 65 5 10.8 2150 250 5 12.0 580 155 5 11.8 1260 165 5 11.6 1210 5 11.8 1600 12 10.5 1880 115 12 11.8 1400 12 11.3 1430 *Blank of approximately 50 ppm of Ta ot 800°C. Several attempts were made to reduce the size of the blank, but all were unsuccessful, **Blank of 350 ppm of W at 800°C. changes in activities of the uranium and niobium fluoride complexes. A previous study of the reaction of UF , with V° in the NaF-KF-LiF eutectic showed vanadium concentrations in the melt to be about 25 ppm. However, repeated checking of this rather surprising result has served to show that the chemical analyses were grossly in error. |t now appears that significant interaction of UF, with vanadium in either NaF-KF-LiF or NaF-ZrF4. oCccurs However, it seems advisable to withhold the data until the inconsistencies in analyses have been clarified. Reaction of UF3 with NaF-KF-LiF Evutectic B. H. Clampitt Materials Chemistry Division Alkali-metal vapors are evoived from a heated solution of UF3 in the molten NaF-KF-LiF eutectic, 97 ANP PROJECT PROGRESS REPORT TABLE 4.14, DATA FOR THE REACTION OF UF4 WITH Nb° AT 600 AND 800°C IN MOLTEN NuF-ZrF4 (50-50 mole %) AND NaF-LiF-ZrF, (22-55-23 mole %) Conditions of Equilibration Present in Filtrate Temperature Time Total U Total Nb* Total Ni (°c) (hr) (wt %) (pPm) (ppm) For Reaction Medium Nc:F-ZrF4 (5050 mole %) 600 3 8.2 25 55 3 7.9 20 60 5 8.9 25 55 5 8.7 25 80 800 3 8.8 20 40 3 8.8 20 70 5 8.8 20 80 5 8.5 20 85 For Reaction Medium NuF—LiF-ZrF4 (22-55-23 mole %) 600 3 8.2 1 100 3 8.8 1 120 5 8.4 2 120 5 8.9 1 45 800 3 8.5 1 110 3 8.6 1 110 5 8.9 5 15 5 8.3 4 40 *Blank of 25 ppm of Nb at 800°C in NaF-ZrF ,; approximately 15 ppm in NaF-LiF-ZrF ,. despite the standard free-energy changes for the reaction KF + UF, —>UF, + K° being unfaverable by about 30 kcal per gram atom of potassium. The vapor is about 95% K° with the baflance almost entirely Na® but, for purposes of the following discussion, it is assumed that KF is the only component of the melt reduced by UF .. That this reaction should proceed perceptibly is additional evidence that UF, is strongly complexed and stabilized in the molten fluoride solution; this stabilization, which in general makes UF |, solutions less corrosive than they would otherwise be, 98 contributes, unfortunately, to instability of UF, in such systems, ' The kinetics of conversion of UF3 to UF , in this solvent are complicated by the disproporticnation reaction 4UF ,—>3UF, + U which takes place simultaneously with the redox reaction and produces an alloy of uranium metal with the container wall, A rough study of the kinetics of attrition of UF, in this solvent has been made by analysis for UF, of samples taken after 2 hr of heating of standard mixtures in open crucibles of copper under helium atmospheres at varying pressure, A previous report33 showed that at 650°C, under these conditions, the rate of reaction (measured by the UF ; remaining after 2 hr) was independent of helium pressure at pressures above 50 mm Hg; at pressures below 50 mm Hg the amount of UF, remaining varied nearly linearly with pressure, The results obtained at 750°C, shown in Fig. 4.14, are quite similar; at this temperature the reaction rate is insensitive to pressures above 130 mm Hg. For samples which were held at pressures above this “‘critical’’ pressure and for which 21.4 wt % of UF ; was initially present, 58% of the UF; remained after 2 hr at 650°C and 47% remained after a similar period at 750°C. A ORNL-LR-DWG 12791 | 10 s : J ./.’ ! ® a0 UF3 REMAINING AFTER 2 HOURS (wf %) 0 . b 0 100 200 300 400 50C 600 700 800 HELIUM PRESSURE (mm Hg} Fig. 4.14. Oxidation of UF; at 750°C in NaF- LiF-KF (11.5-46.5-42 mole %). This behavior suggests strongly that the sharp break in the reaction rate as a function of pressure occurs at the ‘‘bubble-point’’ pressure for the solution at a given temperature, At pressures above this level- the evolution of potassium ap- pears to reach a near-equilibrium steady state and to proceed at a rate controlled by the rate of vaporization of potassium from its solution in the salt. If the values 50 and 130 mm Hg are taken as the bubble pressure of K° from the solution at 650 and 750°C, respectively, then, since the vapor pressures of pure K° at those temperatures are 240 and 750 mm Hg, the activity of the 33g, H, Clampitt and C. J. Barton, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 90. PERIOD ENDING MARCH 10, 1956 potassium is approximately 0.2 at each temper- ature, (The absence of a brilliant blue-green phase seems to be sufficient assurance that the solution is not saturated.) The effect of temperature on the two competing reactions cannot be specified with certainty, The disproportionation reaction should be rapid, initially, ot all temperatures, but it would be expected to slow to a rate controlled by the slow diffusion of U° into the copper wall, In each 2-hr test, therefore, an almost constant amount of UF might be expected to form, regardless of temper- ature, If the reaction to produce K° vapor is assumed to be more rapid, by a ratic of 130/50, at the higher temperature and if the disproportionation rate is assumed to be nearly independent of temperature, then, by using the experimental data for reactions above the ‘‘critical'’ pressure, a set of three simultaneous equations in three unknowns can be formulated and solved, The calculations show that, if these assumptions are correct, about 7 and 18% of the original UF, was oxidized by KF at 650 and 750°C, respectively, and 35% dispro- portionated at each temperature, The 35% lost by disproportionation is con- siderably higher than the 5 to 10% shown experi- mentally for similar conditions in the NaF-LiF eutectic, This discrepancy may well be associated with the fact that a deposited mirror of uranium alloy is always noticed when KF is present but not when NaF-LiF alone is used; potassium seems to ‘““catalyze’’ the disproportionation in some fashion, For the steady-state condition, the apparent heat of activation for the evolution of potassium is 18 kcal if the rates are considered to be proportional to the steady-state potassium pressure at the two temperatures. This is in good agreement with the 20-kcal heat of vaporization of potassium, particu- larly since the 18-kcal figure applies to solutions in which the activity of potassium is only 0.2, It is interesting to note that, since the UF -to-UI:4 ratio does not change markedly with temperature and the near-equilibrium steady-state activity of potassium does not change from the value 0.2, the equilibrium constant for the reaction is nearly independent of temperature. This means that the heat of reaction is very small, in spite of the large positive standard free energy for the reaction, Evidently, a large negative heat of complexing is 99 ANP PROJECT PROGRESS REPORT involved, and the reaction is better represented by the equation: (X + DKF + UF,—> K UF,, + K° . On the assumption that the activity coefficients of KF and UF, are of the order of unity, the relation ayE Ao o 4 K = o=AF°/RT _ a a UF, “KF ()’XUF )(0-2) -4 _qp-1s (0.02)(0.4) yields about 10-7 for the activity coefficient of UF, (the concentration X is approximately 0.02) in the NaF-KF-LiF eutectic at 1000°K. There are as yet no data which permit an independent estimate of the activity coefficient of UF, in this system, This value seems to be surprisingly small, since in the NaF-ZrF solutions (in which UF, must compete with ZrF, for the complexing fluoride ions) the activity coefficient of UF, from the reaction 2UF4 + Fec’:Fer + 2UF3 is about 10-1, On the basis of these results, the NaF-LiF-KF type of melts, for which the beneficial effects of UF, are most needed, are the melts with which UF, is least compatible, This is a consequence of the extremely small activity coefficient for UF, in the presence of the large excess of fluoride ions and the correspondingly high activity of U° required by the disproportionation equilibration to balance the low UF , activity, Experimental Preparation of Fluorides B. J. Sturm Materials Chemistry Division The preparation of several structural-metal fluorides was accomplished during the quarter. Additional Crl':3 and FeF, were prepared by hydrofluorination of anhydrous CrC[3 and FeCI2 at 600°C. Substantial quantities of CeF"3 and Lc:l:3 were synthesized, and additional work was done on complexes of ZrF fluorides, with several metal The identities and the purities of all the preparations were established through chemical, x-ray, ond petrogrophic examinations, 100 The preparation of CeF . was accomplished by treating aqueous solutions of either CeCl3 or C:e(NC):,’)3 with aqueous HF, washing the precipi- tated CeF , thoroughly with water by centrifugation, and drying at 150°C. The compound LaF, was prepared in a similar manner by using a solution of LdCl3 obtained by dissolving La,0,. A small quantity of CUF2 was prepared by dehydration of CuF ,-2H,0 with HF at 600°C. The compounds NaCrF, and NaNiF, were prepared by fusing the proper quantities of NaF and structural-metal fluorides in sealed nickel capsules, Data were presented previously34 for the compounds CrFQ-ZrF4, FeF,-ZrF ,, and NiF ,-ZrF , which weee.prepared by heating Zr 4 With equimolar quantities of the structural-metal fluorides ot 950°C. The instability of CrF,.ZrF, and FeF,-ZrF, was found to be due to atmospheric contamination; if stored properly, these compounds are, apparently, stable indefinitely. A compound ZnF _.ZrF , with x-ray properties very similar to those of Nin-ZrF4 was prepared by heating equimolar quantities of ZnF2 and ZrF4 at 950°C in sealed nickel capsules. An equimolar mixture of AIF_ and ZiF , was heated to 950°C in a sealed nickel capsule in an attempt to prepare A|F3-ZrF4. Examination of the product revealed that no reaction occurred; only A|F3 and ZrF , were observed. An attempt to prepare I'_'ei::,’-ZrF4 by heating equimolar quantities of Fer and ZrF , in a nickel capsule lined with silver gave material which x-ray examination revealed to be FeF _.ZrF, and an unknown phase, Chemical analysis szhoweé that this material contained equivalent quantities of Ag* and Fe** and, thus, that Fe3* had been reduced by Ag® The replacement of the silver liner by a gold liner gave a material that contained small amounts of FeF_ and ZrF, but which consisted primarily of a phase having essentially the FeF,.ZrF, structure but a smaller cell size. It is believed that this phase is FeF .Z¢F . It is interesting to note that decomposition of FeF -ZrF on exposure to the atmosphere yields a ma’reriof with a cell size intermediate between that of FeF ,-ZrF , and the presumed FeF,.ZrF,. 348. J. Sturm, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 91. Color Studies of Alkali Metals in Equilibrium with Alkali Halides at High Temperatures B. H. Clampitt Materials Chemistry Division Davy335 and Bunsen and Kirshhoff3¢ observed long ago that when a molten alkali halide was subjected to electrolysis a highly colored phase appeared near the cathode. This color has been explained as being due to colloidal metal fogs37.38 or to subhalide formation. Bredig and his co- workers39:40 and Cubicciotti (reference 41, for example), who have made extensive studies of metal salt systems, have investigated metallic systems in which no colors were produced. Corbett and von Winbush4? have studied the solubility of metals such as cadmium and antimony in their chlorides and have found relatively slight color changes in the salts as the metals were dissolved, In general, it appears that color changes in molten salts have been often observed but that colors in discrete metal phases have not been described, Recent experiments in this laboratory have shown that, in general, alkali-metal phases in contact with alkali-metal salts are highly and strikingly colored, The experiments were conducted in a dry box which could be evacuated and subsequently filled with an inert atmosphere. A 2-in. steel pipe welded through the bottom of the box and sealed at the projecting end served to provide, when fitted with a surrounding tube furnace, a heated zone intregal with the box, Crucibles of nickel were used for the containing vessels., The salts and metals were of reagent grade, Fluorides and chlorides were fused prior to use; bromides were 3SH. Davy, Trans. Roy. Soc. {Londonr) 97, 1 (18C7). 36R. Bunsen and G. Kirshhoff, Pogg. Ann. 113, 364 (1861). 37R. Lorenz and W. Eite!, Pyrosole, Lupzig, 1926. 38, Mollwe, Akad, Wissensch. Gottingen Nachrichten Math. Physikal. Klasse., Fachgruppe 1l, Neue Folge 1, Neo, 18, 203-207 (1935). M. A. Bredig, J. W. Johnson, and Wm. T. Smith, Jr., J. Am. Chem. Soc. 77, 307 (1955). 40M. A. Bredig, H. R. Bronstein, and Wm. T. Smith, Jr., J. Am. Chem, Soc. 77, 1454 (1955). 41p, Cubicciotti, J. Am. Chem. Soc. 74, 1198 (1952). 42y D, Corbett and $. von Winbush, J. Am. Chem. Soc., 77, 3964 (1955). PERIOD ENDING MARCH 10, 1956 given no previous treatment., The alkali metals were washed in the dry box with cyclohexane to remove the protecting mineral oil, and the cyclo- hexane was removed by evaporation, The colors observed when a small amount of metal is added to a molten salt mixture are listed in Table 4.15, and the colors observed when a small quantity of salt is added to a relatively large quantity of liquid metal are listed in Table 4.16. Colors have been observed with every combination tested, except lithium metal with a lithium salt, Repeated tests have shown that the colors are much more dependent on temperature than on composition of the system, At 400°C the mercury-like sheen of the alkali metal begins to develop a green tinge, At 500°C, the green changes to turquoise, Above 600°C the color becomes increasingly deep blue and eventually changes to violet, The color produced is inde- pendent of relative quantities of metal and salt, The color is produced whether or not the temper- ature is sufficiently high to melt the salt; molten metal is colored by solid salt, Since the metal phase is sometimes on the top and sometimes on the bottom, because of differ- ences in interfacial tensions, it is difficult to determine with certainty whether the salt or the metal is colored. There are several reasons, however, for the belief that the color is in the metal phase. A definite quantity of salt, too small to measure conveniently, can be added to a large quantity of molten sodium without color production; but the addition of one more grain may bring the color in full intensity. The color production is independent of whether or not the salt that can be added without producing the color is so small that it cannot be detected with certainty. If the NaF-KF-LiF eutectic, for example, is contained at 600°C in a crucible with a partition that extends to within ¥ in. of the bottom and sodium metal is added to the salt on one side, the material on that side turns blue immediately; the material on the other side becomes dark but shows no blue color, When NaCl is added to one compartment of a similar crucible containing sodium metal at 600°C, the contents of both compartments turn blue. Since NaC!l sinks in molten sodium, however, this last observation is not necessarily evidence for the color being in the metal phase. Bredig's3? results show a decrease in solubility of the salt in the metal as the cation size becomes 101 ANP PROJECT PROGRESS REPORT TABLE 4,15. COLORS OBSERVED WHEN SMALL AMOUNTS OF METAL ARE ADDED TO LARGE AMOUNTS OF MOLTEN SALT Composition of Salt Phase Experimental Temperature* Position of Metal Alkali Metal o Color (mole %) (TC) Phase NaF-LiF-KF; 11.5-46.5-42 Li 500 Top Turquoise Na 500 Top Turquoise K 500 Top Turquoise NaF-LiF; 40-60 Na 670 Top Blue Li 670 Top Blue LiF Li 845 Top * % L.iCl Li 650 Bottom Metallic LiCl-NaCl; 59-41 Na 600 Bottom Blue Li 600 Bottom Blue LiCl-LiF; 80-20 Li 600 Bottom Metallic LiCl-LiBr; 25-75 Li 600 Bottom Metallic *Determined by vapor pressure of alkali metal and melting point of the salt. **Color not defined because furnace was at red heat, TABLE 4.16. COLORS OBSERVED WHEN SMALL AMOUNTS OF SOLID SALT ARE ADDED TO LARGE QUANTITIES OF METAL Experimental Temperature Alkali Metal Salt Phase (°C) Color Li LiF 600 Metallic LiCl 600 Metallic LiBr 600 Metallic NaF 600 Blue NaCl 600 Blue KF 600 Blue KClI 600 Blue Na LiF 600 Blue NaF 600 Blue NaCl 600 Blue K LiF 500 Green LiCl 500 Green NaF 500 Green KF 500 Green KCI 500 Green 102 smaller. It is possible, therefore, that lithium halides are too insoluble in metallic lithium at 700°C or below for the colors to be observed in such systems, The salt phases do not remain transparently clear in these experiments. The salt phase darkens in all cases but does not, apparently, show true colors, Darkening under these conditions might reasonably be related to F-center production in the molten alkali halides. Bredig called attention to a study of this problem by Mollwo,43 who measured absorption spectra of molten alkali halides which had been exposed to alkali-metal vapor. Mollwo found the abserption spectra to consist of broad bands, without structure, and to have maximums at, for example, 790 my for sodium salts. No relationship with the band maximums of the F-centers was found, and, for a given metal, there was no dependence on the anion used. It appears to be very likely that the dark melts observed in the present work resemble those which gave the broad absorption spectra reported by Mollwo and that they represent ““loose’’ F-centers, that is, loosely bound electrons which can absorb a broad range of energies, This in in contrast to the lattice-locked F-centers in solid alkali halides, for which the frequency of the absorption band is related to the lattice constant, d, by the equation ved? = 5.02 x 10=3 m2/sec Also, the blue-green ‘‘metal fogs'' or ‘pyrosols,’’ mentioned by Mollwo and others as observable during electrolyses of alkali halides, would now be interpreted, according to the results obtained in the present work, as being due to a colored metal phase, PHYSICAL PROPERTIES OF MOLTEN MATERIALS F. F. Blankenship G. M. Watson Materials Chemistry Division Vapor Pressures in the System KF-ZrF S. Cantor Materials Chemistry Division The vapor pressures of various compositions in the system KF-ZrF , are being measured by the 43g, Mollwo, Akad. Wissensch, Gottingen Nachrichten Math. Physikal. Klasse., Fachgruppe II, Neue Folge 1, No. 18, 203-207 (1935). PERIOD ENDING MARCH 10, 1956 Rodebush-Dixon method44 previously used in these laboratories. 4> The method depends on the determination of the pressure of inert gas in the limbs of the vapor-pressure cell at which the passage of the gas through the cell is blocked by the vapor. This pressure is determined by the change of levels in a differential manometer connected to the two limbs of the cell, Calcu- lations that involve the interdiffusion of the vapor and the gas at the temperatures of interest indicate that 5 to 10 mm Hg represents the lower limit of accurate determination of vapor pressure, Calibrated platinum—platinum-rhodium thermo- couples were used in conjunction with a Leeds & Northrup portable precision potentiometer (model 8662) to measure temperatures, The apparatus was tested by determining the vapor pressure of ZrF4 and comparing the value obtained with that obtained by Moore46 with a similar apparatus and with values obtained at Battelle Memorial Institute (BMI} by the transpiration method.47 The vapor- pressure equations obtained from these three investigations are shown in Table 4.17. [t was assumed that the pressure measured was due to ZrF, only for the region studied in the KF-ZrF , system, This assumption appeared to be justified, since the equilibrium vapor pressure of pure KF is quite small at the temperatures studied and petrographic and x-ray analysis indicated essentially pure ZrF4 in the sublimates obtained. The results of these measurements are compiled in Table 4.18 in terms of the constants A and B of the vapor-pressure equation, B fog P (mm) = A - TCK) The constants were obtained by a least-squares treatment of the data., The heat of vaperization, which was obtained from the quantity B, appears to decrease with increases in the KF fraction of the mixture, while the low volatility of ZrF4 from the mixtures suggests that the heat of vaporization 44w, H. Rodebush and A. L. Dixon, Phys. Reuv. 26, 851 (1925). 45R. E. Moore and C. J. Barton, ANP Quar. Prog, Rep. Sept, 10, 1951, ORNL-1154, p 136, 46p, E. Moore, ANP Quar. Prog. Rep. Sept. 10, 1952, ORNL-1375, p 147. 47, A, Sense, M. J. Snyder, and R. B. Filbert, Jr., J. Phbys. Chem. 58, 995 (1954). 103 ANP PROJECT PROGRESS REPORT TABLE 4.17. YAPOR-PRESSURE EQUATIONS FOR ZrF4 OBTAINED BY VARIOUS INVESTIGATORS Calculated Pressure (mm Hg) Investigator Vapor-Pressure Equation o= P = At 1000°K At 1100°K 11,320 Moore log P = 12.46 — 13.8 145 12,376 BMI log P = 1340 — — 10.5 141 11,506 This study log P = 12,70 — — 15.6 174 TABLE 4,18. SUMMARY OF CONSTANTS OBTAINED FROM VAPOR-PRESSURE MEASUREMENTS OF THE SYSTEM KF-ZrF4 Cemposition (mole %) Constants Heat of Vaporization, H (keal) Zrl""4 KF A B 84.50 15.50 11.88 10,820 49.5 69.97 30.03 11.63 10,720 49.0 65.87 3413 9.754 8,881 40.6 . 60.07 39.93 7.635 6,806 31.1 49.94 50.06 8.211 7,925 36.2 - could be expected to increase, A further investi- oy UNCLASSIFIED gation is being carried out to ascertain whether the 150 — i ‘ | . data are subject to some source of systematic ' L error which might be responsible for the apparent - ':LHO'(?R?SU[;ZTA o NGFlz;; [ discrepancy., For this purpose an independent IE Vo o BMI DATA ON Nof -Zrf, 4 . i static method of measuring vapor pressures will be = A iy used, % %0 i LT 7 a RAOULT’S LAW FOR | The B800°C isotherms of vapor pressure vs £ 70 |- UNDISSOCIATED ""QU--'-Q—S“‘\}A;'----NGF*--Z‘& . - | composition for the NoF-ZrF‘11 and KF-ZrF, 2 o L . //l W /| systems are shown in Fig. 4.15. The vapor pressure 2 | s Law //.\\_K}_ZrFq for pure liquid ZrF was calculated from the 20 '__._.;__fifgifpopzrw 215 34 /] sublimation pressure and the heat of fusion /r,//! | /fi“/ ‘ | (9 kcal/mole) as determined from the loweringof 10 s . M/fif ‘ the freezing point of ZrF , by the addition of NaCl. NeF 10 20 30 40 50 60 70 80 90 ZrF, The results indicate clearly that lower vapor kF Z¢F, (mole %) . pressures can be obtained in the KF- ZrF system than in the NaF- ZrF system, presumably because Fig. 4.15. Changes in Vapor Pressure with KF furnishes a greater activity of fluoride ion as a result of the greater radius of the potassium ion, 104 at 800°C Composition for the Systems NaF- ZeF, ZiF, and KF- Vapor Pressure of FeCl, C. C. Beusman Oak Ridge Institute of Nuclear Studies Measurements of the activity of ferrous chloride in melts of the alkali-metal chlorides have been started with the use of vapor-pressure methods, A transportation apparatus has been assembled and will be used in the temperature ranges where both components are volatile, For lower temper- atures, the Rodebush-Dixon method44 will be used. For calibration purposes, the vapor pressure of pure ferrous chloride has been measured in the Rodebush-Dixon apparatus. The data are in fair agreement with literature values and are well represented by the equation 8,985 log P (atm) = 23,080 — TR Surface Tensions of Molten Salts — 5.234 log T (°K) . S. Langer Materials Chemistry Division Surface tensions of molten salts have been examined by a number of investigators48~33 who have documented their findings in the open liter- ature; a determination of surface tension of a NaF-ZrF -UF, fuel mixture has been made by Cohen and Jones®4 of ORNL. However, no systematic study of surface tension as a function 48y, k. Boardman, A. R. Palmer, and E. Heymann, Trans, Faraday Soc. 51, 277 (1955). 49y, |. Semenchenko and L. P. Shekhobalova, J. Phys. Chem. (USSR) 21, 613, 707, 1387 (1947). AEC trans. S0F, M. Jaeger, Optical Activity and High Temperature Measurements, p 235-307, McGraw-Hill, New York, 1930. 51g. s. Ellefson and N. W, Taylor, J. Am. Ceram. Soc. 21, 193, 205 (1938). 52¢. F, Baes, Jr., and H. H. Kellogg, J. Metals 5, 643 (1953). 53y, D. Kingery et al., Development of Metal-Ceramic Compositions Suitable for Service at Elevated Tempera- tures, NEPA-1170 (Sept. 1949); NEPA-1234 (Nov. 1949); NEPA-1848 (April 30, 1951); and F. H. Norton et al., Study of Metal-Ceramic Interactions at Elevated Temper atures, NY0-3136 (Oct. 1, 1951); NYO-3137 (Jan. 1, 1952); NYO-3139 (April 1, 1952); NYO-3142 (Jan. 1, 1953); NYO-3143 (Feb. 1, 1953); NYO-3144 (Feb. 1, 1953); NYO-3145 (April 1, 1953); NY0-6290 (Oct. 1, 1953); NYO-6291 (July 1, 1953); NYO-6294 (April 1, 1954). 545, |. Cohen and T. N. Jones, Preliminary Surface Tension Measurements of the ARE Fuel (Fluoride Mixture No. 30), ORNL CF-53-3-259 (March 27, 1953}, PERIOD ENDING MARCH 10, 1956 of composition, purity, etc., has yet been made for materials of interest to the ANP program. Since surface tension has been shown to be important, in certain cases, to heat-transfer>> and corrosion>6 processes, a study of this property is under way. The method chosen for this study is the sessile- drop technique, which has been successfully employed by Ellefson and Taylor,”7 Baes and Kellogg,52 and Kingery et al.33 This method was selected in preference to the maximum-bubble- pressure technique because of the uncertainty of the maximum-bubble-pressure method with liquids having contact angles greater than 90 deg (measured through the liquid). A profile drawing of a sessile drop is shown in Fig, 4.16. A photograph is taken of the sessile drop on a supporting plaque, and the dimensions 2x, the maximum diameter of the drop, and z, the height from the maximum diameter to the apex, are measured. These dimensions, along with the density of the molten salt, are sufficient to determine the surface tension, The contact angle 8, which is difficult to measure directly, can be calculated if the additional quantities x“ and z° are known, UNCLASSIFIED ORNL—LR--DWG 127933 Fig. 4.16. Profile of a Sessile Drop. The equipment required for this study has been assembled and is in working order, Preliminary experiments are being carried out with NaF-ZrF (53-47 mole %), and nickel, Inconel, platinum, an graphite are being used as supporting plaques, The condition of the surface of the supporting plaque is known to be extremely important in 55w, k. Stromquist and R, M. Boarts, Effect of Wetting on Heat Transfer Characteristics of Liquid Metals, ORO-60 (Jan. 31, 1952). 36, W. Taylor, J. Nuclear Energy 2, 15 (1955). 57B. 5. Ellefson and N. W. Taylor, J. Am. Ceram. Soc. 21, 193, 205 (1938). 105 ANP PROJECT PROGRESS REPORT determining whether the molten drop will wet the plaque, and the effects of various atmospheres on the wetting properties are also being studied. For the sessile-drop method to be successful, the molten drop must not wet the supporting surface, The preliminary experiments show that nickel and Inconel surfaces are not immediately wetted by the sessile drop when it melts, even if the surfaces have previously been hydrogen-fired, The contact angle is only slightly greater than 90 deg, and, on standing, the contact angle slowly decreases until the surface is wetted by the salt, whether the atmosphere is helium or a vacuum of about 2 4. A clean, polished, degreased platinum surface is wetted by the drop immediately upon melting in both vacuum and helium, However, if a nickel plaque that has been previously hydrogen-fired is used as the supporting surface and a hydrogen atmosphere is maintained in the furnace tube so that any thin film of oxide which has formed since the hydrogen-firing process will be reduced, the sessile drop will wet the surface as soon as it starts to melt, The droplet will spread over the nickel surface by the time it is entirely molten, It oppears from these preliminary experiments that truly clean hydrogen-fired nickel surfaces are completely wetted by molten Nc:ZrF5 and that, if thin oxide films are aliowed to form on the metal surface after hydrogen-firing, the films will prevent wetting until they are dissolved by the molten salt, These observations are difficult to reconcile with the lack of evidence of wetting in nickel purifi- cation vessels in which intensive hydrogen treatments of molten fluorides are carried out for hours at 800°C, and therefore a more thorough study of the behavior of the sessile drop is being made. Ellefson and Taylor57 measured the surface tension of LiF by the sessile-drop technique with ‘*electrode’’ graphite as the supporting surface, Their results are in good agreement with those of Jaeger,?0 which were obtained by the maximum- bubble-pressure technique, Thus it can be assumed that at least some fluoride melts will not wet graphite, Experiments with the use of graphite surfaces are under way, 58G. M. Watson and F. W. Miles, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 96. 5%, D. Harkins, Physical Methods of Organic Chem- istry, vol. |, Part 1ll, 2d ed., chap. IX, p 355 (ed. by A. Weissberger), Interscience, New York, 1949. 106 Density and Surface Tension of Molten KCI ot 800°C F. W. Miles Materials Chemistry Division Apparatus 38 added to the conventional equipment for fuel purification is being used for the study of variations of density and surface tension of molten salts with changes in chemical composition, Calibration experiments with molten-salt standards are under way, and measurements were made of the surface tension and the density of molten KCl to determine the accuracy and precision of the methods, Both the density and the surface tension of a molten salt can be determined from the same set of measurements of the maximum pressures required to blow helium bubbles into the melt through a capillary orifice at various accurately known depths. This is the standard method for surface- tension measurement and has been frequently described in the literature, The procedure being used at present was obtained from various sources,”?-61 and the design of the capillary orifices was substantially guided by Porter's analysis,82 For the calculations Sugden’s method of correction®9:63 for bubble distortion, as modified by Tripp and Young®4 for single bubbles, was used. The density and the surface tension of molten KCl at 800°C were determined with three different orifices (two with the same radius) for purposes of calibration. Measurements were made with each orifice at five different depths. The depths used were approximately ‘4, ]/2, Y, ]?32, and 2 in, The results are summarized in TaABIe 4.19. The second determination with orifice No., 1 was purposely performed without recalibration of the tip aofter exposure to the atmosphere following immersion in KCl, The nickel tube carrying 60g, Sugden, The Parachor and Valancy, p 209, Routledge, LLondon, 1930. 61, w. Taylor, Solid Metal-Liquid Metal Interaction Studies. Part I: The Surface Tension of Sodium, AJE.R.E, Report M/R 1247 (Sept. 1953). $2A. W. Porter, Phil. Mag. 9, 7th Series, 1065 (1930). 633, Sugden, J. Chem. Soc. (London) 121, 858 (1922). 64y, Pp. Tripp with T. F. Young, Maximum Bubble Pressure Method for Measurement of Surface Tension, Ph.D. thesis, University of Chicago, 1934. Many help- ful suggestions regarding the experimental assembly grnd procedure were received from Professors Young and ripp. PERIOD ENDING MARCH 10, 1956 TABLE 4.19., DENSITY AND SURFACE TENSION OF KCI AT 800°C Orifice Radius* Observed Density Observed Surface Tension Oritice No. (em) (g/ml) (dynes/cm) 1 0.0729 1.516 95.2 1 0.0729 1.523 92.9 2 0.0729 1.501 91.9 3 0.0430 1.510 93.8 Av 1.513 * 0.007 Av 93.5 * 1.1 *Orifice radius obtained by calibration with water for which the surface tension was simultaneocusly determined with a ring tensiometer, orifice No. 2 was found to be slightly bowed after the experiment; this may account for the lower surface-tension values found with this orifice, Previously published values for these properties of KCl are 1.509 g/cm3 and 95.8 dynes/cm by Jaeger65 and 1.5102 g/em3 by Van Artsdalen and Jaffe. 66 Although the precision and accuracy indicated above are perhaps acceptable, it is hoped that further improvement can be obtained without undue difficulty, Additional tests with KC| and with KCl-NaCl mixtures will be completed before measurements on fluoride mixtures are attempted, PHYSICAL CHEMISTRY OF FUSED SALTS®? E. R. Van Artsdalen Chemistry Division It appears to be established that, in the solid, Cc:tCl2 forms 1:1 complexes with KCl, RbCIi, and CsCl. It has been claimed6® that viscosity measurements of the liquid mixtures indicate that these complexes exist in the molten state. Freezing-point depression studies were made in fused sodium nitrate to test whether any such complexes are stable in the dilute range, Mixtures of CsCl and CaCl, depressed the freezing point in 65F. M. Jaeger, Z. anorg. u. allgem. Chem 101, 185 (1917). 66E. R. Van Artsdalen and I. S. Yaffe, J. Phys. Chem. 59, 118 (1955). 67 Details of this work will be published in separate reports and articles from the ORNL Chemistry Division. 68y, L. Strelets, V, N. Zhludneva, and |. L. Reznikoyv, Zhur. Priklad, Khim, 28, 643 (1955). a strictly additive, ideal manner, Therefore the evidence indicates that no CsCl-CaCl, complexes exist in the dilute range when dissolved in sodium nitrate at about 300°C. Density and preliminary conductance data have been obtained for several molten rare-earth halides. These liquid salts tend to attack quartz rather severely, and, as a result, the conductance measurements show poorer precision than that previously obtained for alkali halides. Some data for LuuC[3 and LaBr, are given in Table 4.20. Density and conductance of several fused alkaline-earth halides and their mixtures with some alkali halides are being investigated, It is hoped that information obtained from this study will throw light on the question of the existence of complexes in these liquid-salt mixtures and, therefore, be complementary to the freezing-point studies of dilute mixtures, PRODUCTION OF PURIFIED FUEL MIXTURES G. J. Nessle G. M. Watson Materials Chemistry Division Use of Copper Maximum Hot-Leg Attack Found in Fig. 5.7. Hastelloy B Thermal-ConvectionLoop 816 Operated with NoF-KF.LiF-UF, (11.2.41-45.3-2,5 mole %) as the Circulated Fluid. 250X. Reduced 32%. ooy with caption) attack in loop 816 was about the same as in the other two loops, but the pits were more concen- trated. The cold legs of these loops were attacked to the same degree as the hot legs, and they ap- peared to be completely free of deposits. Chemical analyses of the fuel mixtures used in these loops are presented in Table 5.3. Effect of Condition of Inner Sutface of Hastelloy B Tubing on Depth of Attack Six thermal-convection loops were constructed of Hastelloy B tubing that had been reamed to ensure uniformity of the inner surface composition and to eliminate surface roughness. Three of the loops were then operated for 500, 1000, and 1500 hr, respectively, with NaF-ZrF ,-UF , (50-46-4 mole %) at a maximum temperature of 1500°F, and three were operated with sodium under the same condi- PERIOD ENDING MARCH 10, 1956 tions. The test results are presented in Table 5.4. The loops operated with the fluoride mixture showed only slight variations in attack as a func- tion of operating time. In comparison with previ- ously operated Hastelloy B loops constructed of as-received or unreamed tubing, the depths of attack were similar. However, the metal crystals found in some of the earlier loops were completely absent in these later tests, The three loops operated with sodium also showed no substantial increase in depth of attack TABLE 5,3, CHEMICAL ANALYSES OF FUEL MIXTURES CIRCULATED IN HASTELLOY B LOOPS 176, 814, AND 816 Total Uranium Content Impurities Found (ppm) Loop No. Sample Taken (wt %) Ni Cr* Fe Mo 176 Before filling 11.7 70 20 75 After draining 11.5 85 245 360 814 Before filling 11.9 130 80 120 After draining 11.9 20 95 120 75 816 Before filling 12.1 150 95 120 After draining 12.2 145 175 125 140 *Chromium is present as an impurity both in the Hastelloy B and in the fuel mixture, which is produced in Inconel pots. TABLE 5.4, RESULTS OF TESTS OF THERMAL-CONVECTION LOOPS CONSTRUCTED OF REAMED HASTELLOY B TUBING Maximum fluid temperature: 1500°F ing Ti i Attack Metallurgical Results Loop Circulated Fluid Operof(ll:lg; 'me Mamnzun.'] ) fac T mits Hot-l.eg Appearance Cold-l_eg Appearance 766 Sodium 500 0.5 Light surface attack No deposits, some attack 767 Sodium 1000 0.5 Light surface attack No deposits or attack 768 Sodium 1500 2.5 Intergranular attack with No deposits or attack surface pits 769 NoF-Zer-UF4* 500 1.5 Surface pits No deposits or attack 770 NoF-ZrF4-UF4* 1000 3 Intergranular attack with No deposits or attack surface pits 771 NuF-ZrF4-UF4* 1500 2 Surface pits Moderate intergranular attack to o depth of 2 mils *Composition: 50-46-4 mole %. 119 ANP PROJECT PROGRESS REPORT with increased operating time., There were no metallic deposits, and there was very little attack in the cold legs of the loops. However, some metallic deposits were observed macroscopically in the traps below the cold legs of all the loops. Screening Tests of Special Fuel Mixtures A series of standard Inconel thermal-convection loops were operated for 500 hr at a maximum fluid temperature of 1500°F to obtain preliminary data for evaluating several special fuel mixtures. The preparation of the special fuel mixtures is de- scribed in Sec. 4, ''Chemistry of Reactor Ma- TABLE 5.5. Since the quantities of the fuel mixtures available for filling the loops were limited, the usual precleaning with the mixture being investi- gated was omitted, Four loops were operated as the first of a series for establishing the level of ZrF, below which mixtures of NaF-LiF-ZrF ,.UF ; exhibit corrosion behavior typical of alkali-metal fluorides rather than zirconium-base fluorides such as NaF-ZrF - UF (50-46-4 mole %). The results of metallo- graphic examination of the loops and chemical analyses of the fuel circulated are presented in Table 5.5. The attack by NaF-LiF-ZrF ,-UF, terials," RESULTS OF INCONEL THERMAL-CONVECTION LOOP TESTS OF SPECIAL FUEL MIXTURES Operating period: 500 hr Maximum fluid temperature: 1500°F Fuel-Mixture Maximum Depth Impurities in Fuel Fuel Sample Total Uranium Loep No. {ppm) Composition (meole %) of Attack (mils) Taken Content (wt %) Ni Cr Fe 841 Na F-LiF-ZrF4-UF4; 14 Before filling 13.2 100 70 265 20-55-21-4 After draining 13.0 30 220 100 842 Na F-LiF-Zer-UF“; 12.5 Before filling 12.8 405 195 170 20-55-21-4 After draining 13.0 30 370 90 843 NoF-LiF-ZrF4-UF4; 10 Before filling 16.2 335 105 130 53-35-8-4 After draining 16.5 80 500 95 844 NaF-LiF-ZrF4-UF4; 15 Before filling 16.5 300 105 115 53-35-8-4 After draining 16.5 85 300 90 839 RbF-ZrFA-UFA; 9 Before filling 6.90 380 240 235 50-46-4 After draining 6.10 9 280 70 840 RbF-ZrF ,-UF ,; 9 Before filling 6.87 370 235 275 50-46-4 After draining 6.40 30 145 70 845 KF-ZrF4-UF4,' 8 Before filling 8.21 110 110 315 50-46-4 After draining 8.36 20 270 30 846 KF-ZrF4-UF4; 8 Before filling 8.18 120 115 100 5Q0-46-4 After draining 8.06 10 315 40 847 LiF-Z¢F 4-U F4i 17.5 Before filling 9.39 290 155 260 50-46-4 After draining 9.69 10 1750 85 848 LiF-ZrF4-UF4; 19 Before filling 9.21 860 275 825 50-46-4 After draining 9.46 25 1850 80 860 NaF-ZrF ,-UF ; 10.5 Before filling 8.00 <2 80 95 {std) 50-46-4 After draining 8.50 15 1060 80 120 (53-35-8-4 mole %) in loops 843 and 844 resulted in heavy intergranular subsurface void formation, as shown in Fig. 5.8, The cold legs of loops 843 and 844 revealed light surface roughening and the for- mation of metallic crystals, Loops 841 and 842, which operated under similar conditions with NaF- LiF-ZrF ,-UF , (20-55-21-4 mole %), had moderate- to-heavy intergranular void formation in the hot leg, as shown in Fig. 5.9. Although some metal crys- tals were found in the cold legs of these loops, the quantities were smaller than those observed in loops 843 and 844, as would be expected because of the difference in ZrF , content. Six other standard loops were operated as the first of a series of loops to study the effect on corrosion of various alkali-metal fluorides as components of the MF- ZrF 4-UF4 (50-46+4 mole %) fuel mixture, where M stands for potassium, rubidium, or lithium. The resuits of metallographic examinations of these loops and chemical analyses of the fuel circulated are presented in Table 5.5. The maximum attack, which was to a depth of 19 mils, occurred in a loop (848) which circulated LiF-Z¢F ,-UF , (Fig. 5.10), and, as may be noted in Table 5.5, the chromium contents of the lithium-containing fuels circulated in both loops 847 and 848 were quite high. Loops 839 and 840, which circulated RbF- ZiF -UF ,, under similar conditions, showed hot- Inconel thermal-convection EEE] T lNC?ES T 1 EEERESR 2 = Maximum Hot-Leg Attack Found in Fig. 5.8. Inconel Thermal-Convection Loop 843 Which Circu- {ated NaF-LiF-ZrF UF, (53.35-8-4 mole %) for 500 hr ot a Hot-Leg Temperature of 1500°F. 250X, Reduced 32%, (Ssmwmet with caption) PERIOD ENDING MARCH 10, 1956 leg attack to a depth of 9 mils (Fig, 5.11). The systems in which KF-ZrF ,-UF, was circulated, floops 845 and 846, showed hot-leg attack to a depth of 8 mils (Fig. 5.12). The cold legs of all these loops showed slight surface roughening and very thin deposits. E __f] 1 w5 w T (&) z Fig. 5.9. Maximum Hot-Leg Attack Found in Inconel Thermal-Convection Loop 841 Which Circu- lated NaF-LiF.ZrF .UF, (20-55-21-4 mole %) for 500 hr at a Hot-Leg Temperature of 1500°F. 250X, Reduced 32%. (dmwww®t with caption) Maximum Hot-Leg Attack Found in Inconel Thermal-Convection Loop 848 Which Circu- lated LiF-ZrF , were similarly tested, one for 500 hr and the other for 1000 hr. The purpose of these three tests was to evaluate the corrosion resistance of the test specimens in sodium and the effects, if any, on the Inconel containers. The weight and dimensional changes of these 124 three specimens were positive, as was to be ex- pected with porous bodies, and they were less than 0.5%. The only macroscopic change was that the bufflike color of the specimens was altered to a gray-black by the tests. The untested and tested specimens are shown in Figs. 5.16 and 5.17. Powder x-ray diffraction comparisons of the untested and tested specimens did not reveal reaction products. The walls of the Inconel test containers were tinted slightly yellow in the liquid-sodium regions. Metallographic examinations of the specimens and the Inconel containers, chemical analyses of the sodium test baths, and microspark spectro- graphic anolyses of the inner surfaces of the Inconel containers will complete the examination of these tests, 5C. E. Curtis et al., ANP Quar, Prog. Rep. Sept. 10, 1955, ORNL.-1947, p 140. PERIOD ENDING MARCH 10, 1956 UNCLASSIFIED Y-1731 INCHES 0.02 - {0.038 Fig. 5.15. A 70% Ni-13% Ge-11% Cr—6% Si Brazing Alloy After Exposure to NaF.ZrF ;.UF, (53,5-40- 6.5 mole %) for 100 hr ot o Hot 2BeF, + UQ, ZrF4 + 2Be0 — ZBeF2 + Zro2 would be expected to take place. It is possible that BeO, specimens protected by ZrO, might be stable, but it is not likely. The B ,C specimen suffered only a slight loss in weight in this simple test. Exposure of the speci- men to air after the test resulted in the liberation of fumes, which may have been BF .. Further tests of this material seem to be justified. Ciffusion of Chromium in Alloys G. F. Schenck Pratt & Whitney Aircraft M. G. Leddicotte Analytical Chemistry Division Interface reaction of molten fluorides with the chromium component of Inconel results in the re- moval of the chromium from the alloy. A chromium concentration gradient is thus established through- out the metal wall, which causes migration of the chromium from the higher concentration area toward the chromium-depleted surface and results in the formation of subsurface voids. The thickness of the layer impoverished in chromium is a function of the depth of the corrosive attack. 138 Experiments are under way in which activation anclyses are being used to study the diffusion of A measure of the diffusion velocity may be obtained by activation-analysis measurements over a period of time, chromium in Inconel. A clean section of Inconel was activated in the ORNL Graphite Reactor for use as a standard on Three hours after removal of the specimen from the reactor the radiation rate from a 12-g section was 7 r/hr at 5 in. from the surface. This specimen was allowed to cool for several days to diminish the activity. At the end of this time most of the activity was that of activated chromium, and an autoradiograph showed the activity to be homogeneously dis- tributed. In the current work an effort is being made to establish whether diffusion studies are sensitive enough to follow the diffusion of chromium in the wall of a )-in.-OD tube. currently being investigated. graphic plate is exposed to the irradiated sample, and the optical density of the plate is then de- termined as a function of distance from the pe- riphery. It is believed from an evaluation of the data obtained that the results are inconclusive because of the limited resolution of the densitom- eter. In the second approach, the radioactivity in various portions of each sample is measured. It is believed that successively milling small amounts of metal from the inside of the tube will produce a satisfactory series of samples from increasing depths, as measured from the inner surface. The amount of the sample, the depth of each milling, and a measurement of the amount of Cr3' (27.8- day) radioactivity in each portion will be used to determine the degree of chromium diffusion, Preliminary work has consisted in determining the radioactivity of the samples used for radio- avtograph analyses by means of a gamma-ray spectrometer. The gamma spectral analysis dis- played a continuous record of the relationship be- tween the counting rate and the gamma-ray energy, and it was thus possible to obtain the relative intensities of all the gamma-emitting radioelements in the sample. The results indicate that it should be possible to measure the Cr®! intensity in each portion milled from the sample., Current work by the Analytical Chemistry Division includes the design of suitable equipment for obtaining and retaining each layer milled from the specimen, which to base future measurements, Two approaches are In one, a photo- PERIOD ENDING MARCH 10, 1956 6. METALLURGY AND CERAMICS Metallurgy Division WELDING AND BRAZING STUDIES P. Patriarca A. E. Goldman G. M. Slaughter Metallurgy Division Examination of NaK-to-Air Radiators After Service at High Temperatures Metallographic investigations are under way on NaK-to-air radiotors that failed after various periods of service at high temperatures. A large number of sections from radiators designated ORNL Nos. 1 and 3 and York No. 1 have been examined to determine the degree of adherence of the braze material at the tube-to-fin joints and the degree of oxidation of the copper fins. A correlation of these factors with heat-transfer performance is expected to give a better understanding of the relative importance of the fabrication features, ORNL radiators Nos. 1 and 3 failed, because of leaks, aofter 608 and 716 hr, respectively, of primarily isothermal service in the temperature range 1000 to 1600°F. York radiator No. 1 failed, as a result of a leak, after it had been in service in the range 1000 to 1600°F for 152 hr, Many metallographic specimens were taken from each radiator, and each specimen contained a portion of three tubes joined to 15 to 20 fins. The tubes were cross sectioned to show two opposite areas of each joint, Counts were made of the tube-to-fin joints which exhibited 0 to 24, 25 to 49, 50 to 74, and 75 to 100% adherence, and estimates were made of the degree of oxidation of the copper fins according to the categories non- oxidized, slightly oxidized, and heavily oxidized. The results of these investigations on the three radiators are presented in Table 6.1, Also, ORNL radiater No. 3 is being examined for possible causes of the failure. As may be seen in Fig. 6.1, the fire that occurred because of the leak destroyed the area of the leck, and therefore the examination is being made of the numbered sections. Similar sections are being cut from Pratt & Whitney radiator No. 2, which also failed because of a leak. Crack Susceptibility of Back-Brazed Tube-to-Header Joints One of the problems associated with the fabri- cation of the ftull-scale ART NaK-to-air radiators, designated air-cooled radiator 7503, is the brazing of the core halves and the welding of these into a single unit. The specifications suggest the joining of two core halves into a single unit by welding after the brazing operation. An expetiment TABLE 6.1. RESULTS OF EXAMINATIONS OF NaK-TO-AIR RADIATORS FOR TUBE-TO-FIN ADHERENCE AND FIN OXIDATION Radiator Designation ORNL No. 1 ORNL No. 3 York No. 1 Period of operation in the range 1000 to 1600°F, hr 608 76 152 Number of tube-to-fin joints examined 4150 2282 3847 Joints with 75 to 100% adherence, % 91.8 87.7 67.4 Joints with 50 to 74% adherence, % 4.2 3.5 13.0 Joints with 25 to 49% adherence, % 1.4 1.4 7.3 Joints with 0 to 24% adherence, % 2.6 7.4 12.3 Nonoxidized copper fins, % 59.3 12.1 75.4 Slightly oxidized copper fins, % 20.1 2.5 22,0 Heavily oxidized copper fins, % 20.6 85.4 2.6 139 ANP PROJECT PROGRESS REFORT UNCL ASSIFLED Y17425 UNCL ASSIFIED Y-17426 Fig. 6.1. ORNL Radiator No. 3 Showing Area Destroyed by Fire and Sections Cut for Metallographic E xomination. was therefore conducted to determine the influence of the stresses set up, during welding, on the crack susceptibility of the back-brazed tube-to-header joints. Two core halves of the type shown in Fig. 6.2 were assembled for brazing. Qne core half was brazed in the conventional upright position, fin collars down, while the other was brazed in the horizontal position. Brazing in the horizontal position eliminates the need for sump plates to accommodate excess brazing alloy and thus also eliminates the restraining effects of the sump plates on the tubes during nonisothermal service. All welding was done in the down-hand position in accordance with established welding procedures. The completed assembly is shown in Fig. 6.3, Dye-penetrant inspection of the tube-tosheader back brazes indicated freedom from cracks both before and after welding, Visual examination of the tube-to-fin joints indicated that good flow and fin-coliar protection had been achieved in both the 140 vertically and horizontally brazed core halves. The feasibility of brazing in the horizontal position will be investigated further in experiments with farger fin banks, 23/4 x 8 x 16 in., and with the brazing alloy preplaced as sintered rings and as dry powder. Measurement of Weld Shrinkage An experimental program was carried out to obtain information on the transverse weld shrinkage that will result from the various welding operations involved in the construction of the ART, The data will be of value in design and fabrication of the various components of the reactor, Butt welds were made by both the Heliarc and metallic-arc processes in %/16" 3/8-, 12-, and ‘z-in.- thick Inconel plates with several joint designs. A summary of weld shrinkage measured with micrometers and appropriate dial gages is pre- sented in Table 6.2, As may be seen from the data, weld shrinkage increases with plate thick- PERIOD ENDING MARCH 10, 1956 UNCLASSIFIED Y-17175 Fig. 6.2. Component Parts of Crack Susceptibility Test Specimen. UNCLASSIFIED Y-17297 Fig. 6.3. Completed Crack Susceptibility Test Specimen. 141 ANP PROJECT PROGRESS REPORT TABLE 6,2. SUMMARY OF SHRINKAGE OF BUTT WELDS ON INCONEL PLATES Joint Designs 1. J-type bevel with 60-deg included angle; welded in accordonce with procedure specification PS-12 2. VY-type bevel with 100-deg included angle; welded in accordance with pracedure specification PS-1 3. V-type bevel with 75-deg included angle; welded in accordance with operator’s qualification test speci- fication QST-12 Plate Weld Joint Welding Process Thickness Shrinkage Design (in.) (in.) Yariables: Joint Design and Plate Thickness 1 Heliarc and metallic arc 3/8 0.091 1/2 0.95 3/4 0.129 2 Heliarc 3/16 0.113 3/8 0.140 1/2 0.192 Variables: Joint Design and Welding Process 1 Heliarc and metallic arc 3/4 0.129 Metallic arc 3/4 0.119 Heliarc 3/4 0.189 3 Metallic arc 1/2 0.090 Heliare 1/2 0.129 Variable: Weld Volume 3 Heliarc 1/2 0.129 2 Heliarc 1/2 0.192 ness, and the Heliarc welding process results in larger shrinkage for a given joint design in a plate of given thickness than does the metallic-arc welding process., An analysis of the cross- sectional areas for the different joint designs in ]/z-in. plate indicates that other parameters, such as the mean weld width, are also important, Ex- periments on buttewelded low-carbon-steel test plates indicate that the shrinkage is slightly larger than that for Inconel, as would be expected from a comparison of the coefficients of thermal expansion: 6.4 x 10~% in./in.:°F for Inconel and 7.3 x 10=6in./in.+°F for low-carbon steel. Dimensional Control During Fabrication of Pump Yolutes One of the problems associated with the fabri- cation of an Inconel pump of the type designed 142 for pumping NaK in the ART is the maintenance of the critical spacing between the two volutes which constitute the pump housing, Two test pieces that are representative of the volutes were machined from 2-in. Inconel plate for use in a study of methods for controlling the spacing during fabri- cation, Four spacers were also machined from Inconel plate to act as rigid supports duting the welding of the wolutes. of the experimental specimen are shown in Fig. 6.4, The spacers were subjected to an aluminizing treatment prior to assembly of the components, to prevent self-welding. The assembly was welded inaccordance with procedure specifications PS-12. Micrometer measurements were made at four radial positions prior to and after each welding operation, The welded assembly was annealed in a helium atmosphere at 1850°F for a period of 2 hr, with These component parts PERIOD ENDING MARCH 10, 1956 UNCLASSIF{ED Y.17332 SPACER BOTTOM TOP PK-2 WELD TEST Fig. 6.4. Component Parts for Experimental Fabrication of NaK Pump Casing. the heating and cooling rates being maintained at 600°F/hr. After the welding and annealing had been completed, the spacers were removed by machining. An effort to remove the spacers intact was abandoned when it became evident that sig- nificant force would be required and that, as a result, the volute surfaces might be scored. The finished pump casing is shown in Fig. 6.5. Micrometer measurements of the inside surfaces of the volutes indicated that the maximum dimen- sional change was 0.015 in. The maximum shrink- age on the diameter was 0.056 in. It was also found after annealing and after removal of the spacers that the dimensions were exactly the same oas they were after welding. [t is evident therefore that the annealing cycle completely removed the residual stresses. A simijlar experi- ment is now under way to determine the degree of dimensional control which can be obtained when a brazing procedure is used in the fabrication of the pump casing. Cermet-to-Metal Joints The usefulness of cermet valve components for ANP valve applications depends to a large extent upon the possibility of successfully joining them to metallic structural materials, such as Inconel, The fabrication procedure described previously,! which utilized thin films of Electroless-plated nickel-phosphorus brazing ailoy on the cermet, has proved to be too sensitive to the plating variables, An experimental program is under way for developing a more reliable and consistent joining procedure, Since high-temperature brazing alloys of the Ni-Si-B type (such as Coast Metals No. 50; brazing temperature, 1120°C) will bond to cermets such as K-152B (64 wt % TiC—6 wt % NbTaLiC,~ 30 wt % Ni), a series of tests were conducted with these materials. The cermets are hard to wet, and it is very difficult to obtain an even layer of brazing alloy over the entire faying surface, Therefore a nickel screen was placed in intimate contact with the cermet to facilitate wetting, The “‘tinned’’ surface was then ground flat and copper- brazed to the nickel-transition layer and to the Inconel, as in the previous procedure.! A brazed joint that is typical of the joints obtained by this procedure is shown in Fig. 6.6. A valve disk fabricated by this new technique was found by dye-penetrant inspection to be free of flaws. p, Patriarca, ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL.-1947, p 131. 143 ANP PROJECT PROGRESS REPORT UNCL ASSIFIED 5 Y-1743 PK-2 WELD TEST Fig. 6.5. Casing. Finished Experimental NaK Pump Fig. 6.6. Cermet-to-Nickel Joint “*Wetted'’ with Coast Metals No. 50 Alloy and Brazedwith Copper. 100X, Reduced 32%. Another highly promising procedure now being investigated eliminates the need for brazing the cermet to the nickel-transition layer. It has been found that at a temperature of approximately 1350°C an interfacial reaction occurs, between the cermet and the nickel-transition layer, that results in a metallurgical bond. A joint made by this tech- nique is shown in Fig. 6.7. At a higher tempera- ture, approximately 1370°C, the extreme nonuniform 144 UNCLASSIFIED * ¥-17611 Fig. 6.7. Bonding at Approximately 1350°C. duced 32%. Cermet-to-Nickel Joint Formed by 100X, Re- UNCLASSIFIED ¥1T64S Fig. 6.8. treme Nonuniform Interfacial Reaction Resulting from Heating to Approximately 1370°C. Cermet-to-Nickel Joint Showing Ex- interfacial reaction shown in Fig, 6.8 occurred. Further tests will be made to establish the optimum conditions. Brazing Alloy Development The low-cross-section Ni-Ge-Cr-Si brazing alloy system has been found to have sufficient corrosion resistance to fuel mixtures and to sodium to be of use in circulating-fuel reactor fabrication. There- fore several sample melts of alloys in the high- nickel-content range of the system were made for flow-point determinations. The results of the tests are presented in Table 6.3, Four of the alloys with low flow points were submitted for corrosion testing in sodium, in NeK, and in fused fluoride salts, and the results are reported in Sec. 5, **Corrosion Research,’’ TABLE 6.3. RESULTS OF FLOW-POINT MEASUREMENTS ON THE Ni-Ge-Cr-5i BRAZING ALLOY SYSTEM Brazing Alloy Composition (wt %) Flow Point Ni Ge Cr Si c) 75 8 ) 6 1120 75 13 6 6 1120 73 13 1 3 1160 70 13 1 6 1100 68 13 13 6 1100 68 10 16 6 1100 67 13 1 9 1120 65 13 " 1 1140 65 13 16 6 1080 62 13 19 6 1080 62 16 16 6 1100 62 13 16 9 1100 59 16 19 6 1080 High-temperature oxidation and corrosion tests on Coast Metals alloy No. 52 (Ni-Si-B) have indi« cated that removal of a constituent or of con- stituents occurred during testing. Since this alloy has been used extensively in the fabrication of NaK-te-air radiators and fuel-to-NaK heat ex- changers, a study was made to determine the type and extent of removal that may be expected to occur during the intended service time. Some results of oxidation tests of cast alloy buttons are available, The structure of the alloy is shown in Fig. 6.9 as cast and after 100 and 500 hr at 1500°F, It may be seen that the depth of constituent removal increased with time and was approximately 0.006 in. after the 500-hr oxi- dation test. Microspark spectrographic and metal- lographic examinations of the sample indicated PERIOD ENDING MARCH 10, 1956 UNCL ASSIFIED CYe174926 Fig. 6.9. Results of High-Temperature Oxida- tion Tests of Cast Buttons of Coast Metals Braz- ing Alloy No. 52. (a) As cast. (b) Exposed for 100 hr at 1500°F, (c) Exposed for 500 hr at 1500°F. As polished. 100X. Reduced 4%. 145 ANP PROJECT PROGRESS REPORT that both boron and silicon were removed, Micro- hardness measurements on the interior of the as- cast specimen showed a hardness of 700 VHN, whereas the hardness of the depleted surface was 140 VHN, Similar tests conducted in a vacuum and in helium showed no removal of alloy constituents, Results of tests in NaK and in fused salts will be reported later, The brazing of aluminum bronze fins to Inconel tubes by conventional dry-hydrogen techniques has been unsatisfactory because of the formation of a thin film of aluminum oxide on the fin surface, Thin electroplates (<0.001 in.) of nickel and iron were inadequate diffusion barriers when the con- ventional heating time of 4 hr was used; however, heavier platings (0.002 in.) facilitated wetting. An Inconel T-joint brazed to a 6% aluminum bronze fin with Coast Metals alloy No. 52 is shown in Fig. 6.10; the 0.002-in, electroplate of iron used UNCLASSIFIED Y-174% Fig. 6.10. Inconel T.Joint Brazed to a 6% Alu- minum Bronze Fin with Coast Metals Brazing Alloy No. 52 Showing 0.002.in. Electroplate of Iron on Fin. 100X, Reduced 10%. 146 to facilitate wetting may be seen, The addition of manganese to nickel-base high- temperature brazing alloys has also been found to promote wetting. Additions of 30% manganese to the Coast Metals alloy No. 53 have permitted the direct wetting of the bronze, while additions of only 10% promoted flow on material plated with only 0.005 in. of nickel, MECHANICAL PROPERTIES OF INCONEL D. A. Douglas J. R. Weir Metallurgy Division C. R. Kennedy Pratt & Whitney Aircraft Creep-Rupture Design Data The creep testing of Inconel in argon and in NaF-ZrF -UF, (50-46-4 mole %) at 1300, 1500, and 1650°F is nearly complete, The data obtained recently have shown that a few minor modifications of previously published data are necessary, The data presented here, in Figs. 6,11 through .15, supersede the data presented in a previous report in this series.?2 The curves of Figs. 6.11, 6.12, 6.14, and 6.15 include data obtained at higher and lower stresses than were used previously, The curves of Fig. 6.13 reflect a better under- standing of the scatter of test resvits in the 3000- to 4500-psi stress range. Metallographic examina- tion of the test specimens showed that strain aging or strain-induced precipitation had occurred in some specimens in this stress range and had caused an increase in rupture time. Since this phenomenon did not occur in subsequent tests, the design curve was redrawn to include only the more reproducible data. Additional design curves for Inconel tested in argon and in fused salts are presented in Figs. 6.16 and 6.17, Low-Stress Creep Data Information on the creep of Incone! at very low stresses at 1300°F was obtained because it was needed in the design of components that must have dimensional stability at low stresses at the tem- perature of interest, The test results obtained to date indicate a total creep strain of 0.03% in 1000 hr for a stress of 1000 psi at 1300°F in air. 2R. B. Oliver et al., ANP Quar. Prog. Rep. March 10, 1955, ORNL-1864, Figs. 7.22, 7.24, 7.26, 7.28, and 7.29, p 122-126. PERIOD ENDING MARCH 10, 1956 ORNL-LR-DWG 12593 20,000 1% 5% |10% RUPTURE 10,000 ."_;. 2 5000 w w e} or - w 2000 1000 R i 10 100 1000 10,000 TIME {hr} Fig. 6.11. Stress-Rupture Characteristics of As-Received Inconel Sheet (Heat B) Tested in NaF-ZrF - UF, (50-46-4 mole %) at 1300°F. UNCLASSIFIED ORNL-LR—DWG 12594 20,000 5000 STRESS (psi) 2000 1000 1 s 100 1000 10,000 TIME (hr) Fig. 6.12. Stress-Rupture Characteristics of As-Received Inconel Sheet (Heat B) Tested in Argon at . 1500°F. 147 \ ANP PROJECT PROGRESS REPORT 20,000 10,000 5000 STRESS (psi ) 2000 1000 UNCLASSIFIED oy L ORNL—LR —DWG 12595 2 5 10 20 50 100 200 500 1000 2000 5000 13,000 TIME (hr) Fig. 6.13. Stress-Rupture Characteristics of Annealed (2050°F) Inconel Sheet (Heat B) Tested in Argon ot 1500°F. 20,000 10,000 5000 STRESS (psi) 2000 1000 UNCLASSIFIED ORNL-LR-DWG 12596 10% RUPTURE 2 5 10 20 50 100 200 500 1000 2000 5000 10,000 TIME (hr) Fig. 6.14. Stress-Rupture Characteristics of As-Received Inconel Sheet (Heat B) Tested in Argon 1650°F. 148 PERIOD ENDING MARCH 10, 1956 aikinpiini QRNL-LR~DWG 12597 20,000 10,000 4 5000 w o wl © 1% RUPTURE w 2000 1000 1 10 100 1000 10,000 TIME (hr) Fig. 6.15. Stress-Rupture Characteristics of As-Received Inconel Sheet (Heat B) Tested in NaF-ZrF .- UF , (50-46-4 mole %) at 1650°F. ORNL-LR-DWG §2598 20,000 RUPTURE 10,000 5000 STRESS (psi) 2000 10,000 1000 1000 1 100 TIME [hr) Fig. 6.16. Stress-Rupture Characteristics of Annealed (2050° F) Inconel Sheet (Heat B) Tested in NaF. ZrF -UF, (50-46-4 mole %) ot 1300°F. 149 ANP PROJECT PROGRESS REPORT e Ay g 5 UNCLASSIFIED ORNL-LR—DWG {2589 20,000 10,000 5C00 2 RUPTURE STRESS (psi) 2000 1000 i 2 5 10 20 20 100 200 500 1000 2000 5000 10,000 TIME (nr) Fig. 6.17. Stress-Rupture Characterisfics of Annedled (2050°F) Inconel Sheet (Heat B) Tested in Argon at 1650°F, B el TABLE 6.4. COMPARISON OF THE DUCTILITY AT RUPTURE OF 0.060-in.-THICK INCONEL SHEET WITH THAT OF 0.060-in.-WALL INCONEL TUBING IN ARGON AND IN i~-lc|F-ZrF4-UF4 (50-46-4 mole %) Elongation (%) Temperature Stress CF) (o) In Argon In NaF-ZrF4-U F4 Tubing Sheet Tubing Sheet 1300 15,000 2.8 50 12,000 8.8 70 10,000 8.8 40 8,000 13.3 25 1500 5,000 9.2 28 10.2 17 4,000 5.2 20 5.5 13 3,000 1.7 12 3.7 8 2,000 1.3 12 1650 3,000 6.7 30 6.8 19 2,500 3.6 30 5.3 16 2,000 3.8 30 3.3 15 1,500 3.1 12 2.8 8 150 Effect of Biaxial Stress on Creep Ductility at High Temperatures The stress-rupture strength of an internally pressurized 0.060-in,-wall Inconel tube is very similar to that of a 0,060-in,-thick Inconel sheet specimen, but there is a significant difference in the ductility of the two specimens at rupture, The relative values for the elongation in the direction of the maximum stress at several temperatures and stresses are shown in Table 6.4 for the two types of specimens tested in argon and in NaF-ZrF ,.UF (50-46-4 mole %). The decrease in ductility shown by the tube-burst specimen may be attributed to the 2-to-1 hoop-to-axial stress ratio set up in a closed-end pressurized tube. Metals deformed at room temperature under various stress systems exhibit increasing or decreasing ductility in com- parison with their uniaxial tensile ductility, depending on whether the stress system tends to increase or decrease the maximum shear stress,3 In the case of a biaxial tensile-stress system, such as is found in the tube-burst test, the maxi- mum shear stress is decreased by the action of the smaller axial stress, and slip is restricted, This results in lower ductility at rupture, aithough the time-to-rupture depends entirely on the larger hoop stress, The data presented here were taken at elevated temperatures, where the deformation mechanism is slightly different from that at room temperature, but the theories applicable to room- temperature ductility appear to apply. SPECIAL MATERIALS STUDIES J. H. Coobs H. Inouye J. P. Page T. K. Roche L. M, Doney J. A, Griffin A. J. Taylor Metallurgy Division M. R, D’Amore R. E. McDonald Pratt & Whitney Aircraft Y. M, Kolba, Glenn L. Martin Co. Neutron Shield Material for High- Tempetature Use H. Inouye A recent report states that irradiation causes extensive physical damage in hot-pressed B,C, Experiments showed that, in a neutron flux of 3\, Gensamer, Strength of Metals Under Combined Stresses, American Society for Metals, 1941, PERIOD ENDING MARCH 10, 1956 1013 to 104 nv, the B,C bodies began to crack at about 3% burnup of the B'0 atoms and that complete granulation occurred at 26% burnup of the B! atoms., Helium evolution from the irra- diated bodies was determined at various tempera- tures up to 1500°F, In general, the quantity of helium which was released increased with the burnup and the temperature to which the specimen was heated, being a maximum of 20% of theoretical at 36% burnup of the B! gtoms at 1500°F, The effects of irradiation on molded and sintered B,C are being determined by the Solid State Division (see Sec. 8, “Radiation Damage’') and will be compared with the effects on a hot-pressed B,C specimen which was irradiated under similar conditions. lrradiation of the specimens has been completed, and a cursory examination indicates that no cracking or crumbling occurred. Since the hot-pressed body did not show physical damage, the B'0 burnup is assumed to have been less than 3%. It was estimated previously that o B,C layer with a density of 2.0 g/em® would attain tempera- tures between 1500 and 2000°F at a flux density of 1013 nv in the ART. A further evaluation indi- cates, however, that the maximum temperature of the B,C layer in the ART during full power opera- tion will be 1800°F, if the helium gap between the B4C and the sodium-cooled lnconel shell is less, tham 0.020 in. Metal-bonded boron bodies are being studied as possible alternates for hot-pressed B,C as neutron shielding material because of the uncertainties with regard to the effects of irradiation on hot- pressed B,C. It was thought that, if boron could be incorporated in a ductile matrix, irradiation of the bodies probably would not cause fragmentation, The thermal expansion of a metal-bonded body would approach that of Inconel, and the thermal conductivity would be better than thaot of hot- pressed B,C, Also, the metal-bonded material would have better resistance to thermal and mechanical shock and could be fabricated in shapes which would fit a curved surface with gaps of less than 0.020 in. Since a decision as to the method by which a high-boron-content layer could be incorporated in an Inconel annulus in the ART was required before all the fabrication studies could be completed, a 4w. p. Valovage, Effect of Irradiation on Hot-Pressed Boron Carbide, KAPL-1403 {Nov. 15, 1955). 151 ANP PROJECT PROGRESS REPORT scheme was devised invelving a double layer of boron to circumvent the unknown effects of irradi- ation and to relieve the concern over the fit that could be achieved with ceramic tiles, the tem- perature and consequently the quantity of helium released being a function of the helium gap. The final configuration, which consists of two layers of boron-containing material, will result in a boron density of about 1 g/em?, whereas a mini- mum of 1.2 g/em? is desired. The details of the annular space containing the boron layer are shown in Fig. 6,18, The clad Cu-B,C layer will be nearest to the neutron source. The sections of the two layers will be staggered to eliminate neutron windows between adjacent pieces. Neutron pinholes will exist at line inter- sections, but they will be insignificant if a fit of better than 0.020 in. can be attained. In spherical areas the clad Cu-B,C sheet, about 8 x 8 in, and 0.100-in. thick, will be attached to the Inconel shell with one centrally located spot weld, Several canned B, C tiles will be spot-welded to each Cu-B,C sheet to anchor them in place. [n the areas that are not spherical the shield will be built up of layers of CU-B4C. These layers will not be vented for helium, since their low boron densities will minimize the amount of helivm formed. ORNL—-LR-DWG " 563;\ Q.010~in, STAINLESS STEEL CLADDING AND COPPER FOIL—=; \1 -, \\ ", 0.08C-in. COPPER B,C CERMET——__ |- ———=——=———-0.005~in. STAINLESS STEEL CLADDING REFLECTOR REGION \\ HEAT EXCHANGER REGION ~~0.27-in. B4C TILE AND COPPER COATING 7\o.osz~;n. INCONEL 0.25-in. INCONEL—" m\‘\ SPOT WELDS/ b ~ - p // ‘Q,;\\: N ———{0395 in Fig. 6.18. The Neutron Shield for the ART. 152 Fabrication of Boron-Containing Materials H. Inouye Copper-B,C, - Since B,C is chemically inert to copper, an investigation of the variables associated with the fabrication of bodies containing from 20 to 74 vol % B,C in copper was initiated. Thus far almost all the effort has been devoted to the fabrication of bodies containing 60 and 74 vol % 54C in order to determine the maximum concen- tration of boron which can be incorporated into a strong matrix. The theoretical properties of several Cu-B,C mixtures are listed in Table 6.5, TABLE 6.5. THEORETICAL PROPERTIES OF Cu-B4C MIXTURES 54C Content Density of Boron Volume Weight Mixture Density* Per Cent Per Cent (g/cm3) (Q/Cmg) 10 3.03 8.32 0.206 20 6.57 7.67 0.413 30 10.75 7.03 0.620 40 15.80 6.38 0.820 50 22.00 5.74 1.03 60 29.75 5.10 1.24 70 39,60 4.45 1.44 80 52.90 3.81 1.65 *Boron in B4C assumed to be 80%. The 74 vol % B,C-Cu mixture was crumbly and relatively weak after being sintered at 1650°F and coined at 33 tsi. A significant increase in both the strength and the density resulted when the sintered bodies were coined at 60 tsi. The boron densities of the best bodies fabricated were between 1.24 and 1.31 g of boron per cubic centi- meter of mixture, or between 83 and 88% of theo- retical. |t was observed that a sintering weight loss of 1% occurred when the compacted powders were sintered in hydrogen at 1650°F for 1 hr and that the bodies expanded, The dimensional in- crease did not vary significantly with the com- pacting pressure between 10 and 40 tsi, An addi- tional weight loss of 0.1 to 0.2% occurred when the bodies were resintered, at 1800°F, but there was no improvement in density, The sintering weight loss could arise from the volatilization of contaminants of the boron or from the reduction of copper oxide, A weight loss due to the volatilization of the boron would be sig- nificant if the boron content were low or if an accurate amount of boron were required in the body. The expansion which occurs does not permit coining without the removal of material from the edges, and consequently the thickness of the body is affected. The sintering variables were studied in detail on mixtures of 60 vol % B,C in copper. Sintering at 1600°F in helium or hydrogen atmospheres caused linear expansions as great as 3.0%, when either commercial- or high-grade B,C was used, The weight losses of bodies sintered in hydrogen were about 0.75%, and those of bodies sintered in helium were negligible. Prefiring the B,C powder resulted in bodies which shrank instead of ex- panding during sintering, with no weight changes. The boron densities of the best bodies of the .60 vol % B,C composition ranged between 0,977 and 0.998 g;cm3, as coined. A :’/lé-in.-thick body will support itself in a 2-in, span at 1950°F and will support a T-ib weight at 1650°F, with no measurable deflection. Further improvements of the properties can be derived by hot-rolling the coined body. Copper-B,C bodies containing 20, 30, and 40 val % B,C have been sintered and then coined to a 21-in, spherical radius at room temperature, The 40 vol % B,C.Cy composition is strong and shows some ductility at room temperature. The forming of this composition is facilitated when the mixture is hot-roll clad with stainless steel. Strips of the clad composite can be formed to a 2-in. radius without cracking, whereas the unclad material cracks and breaks upon moderate bending. Copper-AlIB,,. — Bodies containing 70 vol % AlB,, in copper were made and evaluated, |t was hoped that sintering this mixture would result in alloying and also the precipitation of finely divided boron. Moreover, if alloying occurred, the resulting composition (90% Cu-10% Al) would be ductile and oxidation-resistant, Boron densities of 1.15 g/cm® (78% theoretical) were attained with this mixture when it was sintered for 1 hr in hydrogen at 1650°F and coined at 60 tsi. When the body was resintered at 1850°F, it melted and o bronze was formed, as was expected. Increasing the AIB,, content to 78 vol % resulted in a body with a boron density of 1,17 g/em?3, PERIOD ENDING MARCH 10, 1956 lron-B,C and Molybdenum-B,C. — Cold-pressed and sintered bodies of Fe-B,C and Mo-B,C mix- tures were fabricated, Mixtures containing 80 wt % B,C were cold-pressed at 60 tsi and sintered at 1150, 1700, and 1950°C, In general, the densities of the bodies and the strengths increased with the sintering temperature. The maximum boron densities attained were above 1.50 g/cm®, The bodies showed tendencies toward lamination and brittleness. Ceramic-Boron Bodies., — Several samples of molded and sintered ceramic-B,C tiles have been submitted by the Norton Company and the Carbo- rundum Company for evaluation. Both companies have made bodies with boron densities of at least 1.2 g/cm® that have rabbeted edges. Thus far neither company has submitted spherical segments which could be measured for dimensional tolerances to obtain an indication of the helium gap that would exist between them and a curved Inconel surface, The Carborundum tiles, 3 x 3 x 0.315 in., were bonded with silicon and had densities of about 60% of theoretical, Radiographic examination of the tiles showed macroporosity, with voids as large as ]/8 in. in diameter, Thickness variations of as much as 0.006 in. were found for both the rabbeted edges and the total thickness., Numerous samples coated with ceramic powders and with flame-sprayed copper and iron have also been received for examination. The Norton Company has also submitted tiles of B,C bonded with carbon, The specimens with rabbeted edges were machined. Thickness meas- urements could not be made, because the surfaces were rough and the material was crumbly, Samples flame-sprayed with Rokide A (Al,0;) have also been received for evaluation, Boron Steels, — Boron steels have been studied in order to evaluate the feasibility of using them for a ring to transmit a compression load from the beryllium reflector to the support strut ring. A low-boron-content alloy containing a minimum of 0.65% B0 and a maximum of 1,30% would be re- quired to support a compression load of 500 psi at 1300°F, Small arc melts of iron-boron alloys containing 0.5, 1,0, 1.5, 2.0, 2.5, and 3.0% boron were hot- rolled in air at 1900°F, Edge cracking occurred in alloys containing 2.5 and 3,0% boron, The hard- ness of annealed alloys increased with the boron content, being in the range of 43 to 97 Rockweli-B. 153 ANP PROJECT PROGRESS REPORT The 1.5% boron alloy was cold-rolled 40% in thick- ness before cracking occurred, while compositions with less than 1.5% boron were rolled 65% before cracks developed. A 30-lb ingot containing 1% boron has been cast and will be extruded to 1 ]/z-in.-diu rod. The pressure ring, 37-in, OD, 0.500 in, thick, and ]1/2 in. wide, will be difficult to form and will probably have to be in two separate pieces. In order to determine the feasibility of casting the ring in one piece, iron-boron alloys containing 0.75, 1.25, 2.00, and 3.00% boron were cast into l-in.~dia ingots. All the clloy compositions had hardnesses of about 55 Rockwell-C and showed no response to heat treatments up to 1500°F, A larger ingot had a hardness of 77 Rockwell-B, which indicated that the cooling rate affected the structure and hence its nardness. In future tests examinations will be made of these alloys to determine the effects of irradiation, the yield strength of castings and wrought plate at 1300°F, and the extent of diffusion between the various alloys and Inconel. Tensile bars of the castings have been made; samples of wrought alloys have been submitted for irradiation; and diffusion couples of iron-boron alloys and Inconel have been service-tested for 500 hr at 1300°F, Bonding of Boron-Containing Materials to Inconel. — Experiments are in progress at the Carter Research Laboratories and the Vitro Lab- oratories to develop methods for thermally bonding a boron layer to Inconel. The experiments at the Carter Research Laboratories are feasibility studies which involve the codeposition of amor- phous boron and electroplates of copper and nickel. Thus far 10 vol % boron in copper has been deposited, and the deposit has good mechanical properties. The investigation at the Vitro Lab- oratories involves the deposition of uniform metal coatings on ceramic-B,C tiles and the bonding of Cu-B,C and Fe-B,C layers to Inconel by electro- phoresis. Conversion of Boron to B,C, — Another attempt was made to convert amorphous boron to B, C, The earlier experiment was repeated, and it was found that the conversion could be accomplished by firing of a mixture of boron and graphite at 1750°C, A 20% weight loss that occurred during firing can be attributed to loss of boron, inasmuch as x-ray patterns of the product showed an excess of carbon, The method appears to be unsuitable 154 for the conversion of expensive B0 unless the weight loss can be controlled and the product can be made crystalline. Inconel-Boron Compatibility M. R, D'Amore The compatibility of Inconel and beron-containing materials in intimate contact at temperatures in the range 1100 and 2000°F was investigated. Also, nine boron-free compounds were tested to determine their suitability as barrier coatings for Inconel in Inconel-boron assemblies. H. Inouye Diffusion couples of Inconel-B,C and Inconel- (B,C-Cu) were tested for periods up to 1000 hr at temperatures between 1100 and 1650°F. The test specimens were prepared by hot roll-cladding B,C powder and a 50 vol % B,C~50 vol % Cu mixture between Inconel cover plates. In addition the diffusion between Inconel and bodies of B ,C-Cy, B,C (Norbide), B,C-SiC, and boron powder was determined under a contact pressure of 50 psi. The extent of diffusion was evaluated metallo- graphically, and the results are presented in Table 6.6. Under the conditions of these tests, brittle re- action products were formed at the B,C-Inconel interface, and boron and carbon penetrated the Inconel along the grain boundaries, as shown in Fig. 6.19, which is a photomicrograph of a speci- men tested at 1500°F with no contact pressure, As may be seen from the test data, the diffusion is low at 1100°F, but it increases rapidly with temperature and with even moderate pressure. Intergranular penetration was found up to 28 mils in depth, but it appeared to have little hardening effect on the Inconel matrix, as illustrated in the microhardness traverse in Fig. 6.20. Additional tests were made to determine the compatibility of Inconel with metal borides, car- bides, carbonitrides, oxides, and nitrides. The test specimens, which were prepared by cold- pressing powder mixtures of Inconel and the com- pound powder, were heated at 1500 and 2000°F for 250 hr in evacuated quartz capsules or in helium, The reactions which occurred between the Inconel and the compounds were determined visually and by x-ray diffraction. The extent of the re- action was estimated by the appearance of new phases and the disappearance of the starting compounds. The data obtained are presented in Table 4.7. TABLE 6.6, PERIOD ENDING MARCH 10, 1956 RESULTS OF COMPATIBILITY TESTS OF INCONEL AND BORON-CONTAINING MATERIALS IN CONTACT AT HIGH TEMPERATURES Test Conditions Compatibility Depth of Boron or Carbon Diffusion (mils) ) Contact Atmosphere Specimen Time Temperature - hr) CF) Pressure Reaction Intergranular (psi) Layer Penetration |nconel-B4C 500 1100 None Evacuated Inconel capsule 1.0 1.0 1000 1100 None Evacuated lnconel capsule 1.0 1,0 500 1300 None Evacuated Inconel capsule 2 to3 4108 500 1500 None Evacuated Inconel capsule 1.5t0 2 15 1000 1500 None Evacuated lnconel capsule 5 to8 15t0 18 lnconel-(Cu-BAC) 1000 1100 None Evacuated Inconel capsule None 1.0 500 1300 None Evacuated Inconel capsule 2 to3 4108 1000 1300 None Evacuated Inconel capsule 5to 6 500 1500 None Evacvated Inconel capsule 1210 13 1000 1500 None Evacuated Inconel capsule 16 to 28 100 1500 50 Helium 1 tob 6 to 10 lnconel-54C 100 1500 50 Helium 2 to3 6 to 10 {Norbide) Inconel-boron 100 1500 50 Helium 1.0 6to 9 {powder) Inconel-boren 100 1500 50 Helium 1 to2 10 to 13 (hotpressed) |ncone|-(B4C-SiC) 96 1600 50 Helium 2 1510 20 Inconel-(BAC-SiC) 26 1600 50 Helium No apparent reaction (A|203 coated) |ncone|-B4C 96 14600 50 Heliom 2 15t0 25 (Norbide) Of the nine boron-free compounds which might be suitable as barrier coatings, the oxides ZrO,, Al,O,, and MgO were found to be inert to Inconel at temperatures up to 2000°F. The carbonitride (ZrCN), the SiC, and the SijN, reacted to some extent with Inconel, as evidenced by a shift in the diffraction lines of the Inconel and by the instability of the compound, which was indicated by the appearance of oxides. The absence of a reaction between Inconel and BN was unexpected, and, as yet, there is no explanation. It was ex- pected that nickel borides would be formed. This test is being repeated. The reactions between Inconel and the metal borides make them unsuit- able for use above 1500°F, presumably, because of the formation of nickel borides, which melt below 1900°F, Nickel-Molybdenum-Base Alloys T. K. Roche H. Inouye Fabricability Studies. — The forgeability of an alloy with the nominal composition of Hastelloy B (68% Ni—28% Mo—4% Fe) has been studied as a 155 ANP PROJECT PROGRESS REPORT | UNCLASSIFIED CUYarse Fig. 6.19. Inconel-B C Compatibility Specimen Tested ot 1500°F for 1000 hr with No Contact Pres- sure. 250X, Fig. 6.20. Microhardness Traverse of Inconel- (B ,C-Cu) Compatibility Specimen Tested at 1500°F for 1000 hr with No Contact Pressure. 100X, Reduced 35%. 156 function of carbon content in an attempt to ascer- tain the cause for the difficulties encountered in the fabrication of Hastelloy B, The poor tfabri- cability of Hastelloy B is believed to be attribut- able to a stable high-temperature phase that has been tentatively identified as (Ni Mo,},C. Arc melts of the alloy were prepared in an argon atmos- phere with nominal carbon additions of 0.01, 0.02, 0.04, 0.06, and 0.10% in 100-g samples. The alloys were cast into rectangular bars approxi- mately 3 in. long, %’ in. wide, and '/2 in. thick, and then hot-rolled at 1150°C. Reductions in thickness of 5% were used for the first few passes, and then 10% reductions per pass were used until a final thickness of 0.160 in, was attained, Visual examination of the hot-rolled strips showed slight edge cracking of the alloys with 0.01, 0.02, and 0.04% carbon additions, The alloys with 0.06 and 0.10% carbon additions were satisfactory. All the strips were then cold-rolled without difficulty to 0.100 in. at a rate of 0.01Q in. per pass, A study of the microstructures of these alloys revealed that the amounts of second-phase material, TABLE 6.7. PERIOD ENDING MARCH 10, 1956 AT 1500 AND 2000°F FOR 250 hr COMPATIBILITY OF INCONEL AND YAR!OUS COMPOUNDS HEATED Material Tested at 1500°F Tested at 2000°F X+Ray Examination Yisual Examination X-Ray Examination Visual Examination Incone|-5i3N4 Inconel-BN Incone |-(SiC-Zr02) Inconel-(Ni-MgO) Inconel-(ZrCN) |nconel-B4C Inncc:onel-":"_rO2 !nconel-A|2O3 Inconel-(A|203-Si3N4) lnconel-(B4C-SiC) |ncone|-(Zr02-M90) Inconel-SiC |n<:on<-:l-('§r82 Innt:or'ael-VB2 lncone |-TiB2 |nconel-Csz Inconel-WB |m::ont=,-|--TuB2 Inconel-MoB, Inconel-ZrB, InconeI-A|B‘2 Slight reaction No reaction Ne reaction Slight reaction ZrO2 No Inconel present; ‘no B4C present No reaction No reaction Neo reaction Slight reaction No reaction Slight reaction Slight reaction Slight reaction Slight reaction Slight reaction Slight reaction Slight reaction Slight reaction Slight reaction Dense, strong body Cracked and expanded body Cracked and expanded body No visible reaction Porous, friable body Porous, friable body No visible reaction No visible reaction No visible reaction Porous, friable body No visible reaction Expanded, porous body No apparent reaction No apparent reaction No apparent reaction No apparent reaction No apparent reaction No apparent reaction No apparent reaction No apparent reaction Moderate reaction No reaction No reaction No Ni-MgO present ZrO2 and ZrC formed No Inconel present; no BAC present No reaction No reaction Stight reaction SiC present; no B4C present No reaction Moderate reaction Moderate reaction Moderate reaction Moderate reaction Complete reaction Moderate reaction Moderate reaction Moderate reaction Moderate reaction Complete reaction; no AIB]2 or lnconel present Porous, streng body Porous, crumbly body Cracked and ex- panded body No visible reaction (Sample con- taminated) Metallie globules in a friable gray powder Hard, dense body No visible reaction Gray, friable body Gray, friable body Porous, friable body Dense, strong metallic body Hard, dense body Porous, hard body Melting occurred; molten phase mefallic Hard, dense body Porous, hard body Dense, hard body Porous, hard body Melting occurred presumably a carbide, in the alloys increased with increased carbon content. On the other hand, the carbon additions were effective in reducing the fine oxide type of impurity carried over from the melting stock and observed in the control alloy (0.01% C). The microstructures are illustrated in Figs. 6.21 and 6.22. A correlation of the hot forgeability of these alloys with their microstructures indicates that the oxide precipitates are more detrimental to 157 ANP PROJECT PROGRESS REPORT Fig. 6.21. Microstructure of 68% Ni—28% Mo-4% Fe Alloy Containing 0.01% Carbon. Oxide precipitates may be seen in matrix. 500X, Fig. 6.22. Microstructure of 8% Ni~28% Mo~4% Fe Alloy Containing 0,.10% Carbon. Carbide precipi- tates present, but matrix cleaned of oxide precipitates. 500X, 158 fabricability than is the presence of the carbide, up to 0.10% total carbon. Thus, close control of the melting practice will be required in order to improve the fabricability of nickel-molybdenum alloys. As was reported previously, attempts to extrude Hastelloy W billets to tube blanks for the pro- duction of seamless tubing have failed, A section was cuf through a fractured area of one of the extruded billets and examined metallographically. Evidence of melting was found adjacent to some of the fractures, as shown by the presence of the eutectic in Fig. 6.23. X-ray analysis of the frac- tured area did not reveal the nature of the eutectic, and, as a result, no definite information on the cause of the hot-shortness of the alloy was ob- tained, It is hoped that the use of a lower billet preheat temperature (1950°F rather than 2050°F) and modifications of the present billet lubrication methods will improve the extrudability of the Hastelloy-type alloys. ' PERIOD ENDING MARCH 10, 1956 Effect of Melting Practice. — A program has been initiated in which the melting practice will be closely controlled and will be correlated with mechanical properties and the fabricability of the Hastelloy B and Hastelloy W alloys. Arrangements have been made with Battelle Memorial Institute for the production of arc-melted ingots with the following compositions: Nickel, Hastelloy B, Hastelloy W, 76% Ni-17% Mo-7% Cr, and 83% Ni~17% Mo. The melts will be made by the consumable-electrode process to take advantage of the high arc temperatures for vaporizing ‘‘tramp’’ elements. Extrusion billets will be prepared from the ingots for the fabrication of suitable test specimens, Effect of Chromium Additions. —~ Previous work showed that chromium additions to a basic 20% Mo—80% Ni alloy resulted in poor forgeability of the alloy, which was attributed, in general, to the high oxygen content of the chromium. Since carbon additions improved the fabricability of nominal Hastelloy B, they were made to 5-Ib vacuum melts UNCLASSIFIED] Fig. 6.23. Section Through a Fracture Area of a Commercial Hastelloy W Billet Extruded at 2000°F, Eutectic structure may be seen adjacent to the fracture. 2000X. 159 ANP PROJECT PROGRESS REPORT of the basic 20% Mo—80% Ni alloy with 7 and 10% Cr added. If these alloys can be hot-rolled, larger heats containing 3, 5, 7, and 10% Cr will be pre- pared for the fabrication of seamless tubing to be evaluated in corrosion tests, Special Alloys. — Five special alloys prepared in 40-1b heats have been received from the Inter- national Nickel Company feor corrosion testing and strength evaluation, The compositions of these alloys are as follows (in each alley, nickel con- stitutes the balance): Component Amount (wt %) Mo 15 17 15 15 15 Cr 5 W 3 3 3 Nb 3 3 3 Al 0.5 0.5 0.5 1.0 0.5 Ti 1.5 C 0.25 Extrusion billets are being machined from these ingots for fabrication of seamless tubing and test specimens, Niobium Fabrication J. P. Page H. Inouye V. M. Kolba Arc-Melted Niobium. — Four fabricated ingots of pure arc-melted niobium prepared by Battelle Me- morial Institute have been received. These ingots were melted in crucibles lined with niobium foil, and getters were used. The dimensions and im- purity analyses of the ingots are presented in Table 6.8. A program has been outlined for determining some of the physical and mechanical properties of this TABLE 6.8, DIMENSIONS AND IMPURITY ANALYSES OF ARC-MELTED NIOBIUM INGOTS PREPARED BY BATTELLE MEMORIAL INSTITUTE Approximate Impurity Analysis Weight I:l?:f Dimensions (ka) (ppm) (in.) C H N © 10 0,5x1.5x75 1.07 100 6 440 717 13 0.5x 1.5x 4.5 0.56 100 4 330 448 14 0.5x2.0x 8.0 1.10 100 4 240 270 15 0.5x2.0x9.5 1.21 100 4 200 200 material with similar properties of the more common wrought material prepared by powder metallurgy techniques will be of particular interest., The wrought material has a severe limitation as an engineering material are available, Pack-Rolled Niobium. — Attempts to overcome the limitation of the size of the wrought niobium pieces have been made by pack-rolling several laminae in that only small pieces in an evacuated capsule. Preliminary tests indicate that the room-temperature mechanical properties of the laminated and nonlaminated ma- terials are quite similar, Room-temperature tensile properties for 0.065-in.-thick Inconel-clad speci- mens are presented in Table 6.9, The clad-core- clad thickness ratio of these specimens was ap- proximately 1:3:1. The clad-to-core bonds te- mained intact until fracture, The scatter in results can be attributed to small differences in the clad- core-clad ratios and, to some extent, to minor variations in the rolling technique. Creep Test Specimens. — The extreme sensitivity of niobium at elevated temperatures to even trace amounts of nitrogen and oxygen has made the va- material. A comparison of the properties of this lidity of available inert-atmosphere creep data TABLE 6.9. ROOM-TEMPERATURE TENSILE PROPERTIES OF INCONEL-CLAD NIOBIUM Number of Range of Range of o . Number Niobivm Laminae Diffusion Barrier of Tests Elongation Tensile Strength in Core (2% in 2 in.) (psi x 1073) 1 Tantalum 3 6.3 to 10.0 83.7 to 87.4 2 Tantalum 2 8.8te 9.0 78.5 to 78.9 5 Tantalum 3 7.5 t0 10.0 82.0 to 83.7 1 Copper~stainless steel 2 10.5 72,8 to 78.5 160 questionable. Attempts are therefore being made to fabricate Inconel-clad niobium creep specimens. Two methods of protecting the edge of the lami- nated sheet are being examined, In one method a clad-niobium panel is machined to slightly over- size creep-bar dimensions, the niobium is undercut by preferential chemical attack, an Inconel wire is inserted in the ‘‘groove,’’ and the assembly is welded shut. A small test section has been suc- cessfully sealed by this technique. in the other method a niobium core and its matching Inconel frame are machined to creep-bar dimensions in one direction and to one-fifth the final dimensions in the other direction; the core is then clad by hot-rolling in the latter direction to creep-bar size. A test section, in which the niobium was simulated by stainless steel, has been prepared by this method, This fabrication technique vyields specimens that are completely edge-protected. For examinations, the core could be located on radiographs and the excess Inconel could be cut away, Creep data should indicate whether the Inconel cladding supplies appreciable strength to the sheet, According to published data,® the creep strength of Inconel is so far below that of niobium that the effect of the cladding will be negligible. Fabrication of Large Panels. ~ Large panels of clad niobium must be fabricated by welding, and therefore the minimum core-to-core distance (across the weld) is of considerable interest. Metallographic examination of an Inconel-clad niobium weldment has established a minimum core-to-core distance of approximately ]{1 in, for ',é-in. sheet, with both copper—stainless steel and tantalum-foil barriers, Diffusion Barriers. — Further comparison of tanta- lum and copper—stainless steel diffusion barriers has yielded some anomalous data. It was reported previously® that the cold-rolling properties of the tantalum-barrier material were superior to those of copper—stainless steel barrier material, These sheets had been hot-rolied at 950°C prior to cold- rolling. Similar cold-rolling tests of material that had been hot-rolled at 1050°C indicate the reverse to be true. A program has been outlined for de- 5General Electric Co., Aircraft Nuclear Propulsion Department Engineering Progress Report No. 10, APEX- 10 (March 1954§’ 6). H. Coobs et al., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 144. PERIOD ENDING MARCH 10, 1956 termining the effect of rolling temperature on these and other properties of roll-clad niobium sheet, Niobium-UQ, Compatibility. — Niobium is being considered for use in solid fuel elements because of its good high-temperature strength. In order to evaluate the usefulness of niobium in this type of service, a program was initiated for investi- gating the compatibility of niobium and UO,. In a previous examination of swaged and rolled ni- obium powder-UO, encapsulated compacts, a phase, other than the UO, particles, was found in the niobium matrix. An x-ray diffraction pattern of the compact after it was aged for 500 hr at 1000°C indicated essentially niobium and UO,, but lines of Nb,O, were also found. Metallographic samples of as-received and as- swaged niobium powder also showed the third phase found in the initial niobium-UO, compacts. A Bergsmann hardness check on the as-received powder particles revealed that the hardness of the niobium matrix was about 210 VHN, whereas that of the second phase was about 2400 VHN, as shown in Fig. .24, Wrought niobium plate obtained from Fansteel Metallurgical Corp. was metallographically ex- amined and found to have a few small inclusions of the compound in an essentially clean matrix of niobium. Uranium oxide powder was placed between two pieces of the Fansteel plate, encap- sulated, and rolled, and a portion of the rolled plate was etched and then examined at 1000X, No reaction between the UO, particles and the niobium matrix was found, A portion of this rolled specimen has been aged for 500 hr at 1000°C and is being examined metallographically, The results obtained thus far indicate that ni- obium powder of higher purity must be obtained, A high-purity grade of powder is being prepared by converting wrought material to the brittle hy- dride, grinding, and, finally, vacuum annealing to remove hydrogen. A further test of the compat- ibility of UO, and niobium will be conducted with this high-purity powder. A high-temperature (~2000°C) vacuum furnace will be constructed to accomplish the sintering of the niobium powder compacts., Seamless Tubular Fuel Elements J. H. Coobs M. R. D'Amore The work on simulated seamless tubular fuel elements was continued with the preparation and 161 ANP PROJECT PROGRESS REPORT AVERAGE HARDNESS OF COMPOUND, 2390 VHN (4 IMPRESSIONS) (3 IMPRESSIONS) UNCLASSIFIED Y-17489 Fig. 6.24. Hardness Impressions on a Polished Section of As-Received Niobium Powder. 500X, Re- duced 2%. extrusion of four three-ply extrusion billets, The core materials were ~100 mesh type 302 stainless steel powder with 30 vol % Al,O, to simulate U02. The billet can material was type 316 stainless steel. Two billets were filled with loose powder tamped to a density of about 55% of theoretical. The other two billets had hot-pressed cores with densities of 80 and 89% of theoretical. The ex- trusion data for the billets are presented in Table 6.10. Sections 18 in. long were cut from the center of extrusions 3 and 4 and shipped to the Superior Tube Company for redrawing to small-diameter tubing. Extrusions 1 and 2 and the remaining por- tions of extrusions 3 and 4 were then split longi- tudinally, and the flow of material was examined visually. It was found that slight misalignment of billets 2 and 3 in the container of the extrusion had caused uneven material flow during The cores in these two extrusions were press extrusion, 162 offset in the direction of extrusion. The core cross sections of the extruded tubes appeared to be concentric except in extrusion 3. The eccentricity of this core was attributed to nonuniform tamping of the core powder during billet preparation. Segments have been cut from extrusions 1, 2, and 4 at various intervals, and the layer thick- nesses are being measured. Thus for the measure- ments have been completed only on extrusion 4, A section approximately 33 in. long, which con- tained about 70% of the original core material, was found to be uniform, The layers were cal- culated to have extruded in a Y:%:¥% thickness ratio, with the actual thicknesses being 0.048 in, for the outside cladding, 0.041 in. for the core, and 0.037 in, for the inside cladding. The work on seamless tubular fuel elements is currently being directed toward a study of the re- drawing properties of the material. Efforts are also being made to increase the length of the PERIOD ENDING MARCH 10, 1956 TABLE 6,10, EXTRUSION DATA FOR THREE-PLY SIMULATED FUEL ELEMENTS Billet soaking temperature: 2100°F Extrusion No, Type of Core A|203 Particle Mesh Size Extrusion Ratio ] Tamped powder 2 Hot pressed 3 Tamped powder 4 Hot pressed =325+ 5:1 -140 +325 9:1 -140 4325 21:1 -325* 21:1 *Grade 38-500. uniform core section of the extrusion, The Alle- gheny Ludlum Steel Corp. is aiding ORNL in the study of three-ply extrusions, Control Rod Fabrication J. H. Coobs R. E. McDonald M. R, D'Amore An investigation has been initiated of the feas- ibility of extruding control rods to close dimensional tolerances, The control material of interest is a 30 wt % Lindsay oxide—~70 wt % Ni mixture, which is to be formed into a 2.5-in.-0OD, 2.0-in.-ID, 24- in.-long cylinder ¢lad externally and internally with 0.020-in.-thick Hastelloy X. Similar control materials containing iron rather than nickel and Inconel rather than Hastelloy X are also of interest, Small rods of these materials will be used for the initial extrusion studies because the extrusion press at ORNL does not have the capacity to ex- trude a full-sized control rod. Several methods have been investigated in an effort to determine the optimum means of fabri- cating cores of the Lindsay oxide—metal mixture for 3-in.-dia extrusion billets. Small compacts containing 30 wt % Lindsay oxide and 70 wt % Ni have been fabricated successfully both by hot- pressing and by cold-pressing plus sintering., Hot- pressing of the powder mixtures at 1200°C produced a sound compact with a density of 86.4% of theo- retical, Cold-pressing plus sintering resulted in a compact with a density of 83% of theoretical, In preparing the compact the powder was pressed at 80 tsi, sintered at 1900°F for ]/2 hr, coined at 50 tsi, sintered again at 2100°F, and coined again at 50 tsi. The cold-pressed body cracked in one area during sintering at 2100°F. The method used gives satisfactory densities in small bodies, but it is not amenable to fabrication of cores for 3-in.- dia extrusion billets because of the excessive pressure requirements, Shield Plug for ART Pumps J. P. Page J. H. Coobs The shield plugs surrounding the shafts of the ART fuel and sodium pumps must fulfill three primary functions; they must serve as neutron, gamma-ray, and thermal shielding. In their thermal- shielding capacity they must prevent freezing of the fuel below them during zero power operation. Since no known material will satisfy all these conditions adequately, it has been proposed that the following layers of materials be stacked in a vented Inconel can: (1) a ]/B-in. disk of a B,C-Cu mixture, for neutron shielding; (2) a 5/é°in. disk of zirconia, for thermal shielding; (3) a 4- to 5-in, slug of high-density (>12 g/cm3) low-conductivity (<0.12 cal/em?/sec) material, for gamma-ray shield- ing. The top surface of the gamma-ray shielding material is to be brazed to the Inconel can, which will be cooled on its outer surface. Tantalum-constantan and tungsten carbide-con- stantan compacts are being investigated for use as the gamma-ray shielding material. A successful diffusion bond of tantalum-constantan to nickel has been made by heating the parts at 1100°C in a hydrogen atmosphere for 20 min and then cooling in @ helium atmosphere. The specimen was cooled in helium in order to minimize the formation of tantalum hydride, which probably caused the crack- ing of several earlier specimens during cooling. Lithium-Magnesium Alloys R. E. McDonald A study of the fabricability of a light alloy con- taining 20% Li has been initiated. Such an alloy would be useful as shielding material. Tests were 163 ANP PROJECT PROGRESS REPORT started on a 20% Li-80% Mg alloy, but it was found to be highly reactive in both air and water. A method of surface protection is necessary before mechanical and physical data can be acquired. Chemical coatings, flame spraying, and roll cladding were investigated without success. Thus far the Dow No., 17 anodizing treatment, which can be supplemented by a Unichrome plastic coating, has formed the only successful surface protection. A roll-claddable alloy would be more suitable for shielding, and a less reactive surface shouid be more amenable to roll cladding. Therefore an effort is being made to produce a less active surface by adding small amounts of aluminum to the lithium-magnesium alloy. The lithium-to- magnesium ratio is being held constant at 1:4, and ternary alloys with 0.5, 1, 1.5, and 2% Al are being investigated. Fabrication of Ceramic Materials L. M. Doney J. A, Griffin A. J. Taylor L ow:Density Rare-Earth-Oxide Compacts. — Meth- ods are being studied for preparing compacts of samarium and gadolinium oxides (Lindsay Code 920) for use in ART control rods. The compacts are to be 1.275 in, OD, 0.775 in. ID, and 1.000 in. long, and they are to have densities of between 3.95 and 4.4 g/cm3. Experimental compacts have been produced with a density of 3.53 g/em?® and a porosity of 53.5%. These compacts have been exposed to molten sodium at 1300°F for 100 hr and have been found to have a porosity to sodium of 53%. The compacts were not damaged by the immersion in sodium. The compacts required for the actual ART control rods will probably be pre- pared by a method very similar to that used in the fabrication of these experimental compacts. Cermets of Iron and Rare-Earth Oxide. — Cermets of 50 vol % iron and 50 vol % rare-earth oxide have been requested for use in the extrusion of Inconel control rods with cermet cores. The fabri- cation method devised for these cermets involves mixing 25 vol % iron metal, 25 vol % iron as Fe,0,, and 50 vol % samarium and gadolinium oxide (Lindsay Code 920) and pressing the mixture into a hollow, cylindrical shape at o very low (10,000 psi) pressure in a steel die. The cylinder is then isostatically pressed at 80,000 psi in a rubber bag immersed in an oil-filled pressure vessel, The isostatic pressing assures uniform density, 164 After pressing, the cylinder is sintered in hydrogen at 1475°C for l/2 hr. Cermets fabricated by this method have a density of 4.6 g/cm® and a porosity of 38.8%. Pieces of the size required for extrusion will be produced by this method. Cermets of Nickel and Rare-Earth Oxide. — Cermet compacts of 70 vol % nickel and 30 vol % rare-earth oxide have also been requested for ex- trusion studies of Inconel-rare-earth-oxide control elements. These compacts were fabricated by the same method as that used for the production of the iron—rare-earth-oxide cermets, except that the compacts were sintered in hydrogen at 1400°C for 20 min, The compacts had a sintered density of 6.1 g/em? and a porosity of 28.4%. NONDESTRUCTIVE TESTING R. B, Oliver J. W, Allen K. Reber Metallurgy Division Success in the inspection of small-diameter Inconel and Hastelloy B tubing with the Cyclograph, de- scribed previously,” prompted further study of the instrument, It is primarily a tuned oscillator in which the oscillator coil encircles the tubing, The instrument measures the amplitude of the oscil- lations, the amplitude being a measure of the resistive component of the coil’s impedance. Theo- retically, since only the resistive component is measured, all changes in the tubing cause the same type of variation,® and good tubing could possibly be rejected because of an inconsequential variation. However, experience has proved that the wobble of the tubing in the coil, which is the worst offender in producing spurious flaw indi- cations, can be eliminated by a well-designed mechanical feed mechanism, such as that shown in Fig. 6,25, The instrument is sensitive to slight changes in outside tubing diameter and wall thick- ness, but these changes occur very slowly over the length of the tube and can be easily separated from flaws, which give abrupt signals, An important feature of the Cyclograph is that its operating frequency is determined by its sensing coil and, hence, may be altered by merely changing the coil. The proper frequency for a particular tube is selected according to its outside diameter, 7R. B. Oliver et al., ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 164. 8R. B. Oliver et al., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 139. wall thickness, permeability, and conductivity. A graph of the Cyclograph reading vs tubing wall thickness, expressed as a percentage of the out- side diameter, is shown in Fig., 6.26. The pa- rameter for the graph is frequency-normalized to a valve /_, which is determined by the outside OUNTING HOLES i Yg~in MICARTA BOARD i - L aiFiLar WINDING/ = SECTION A-A iy oA et PERIOD ENDING MARCH 10, 71956 diameter, the permecbility, and the conductivity according to the relationship given on the graph, The operating frequency that gives the best results is the one which locates o point near a peak on the graph of Fig. 6.26, The dashed lines indicate the nonlinear characteristic of the instrument near UNCLASSIFIED ORNL-LR-DWG 11015 1 T 1’,* 0 1 DETAIL OF CYCLOGRAPH COIL FOR TUBING FLOATING MOUNTING FIXED MOUNTING WITH SYNCHRONOUS MOTOR DRIVE WEIGHTED ROLL / Fig. 6.25. Cyclograph Coil Assembly and SYNCHRONOUS MOTOR DRIVE FIXED MOUNTING WITH ™. SUPPORT PLATE 2 TUBING UNDER INSPECTION T T, 0 v " S Q % COIL ASSEMBLY -. o T i, Feed Mechanism for Tubing Inspection. 165 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORML—=LR—DOWG 11105 2 Try.yDE fo = jes) o 1 =PERMEABILITY (henries/meter) — ¥ =CONDUCTIVITY (mhos/meter) -—- D =0UTSIDE DIAMETER (meters) D O o o CYCLOGRAPH READING (% OF FULL SCALE) N O O 5 10 15 20 25 30 35 40 45 50 TUBE WALL THICKNESS (% OF D) Fig. 6.26. Cyclograph Reading vs Tube Wall Thickness with the Operating Frequency of the Cyclograph as a Pargmeter. the '‘quench’ or zero-reading axis, which allows extreme sensitivity to small flaws in the tubing. A plot of the frequency which will locate a peak on the graph of Fig. 6.26 for each value of wall thickness is given in Fig. 6.27. By consistent use of operating frequencies chosen from these plots, flaw types will always be detected in the same relative magnitudes. For instance, an inside- diameter crack extending through 10% of the wall thickness would cause the same indication from a 0.25-in.-0D, 0.025-in.-wall tube as from a 0.5- in.-OD, 0.049-in.-wall tube. Although it is not possible to obtain a calibration of flaw size vs instrument indication, it is possible to obtain an approximate determination of flaw size in terms of percentage reduction in wall thickness averaged over the length of the coil. Typical Cyclograch indications of outside-diameter cracks in Hastel- loy B tubing and inside-diameter intergranular attack in Inconel tubing are shown in Fig. 6.28, and photomicrographs of two of the flaws thus detected are shown in Figs. 6.29 and 6.30. The "B scan presentations of the eddy-current probe coil and of immersed ultrasonic indications are also photographed, in an effort to identify the several characteristic types of defects found in small-diameter tubing., Photomicrographs are also prepared of appropriate sections of the tubing, but difficulty has been encountered in the preparation of representative metallographic sections because of the minute dimensions of the defects and in- ability to sufficiently pinpoint the defects for good metallographic presentation, 166 UNCLASSIFIED ORNL-LR—-DWG 11106 100 2 5 ;2: Tr;..cyDz = PERMEABILITY (henries /meter) 7 = CONDUCTIVITY (mhos /meter) D= DIAMETER {meters) | 1 2 5 10 20 50 100 TUBE WALL THICKNESS (% OF QD) Fig. 6.27. Frequency vs Tube Wall Thickness for Maximum Resistive Coil Impedance. A type of defect that has been prevalent in small- diameter Hastelloy B tubing is a longitudinal crack which may penetrate as much as 80% of the tube wall from the outside surface. A typical example is illustrated in Fig. 6.29. The ultrasonic “B'" scan, as well as the Cyclograph, indicated this crack, as shown in Fig. 6.31, A crack of this size can be detected by ultrasound through almost 300 deg of tubing rotation, as is demonstrated by the diagonal indication that extends almost com- pletely across the screen, The eddy current probe coil “B"" scan of this crack is shown in Fig. 6.32. A characteristic defect in Inconel is the inter- granular type of attack illustrated in Fig. 6.30. Attempts to detect this defect ultrasonically gave inconclusive indications, probably because the penetration was shallow and the orientation was not parallel to the axis of the tubing, cumulative factors which might preclude valid ultrasonic in- spection. Mechanical difficulties caused by the excessive wobble of the tubing prevented an ade- quate probe-coil inspection. The small l/lé-in.-long gouge shown in Fig. 6.33 on the surface of an Inconel tube was readily detected ultrasonically, A representative ‘‘B" scan picture for the ultrasonic detection of this defect is shown in Fig. 6.34. PERIOD ENDING MARCH 10, 1956 UNCLASSIFIED ORNL—LR-—DWG 11104 . \ \\ Ty L L LY \ Y CYCLOGRAPH TRACE AT 160 ke OF REJECTED W Lo L 0.25-in.-0D x 0.049-in.~WALL HASTELLOY B \ \\ \ \ \L \\ , \\ Ro8EA I ‘\ \\ TUBING. OUTSIDE DIAMETER RADIAL CRACK ~ \ ! L j | | -3y | 0.027 in. DEEP AT POSITION R586A-1{—-1 IS | Lfll .l——v—k’t JM\J ) i ‘ SHOWN IN FIG. 6.29. T T T / / ,.-' / | / ; : — —— ? / / // [ /) // [ / 3 CYCLOGRAPH TRACE AT 200 kc \\ OF REJECTED 0.25-in.-0D x W 0.025-in.-WALL INCONEL TUB- ING. INSIDE DIAMETER INTER- GRANULAR ATTACK AT POSITION /L 34RA TO A MAXIMUM PENETRA- TION OF 0,002 in. IS SHOWN IN FIG. 6.30. Fig. 6.28. Cyclograph Traces of Defective Tubing. . Fig. 6.29. Radial Crack 0.027 in. Deep on Outside Surface of a 0.25¢in.-0OD, 0,049-in.-Wall Hastelloy B Tube Detected by Indication Shown in Fig. 6.28. 100X, Reduced 2%. 167 ANP PROJECT PROGRESS REPORT Fig. 6.30. Intergranular Attack to a Maximum Depth of 0.002 in. on a 0.25-in.-0OD, 0.025-in.-Wall Inconel Tube Detected by Indication Shown in Fig. 6.28. 150X, UNCLASSIFIED PHOTO 15158 Fig. 6.31. Ultrasonic ‘‘B’’ Scaon of Crack Shown in Fig. 6.29. 168 PERIOD ENDING MARCH 10, 1956 UNCLASSIFIED PHGTO 16155 Fig. 6.32. Eddy Current ‘*B’’ Scan of Crack Shown in Fig. 6.29. Fig. 6.33. Gouge 0.004 in. Deep on the Outer Surface of a 0.25-in.-0D, 0.025-in.-Wall Inconel Tube. 100X, ' 169 ANP PROJECT PROGRESS REPORT UNCL ASSIFIED Y-16454 Fig. 6.34. Ultrasonic ‘'B’’ Scan Indication of the Gouge Shown in Fig. 6.33. In an effort to compare the results obtained from immersed ultrasound, radiography, and fluorescent penetrant inspections, six pieces of 0.1875-in.-0D, 0.025-in.-wall, 7-ft-long CX-900 Inconel tubing were inspected by the respective methods. Mere detection was considered to be sufficient to indi- cate a defect, and no attempt was made to estimate the defect size. The total number of such defect indications and the correlation obtained between the various methods are presented in Table 6.11, The sampling is insufficient for final decisions to be made concerning inspection methods, but definite trends are indicated. |t appears that no one of the methods alone would be adequate, and if any one of them was eliminated a few defects would be overlooked. TABLE 6.11. CORRELATION OF RESULTS OF INSPECTION OF CX-900 INCONEL TUBING BY X-RAY, ULTRASOUND, AND FLUORESCENT PENETRANT METHODS Number of Defects Correlated by Number of Defects Found by Number of Defects Which Did Not Toro Morbodde Number of “Tube Each Method Correlate With Other Methods Defects " Krey Ultmosound TSI e Flerescomt ey Tey Ultosond - Correlared by Penetrant Ultrasound Penetrant CxX-1 7 13 14 2 5 6 3 3 6 i CxX2 2 4 16 2 0 12 0 0 4 0 CX3 1 25 13 0 13 1 1 1 12 1 CX<4 6 5 12 6 2 9 0 0 3 0 CX5 3 4 14 1 1 11 1 1 2 0 CX-6 2 Z 13 0 3 Z 2 0 4 o Total 21 58 82 1 24 48 7 5 30 2 170 PERIOD ENDING MARCH 10, 1956 o 7. HEAT TRANSFER AND PHYSICAL PROPERTIES H. F. Poppendiek Reactor Experimental Engineering Division FUSED.SALT HEAT TRANSFER H. W, Hoffman Reactor Experimental Engineering Division D. P, Gregory Pratt & Whitney Aircraft Preliminary heat-transfer studies were carried out on a fresh batch of the fuel mixture NaK-KF- LiF-UF, (11.241-45.3-2.5 mole %) in a newly constructed system containing a Hastelloy B test section. The results, given in terms of the Colburn function (Fig. 7.1), lie above those obtained with an earlier batch of the same salt but are still below the general heat-transfer correlation. The Reynolds modulus range of the data will be ex- tended in further experiments. It is believed that the difference between the two sets of data may be the result of suspected variations in the two salt batches. The new set of heat-transfer data falls about 30% below the correlation for ordinary fluids, and this difference may be attributed, in part, to the inaccuracy of the estimated value of 0.010 Pr 2/ . 3 /st N o o S w 0.002 COLBURN FUNCTION, / 0.001 1000 2000 5000 thermal conductivity used in the correlation of the salt data. Operation of a heat-transfer system that includes a pump to circulate the fused salts was initiated. Operational experience with NaNO,-NaNO,-KNO, (40-7-53 wt %) shows that the system performs satisfactorily at temperatures up to 700°F. Heat- transfer experiments with water (Fig. 7.2) indicate that no systemic error exists. The pump system will be used to study the salt mixture NaF-KF- LiF-UF, (11.2-41-45.3-2.5 mole %) rather than NaF-ZrF ,-UF, (50-46-4 mole %), as reported earlier, However, the heat-transfer characteristics of the latter salt will be examined in o pressurized system. At present, fluid flow rates are obtained in the pressurized systems with volume probes. Although extreme precautions are taken, small quantities of oxygen and water vapor possibly enter the system when probe changes are necessary. In addition, it is difficult to maintain the calibration of the volume probes. Therefore a more positive method ORNL-LR-DWG 12600 O PREVIOUS DATA ® NEW DATA . : | GENERAL CORRELATION FOR ORDINARY FLUIDS: /= 0.023 N;EOE 10,000 20,000 50,000 100,000 REYNOLDS MODULUS, Vg, Fig., 7.}, Comparison of Heat Transter Measurements on Two Batches of NaF-KF-LiF-UF , (11.2-41. 45,3-2.5 mole %) with the General Correlation for Ordinary Fluids. 171 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 12601 500 200 100 C. NNu/NPr 4 50 20 5x 0% REYNOLDS MODULUS NRe Fig. 7.2. Results of Heat Transfer Measurements with Water in the Test System that Includes a Pump. of obtaining the fluid flow rate has been investi- gated. The fluid weight is obtained by measuring the deflection of a cantilever beam which supports one of the tanks, Direct calibration of the system is accomplished by loading it with known weights, The beam movement can be measured by using either an iron core microformer or strain gages. The deflection signal, when amplified, can also be used to operate the system cycling mechanism. Thus, the necessity of opening the system during the experiment will no longer exist. ART FUEL-TQ«NaK HEAT EXCHANGER J. L. Wantland Reactor Experimental Engineering Division Heat-transfer and isothermal friction character- istics of the ART fuel-to-NaK heat exchanger have been determined for three degrees of spacer density: 9 sets of spacers at 71/2-in. intervals, 13 sets of spacers at 5-in, intervals, and 25 sets of spacers at 2Y-in. intervals, The data, a portion of which have been extrapolated to zero spacers, are shown in Figs., 7.3 and 7.4. For the normal spacer density (13 spacers at 5-in, intervals) the heat-transfer and isothermal friction characteristics in the Reynolds modulus range of 1,500 to 10,000 can be expressed by NNU/NPI'O-4 = 0.0155 NRGO'B 172 —— ORNL-LR-DWG {2602 100 | } 1. | L L | - . e 25 SPACERS AT 2%-in. INTERVALS | | >0 4 13 SPACERS AT 5-in. INTERVALS (NORMAL) w9 SPACERS AT 7',-in. INTERVALS || | | — GENERAL CORRELATION ‘ — . [ iy s . | ! i /.':’L / . I | "({l .‘, 'I} . } /‘A /| . P o s~=—EXTRAPOLATED b S, .»/x‘é}}g, TO ZERO SPACERS T 10 S A LT L ) B ¥ | /‘{":;’ | ]T - Pab- s el ] R ] oA | o2 a0 > 04 _ ", 08 AL Ny Mo, = 0.0155 N, 5 4, A . . . [ SN IR ] - — . - . . e L & - | | 10° 2 5 104 REYNOLDS MODULUS, Mg, Fig. 7.3. Heat Transfer Characteristics of an Experimental ART Fuel-to-NaK Heat Exchanger for Three Spacer Densities. and f = 0.175/Ny _0-39. The limited data presented in Figs. 7.3 and 7.4 appear to indicate that for a given heat exchanger pressure drop the heat- transfer coefficient is almost independent of the spacer density. For example, as the spacer density increases, the Reynolds modulus decreases to such an extent that the Nusselt modulus does not change significantly, The available heat-transfer data for flow parallel to a square array of cylindrical tubes and for flow through square cross-sectional ducts' are summa- rized in Fig. 7.5. It is thought that a square duct 1J. L. Wantland, Thermal Characteristics of the ART Fuel-to-NaK Heat Exchanger, QRNL CF-55-12-120 (Dec. 22, 1955). o> ORNL-LR-DWG 12603 05 0 25 SPACERS AT 2% —in. INTERVALS A 13 SPACERS AT 5-in. INTERVALS (NORMAL!} ® 9 SPACERS AT 7% —in. INTERVALS -—— EXTRAPQOLATED TO ZERC SPACERS -—— GENERAL CORRELATION 0.2 FRICTION FACTOR, £ o 0.05 0.02 0.01 103 2 5 104 REYNOLDS MODULUS, NRe Fig. 7.4. lsothermal Friction Characteristics of an Experimental ART Fuel-to-NaK Heat Exchanger for Three Spacer Densities. approximates the flow configuration of the ART heat exchanger better than any other geometry for which heat-transfer data are known., At high Reynolds moduli the data are in agreement with the relation NNu/NPro'A = 0.023 NRGO'S; in the transitional flow regime all experimental data fall below this equation. The data for curve 2 were obtained with 20- and 100-tube heat exchangers circulating NaK and fuel. An analysis has been made in an effort to recon- cile the disagreement between the data presented PERIOD ENDING MARCH 10, 1956 W 5 ORNL—LR—-DWG 12604 10 T ! T T = 1 WATER FLOW PARALLEL TC TUBES IN A 17T 5 | SQUARE ARRAY, S = 1.{§ J | 2 ART FUEL FLOW PARALLEL TO TUBES IN A SQUARE ARRAY, § = {11 2 — } 3 AIR FLOW IN A SQUARE DUCT, £/0 =100 E 16* |~ 4 WATER FLOW IN A SQUARE DUCT, L/ = 130 | — 5 WATER FLOW PARALLEL TO TUBES IN A ST SQUARE ARRAY, § =112 & WATER FLOW PARALLEL TO TUBES IN A |7 2 |- SQUARE ARRAY, § =1.2 e s 0.4 _ 08 e ~——GENERAL CORRELATION W, /W °* = 0.023 a, i I L2, 1 i T g i 5 e = o oo b . st o] & i P et - = ; Lo i 1 G 15 S N fi/ i EZ 2 : // . 102 - "/ il - e - i -1 5 — > TN | ‘;’/ 2 : ,;Z;’ /o 10 pooT < : /1 = - A i 2 & - { 1 | I | 10 2 5 10t 2 5 10° 2 5 108 REYNOLDS MODULUS, Ay, Fig. 7.5. Heat Transfer Characteristics for Flow Through Square Ducts or Parallel to Cylindrical Tubes in a Square Array. in Fig, 7.4 on the effect of spacers on the fluid friction characteristics in an ART heat exchanger tube bundle and the data reported by Cooper.? This analysis indicated that the differences in the spacings between the tube bundle and the container wall for the two experimental configura- tions were not great enough to explain the differ- ence between the two sets of friction data, Actually, this disagreement is not serious, because the spacer friction is only a small part of the total friction, M. H. Cooper, Pressure Drop of Heat Exchanger Tube Spacers, ORNL CF-55-11-180 (Nov. 28, 1955). 173 ANP PROJECT PROGRESS REPORT . . ART CORE HYDRODYNAMICS C. M. Copenhaver F. E. Lynch Reactor Experimental Engineering Division G. L. Muller Pratt & Whitney Aircraft A 5/22-sca|e model of the ART sodium coolant annulus, which was designed for a study of the fluid flow distribution, was fabricated and as- sembled without the spacers and lands, Although the model does not incorporate the actual reflector- coolant hole distribution, provisions have been made for tapping off the flow (representative of reflector-coolant holes) at various sectors of the inlet header, The preliminary experimental results obtained were based on 38.6% of the reflector-coolant flow going to the annulus and the remaining 61.4% being distributed uniformly to the coolant holes. The limitations of the experimental system made it impossible to attain the average Reynolds modulus of the actual coolant annulus, approximately 10%; however, Reynolds moduli up to 10* were obtained, and the results were extrapolated to give an esti- mate of the flow distribution, For a concentric annufus (width, 0,135 in.) the average flow through the portions of the annulus farthest away from the inlet appears to be about 10% greater than that through the portions nearest the inlet. Spacers have been inserted in the annulus according to the ART design, and flow-distribution and pressure- drop measurements are being obtained for concentric- annulus conditions, for various radial displace- ments of the outer core shell, and for other conditions. Yisual studies were made of the flow through the 21-in, ART core model after the installation of vortex generators supplied by Pratt & Whitney Aircraft, Vane angles of 50, 55, 60, 65, and 70 deg were used. The 50- and 55-deg vortex generators improved the flow in comparison with that obtained with the 45-deg generators, but the generators with larger angles gave no further improvement, In all cases there was a persistent region of fluctu- ating flow at the equator on the island wall, and flow through the core was characterized by macro- scopic turbulence. Although no large reverse-fiow regions existed, steady and unsteady flow were observed in various regions of the core. A faired- in entrance has been built to use with the vortex 174 generad®eemin order to get better flow distribution going into the vanes, TEMPERATURE STRUCTURE IN REGION BEYOND THE ART REFLECTOR H. W. Hoffman J. L. Wantland C. M. Copenhaver Reactor Experimental Engineering Division One of the several interrelated problems arising in the ART design is the determination of the temperature increase of the reflector-cooling sodium as it returns outside the reflector to the sodium pumps. Since the outlet temperature of the sodium is fixed by the allowable sodium-beryllium inter- tace temperature, the sodium temperature rise on the return side is important in establishing the axial temperature gradient available for cooling both the reflector and the outer core shell. |In order to evaluate this temperature rise, it was necessary to determine the heat sources and their locations. The idealized slab system for which the temperature analysis was made is indicated in Fig. 7.6. The actual system, in which the boron carbide tiles contact the walls at a finite number of points, has been simplified in terms of two In system A it is assumed that no thermal resistance exists between the boron carbide and the Inconel shells containing it, while in system B a 10-mil-thick helium gap isolates the boron carbide from the Inconel shells. Systems A and B are considered to be lower and upper limits of the problem. The heat-deposition rates used in these calculations are given in Table 7.1, limiting systems. The temperature profile in the region between: the sodium and the fuel return circuits was cal- culated by the method of superposition for a slab geometry. The temperature profiles obtained at the outlet end for systems A and B are compared in Fig. 7.6. The temperature of the sodium was evaluated by using the heat-flux distributions on each side of the sodium return passage. On the reflector side, the heat flux at any point was the heat generated in the 2 in, of beryllium immediately adjacent to the sodium as corrected for the heat flow in the beryllium resulting from conduction. On the out- side, the flux was determined from the temperature profile obtained by iteration. The resultant tem- perature rise of the sodium was 33°F in system A and 25°F in system B, PERIOD ENDING MARCH 10, 1956 S ORNL~LR —DWG 12605 1450 SYSTEM A | 1400 —-1—=S50DIUM 1350 INCONEL BORON CARBIDE INCONEL FUEL 1300 TEMPERATURE (°F} 1250 / 1200 1450 SYSTEM B 1400 SODIUM HELIUM HELIUM o 1350 iINCONEL BORON CARBIDE INCONEL TEMPERATURE o C C 1250 1200 60 50 40 30 20 10 o —10 DISTANCE FROM INCONFL—FUEL INTERFACE ( x1073in.} Fig. 7.6. Comparison of Temperature Profiles in Region Beyond the ART Reflector. TABLE 7.1, HEAT-DEPOSITION RATES IN ART FOR REGION OUTSIDE THE REFLECTOR Region Heat Deposition Rate Beryllium 2 in. from Be-Na interface 4.639 x 104 Btu/hr -ft3 1 in. from Be-Na interface 6.282 x ]04 Btu/hr -h3 At Be-Na interface 9.568 x 104 Bru/hr « it . 4 3 Sodium 3.479 x 10" Btu/hrft Inner Incone! shell 5.5467 x 105 Btu/hr -fts Boron carbide Volyme 1.894 x 107 Btushr - 5 Flux at reflector-side surface 4.154 x ]04 Btu/hr -f12 Outer Inconel shell 4,842 x 105 Btu/hr -ft3 175 ANP PROJECT PROGRESS REPORT JEMPERATURE STRUCTURE IN AN IDEALIZED ART CORE H. F. Poppendiek L. D. Palmer Reactor Experimental Engineering Division An analysis was-made of the temperature struc- ture in an idealized ART core. From the activity data obtained in the hot-critical experiment,® it was possible to obtain a simplified radial volume- heat-source distribution. This source distribution was substituted into the heat-transter equations, which were then evaluated for the uncooled-wall case. The resulting uncooled-wall temperature difference above the fluid temperature was more than twice as great as the corresponding tempera- ture difference in a uniform volume-heat-source system. This radial temperature information was then used in a general cooling analysis of the idealized ART core and cooling annuli such as that outlined previously.? The resulting axial sodium, core wall, and fuel temperature distribu- tions are shown in Fig. 7.7; the wall cooling by the sodium coolant in the annuli was approxi- mately 2.5 Mw, ART CORE HEAT-TRANSFER EXPERIMENT N. D, Greene G. W. Greene H. F. Poppendiek Reactor Experimental Engineering Division G. L. Muller Pratt & Whitney Aircraft Fabrication of the half-scale model of the ART core, together with its associated components, was completed. The assembly, which will be used for volume-heat-source heat-transfer experiments, has been operated with water (Fig. 7.8). Prelimi- nary measurements indicate that a mean Reynolds number of 100,000 will be achievable with the existing pumps. All recording instrumentation for the transient and steady-state thermocouples, which are located near the wall-fluid interfaces, as well as below the wall surfaces of the core and island, has been developed and installed, Operating procedures and schedules are now being established, and explora- 3A. D. Callihan et al., ANP Duar, Prog. Rep. Dec. 10, 1955, ORNL.-2012, p 71. 4H. F. Poppendiek and L. D. Palmer, Application of Temperature Solutions for Forced Convection Systems with Volume Heat Sources to General Convection Problems, ORNL-1933 (Sept. 29, 1955). 176 —— ORNL~LR-DWG 12606 Re=94,000 Pr=273 1600 1500 1400 TEMPERATURE (°F) 1300 1200 14G0 1000 DISTANCE FROM CORE INLET {ft) Figo 7-7- AJ!T-(:Oreo Temperature Structure in an ldealized tory, low-power runs will be made within a few weeks., HEAT CAPACITY W. D, Powers Reactor Experimental Engineering Division The heat capacities and enthalpies of four spe- cific NaF-ZrF, mixtures were measured in the liquid and solid states in order to determine whether there was an appreciable variation in the heat capacity with composition. No major change in the heat capacity or in the product of the heat capacity and density was found over the compo- sition range studied: 53 to 65 mole % NaF and 35 to 47 mole % ZrF,. The following equations represent the data obtained. NaF-ZrF , (65-35 mole %) Solid (120 to 465°C) PERIOD ENDING MARCH 10, 1956 o o vy o™~ O - o X 0. 1on, - ART Volume-Heat-Source Test Sect ig. 7.8. F 177 ANP PROJECT PROGRESS REPORT H, — Hygoo =—4.1 +0.1835T + (6.12 x 10=3)T2 -, b c, =0.1835 +(12.23 x 10=3)T Liquid (605 to 890°C) Hop = Hygoo ==1.7 +0.3322T ~ (3.87 x 10~%)72 <, =0.3322 — (7.73 x 10~3)T NaF-ZrF , (61-39 mole %) Solid (120 to 470°C) Hyp = Hysoc ==2.7 +0.1648T +(8.30 x 10~3)7172 c, =0.1648 + (16.61 x 10-3)T Liquid (550 to 900°C) Hp = Hycoe =16.8 +0.28357 ~ (0.84 x 107%)7? ¢, =0.2835 - (1.67 x 10-9)T NaF-ZrF, (57-39 mole %) Solid (120 to 495°C) Hp = Hygoo ==6.6 +0.1914T +(2.97 x 10~3)12 c, =0.1914 + (5.93 x 10-3)T Liquid (550 to 900°C) Hp — Hygoc =60.2 +0.15307 + (8.48 x 10~5)T2 ¢, =0.1530 + (16.95 x 10=3)T NaF-ZrF , (53-47 mole %) Solid {120 to 515°C) Hop = Hygoo =-5.2 +0.1847T +(2.38 x 10=%)T2 c, =0.1847 + (4.76 x 10=5)T Liquid (560 to 905°C) Hp = Hygoe =555 +0.1884T + (4.65 x 10~3)72 c, =0.1884 + (9.30 x 10=3)T In these expressions, H = enthalpy in cal/g, c, = heat capacity in cal/g . °C, T = temperature in °C, Table 7.2 gives a comparison of the heat ca- pacities in the liquid state in the following units: cal/g:°C, cal/g-atom-°C, and cal/cm3.°C. The volumetric heat capacities and the heat capacities on a per-gram basis have been found to be nearly constant over a wide range of compositions, In a report® published recently the heat capac- ities of 17 fluoride mixtures are listed and com- pared. Equations were formulated so that the enthalpy and heat capacity of solid and liquid fluoride mixtures could be predicted from the composition, YISCOSITY S. |, Cohen Reactor Experimental Engineering Division A study was initiated to determine the effect of fission products on the viscosities of fuel mix- tures, The mixture used to simulate the ART fuel after exposure to reactor conditions was NaF- ZrF ,-UF , (50-46-4 mole %) with 11 g of RbF, 17 g of BaF,, and 52 g of TaF; added per kilogram of pure fuel, A sample of this fuel was prepared by dry mixing of the constituents, Measurements were made on the mixture and on a sample of the pure fuel from the same batch to furnish a control. However, the solubilities of the additives were apparently poor, and it was not possible to obtain a solution of the various components. Hence, the data obtained are felt to be invalid. A second sample of this proposed mixture is being prepared that will be hydrofluorinated after the additives 5W. D. Powers and G. C. Blalock, Enthalpies and Heat Capacities of Solid and Molten Fluoride Mixtures, ORNL- 1956 (Jan. 11, 1956). TABLE 7.2. HEAT CAPACITIES IN THE Nch-ZrF4 SYSTEM Composition (mole %) Heat Capacity NaF ZrF 4 cal/g «°C cal/g satem -°C cal/em®2C 65 35 0.278 7.82 0.812 61 39 0.272 7.79 0.809 57 43 0.272 7.9 0.824 53 47 0.253 7.50 0.776 178 have been introduced, The resulting mixture will be analyzed, and a second attempt to obtain vis- cosity data will be made. An important observation was made during this study: the viscosity measurements on the pure fuel used as a control yielded values which were about 15% lower than those obtained previously for this mixture.® Consequently, measurements were made on a second sample of fuel from the batch used previously. The results obtained were in satisfactory agreement with those previously reported. Yisual comparison of the melts from the old and new batches of fuel indicated that the new material was of much higher purity. These ob- servations clearly show the influence of fuel purity on the viscosity, Data on three other fluoride mixtures were also obtained, The results are presented in Table 7.3. The NaF-ZrF, data were obtained with the modi- fied Brookfield and capillary viscometers, and the results were in satisfactory agreement with capil- lary viscometer results determined over a year ago. The two beryllium-bearing salts were studied in the beryllium dry box with the capillary vis- cometers, 65, |. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 30, ORNL CF-55-3-62 (March 2, 1955). PERIOD ENDING MARCH 10, 1956 PERFORMANCE COMPARISONS OF SEVERAL FLUORIDE FUELS H. F. Poppendiek Reactor Experimental Engineering Division There are many ways in which the performance or effectiveness of fuels in a circulating-fuel re- actor can be compared. In the comparisons pre- sented here it is presumed that the reactor power, fuel inlet and outlet temperatures, and reactor core and heat exchanger geometries are given and fixed. Also, it is considered that the core Reynoids num- ber is about 100,000 and that the heat exchanger Reynolds number is above the transition region, Under these circumstances it is possible to ex- press the heat and momentum transfer in the reactor core and the heat exchanger in terms of simple relations involving only the thermal properties, namely, heat capacity, thermal conductivity, vis- cosity, and density. These relations have been evaluated for several of the more important ANP fuels, and the data are presented in Table 7.4. All quantities were normalized, that is, all numbers in a column have been compared with the lowest, which has been called unity., It is noted that the major differences between the fuels occur in the heat exchanger. The last two mixtures, which may be called alkali-metal-base fuels, have radial fuel temperature differences that are from 20 to 40% lower than those in the zirconium-base fuels, Also, the heat exchanger pressure drops for the alkali-metal-base fuels are only about’ half as great as those for the zirconium-base fuels, TABLE 7.3, FLUORIDE VISCOSITY DATA Composition Temperature Absolute Viscosity Kinematic Viscosity B/T(°K) (mole %) (°C) {centipoises) (centistokes) = Ae Reference NcJF-ZrF4 550 11.3 3.45 p o= 0.07294]60/T - (50-50) 850 2.9 0.96 Nc:F-LiF-BeF2 600 55.65 2.65 po= 0.195e2930/T o (49-36-15) 800 2.95 1.44 NaF-LiF-Ber-UF4 575 8.4 3.45 p= 0.07364010/'1‘ s (56-21-20-3) 850 2.7 1.18 *S. 1. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 31, ORNL CF.55-12-128 (Dec. 23, 1955). **S |, Cohen and T. N. Jones, Measurement of the Viscosity of Compositions 97 and 98, ORNL CF-55-12-127 (Dec. 23, 1955). 179 ANP PROJECT PROGRESS REPORT TABLE 7.4, PERFORMANCE COMPARISONS OF SEVERAL FLUORIDES Composition Comparison Ratios™ Reactor Heat Exchanger (mole %) dq At — Ay Ap om dA T c NaF-ZrF ,-UF 1.16 1.15 1.25 2.04 {50-46-4) LiF-NaF.ZrF ,-UF , 1.36 1.18 1.45 1.53 (55-20-21-4) RbF-ZrFA-UF4 1.06 1.00 1.33 2.74 (48-48.4) LiF-NaF-KF-UF 4 1.00 1.25 1.00 1.17 (45.3-11.2-41-2.5) NaF-BeFZ-UFJ‘ 1.06 1.26 1.07 1.00 (70-28-2) *Atom = radial fuel temperature difference in the core for an uncooled wall, dq — | = wallecooling flux required to lower the wall temperature to the fluid temperature, dA c Atr = radial fuel temperature difference in the heat exchanger, Ap = the fuel pressure drop through the heat exchanger. 180 PERIOD ENDING MARCH 10, 1956 8. RADIATION DAMAGE G. W, Keilholtz Solid State Division REVIEW OF TESTS OF CORROSION OF INCONEL AND STABILITY OF FUEL UNDER IRRADIATION G. W. Keilholtz Solid State Division The effects of irradiation on the corrosion of Inconel exposed to a fluoride fuel mixture and on the physical and chemical stability of the fuel mixture have been investigated by irradiating Inconel capsules filled with static fuel in the MTR and by operating in-pile forced-circulation Inconel loops in the LITR and in the MTR. The relatively simple capsule tests have been used extensively for the evaluation of new materials. The principal variables in these tests have been flux, fission power, time, and temperature. In a fixed neutron flux the fission power is varied by adjusting the uranium content of the fuel mixture, Thermal-neutron fluxes ranging from 101} to 104 neutrons/cm?.sec and fission-power levels of 80 to 8000 w/cm3 have been used in these tests, Almost all the capsules have been irradiated for 300 hr, but in some of the recent tests the irradi- ation period was 600 to 800 hr. After irradiation the effects on the fuel mixture are studied by measuring the pressure of the evolved gas, by determining the melting point of the fuel mixture, and by making petrographic and chemical analyses. The Inconel capsule is also examined for corrosion by standard metallographic techniques. In the many capsule tests made to date, no major changes that can be attributed to irradiation, other than the normal bumup of the uranium, have occurred in the fuel mixtures. However, the analytical method for the determination of chromium in the irradiated fuel mixture is being rechecked for accuracy, The metaliographic examinations of Inconel capsules tested at 1500°F for 300 hr have shown the corrosion to be comparable to the corrosion found under similar conditions in unir- radiated capsules, that is, penetration to a depth of less than 4 mils. In capsules that briefly reached temperatures of 2000°F and above, there was penetration to a depth of more than 12 mils and grain growth, Three types of forced-circulation in-pile loops have been tested. A large loop was operated in a horizontal beam hole of the LITR, The pump for circulating the fuel in this loop was placed outside the reactor shield. A smaller loop, in- cluding the pump, was opetated in a vertical position in the lattice of the LITR, A third loop was operated completely within o beam hole of the MTR. The operating conditions for these loops are presented in Table 8.1, and results of chemical analyses of the fuel mixtures circulated are given in Table 8.2. The LITR horizontal loop operated for 645 hr, including 475 hr at full reactor power. The loop generated 2.8 kw, with a maximum fission power of 400 w/em3. The Reynolds number of the circulated fuel was 5000, and there was a temper- ature differential in the fuel system of 30°F. The volume of the loop was large, and therefore there was a large dilution factor. Metallographic analyses showed less than 1 mil of corrosion of the Inconel walls of the loop., Chemical analyses showed the irradiated fuel mixture to contain 180 ppm Fe, 150 ppm Cr, and 30 ppm Ni. Therefore there was no evidence of accelerated corrosion in this experiment, The LITR vertical loop operated for 130 hr, with only 30 hr of the total operating period being at full reactor power. The loop generated 5 kw, with a maximum fission power of 500 w/cm3, The Reynolds number of the fuel was 3000, and the temperature differential was 71°F. The surface- to-volume ratio was 20, and the dilution factor was 10. Chemical analyses of the irradiated fuel showed 370 ppm Fe, 100 ppm Cr, and 50 ppm Ni. Metallographic analysis of this loop showed less than 1 mil corrosion at the curved tip. The horizontal loop inserted in the MTR (MTR in-pile loop No. 3), described previously,! operated for 467 hr, including 271 hr at power. The loop generated 20 kw, with a maximum power of 800 w/em3, The Reynolds number of the fuel was 5000, and the temperature differential was 155°F for 103 hr and 100°F for 168 hr. The dilution factor was about 5, ]D. B. Trauger et al., ANP Quar, Prog. Rep. Dec. 10, 1955, ORNL-2012, p 27. 181 ANP PROJECT PROGRESS REPORT TABLE 8.1. OPERATING CONDITIONS FOR INCONEL FORCED-CIRCULATION IN-PILE LOOPS i . LITR Horizontal LITR Vertical MTR In-Pile Operating Yariables Loop Loop loop No. 3 Fuel composition (mole %) NaF-ZrFA-UF4; NqF-ZrF4-UF4; NqF-ZrF4-UF4; 62.5-12.5-25 63-25-12 53.5-40-6.5 Maximum fission power, w/c:m3 400 500 800 Total power, kw 2.8 5.0 20 Dilution factor 180 10 5 Maximum fuel temperature, °F 1500 1500 1500 Maximum temperature differential, °F 30 71 155 Reynolds number of fuel 5000 3000 5000 Operating time, hr 645 130 467 Time at full power, hr 475 30 271 Depth of corrosion ottack, mil <1 < <1 TABLE 8.2. CHEMICAL ANALYSES OF FUEL MIXTURES CIRCULATED IN INCONEL FORCED-CIRCULATION IN-PILE LOOPS Minor Constituents (ppm) Loop Designation Sample Taken lton Chromium Nickel LITR horizontal lcop Before filling 80 ¥ 10 1015 200 * 100 After draining 180 T 40 150 £ 10 30 t5 LITR vertical loop Before filling 90 £ 10 80 £ 10 145 T 20 After draining 370 * 20 100 * 20 50 10 MTR in-pile loop No. 3 Before filling 40 £10 60 £ 10 40 T 10 After draining 240 1 20 50 = 10 100 £ 20 Chemical analyses of fuel from MTR in-pile loop No. 3 showed 240 ppm Fe, 50 ppm Cr, and 100 ppm Ni. The high iron concentration probably resulted from a sampling difficulty. Examinations of unetched metallographic sections of the Inconel tubing showed no corrosion penetration. Etched sections showed no attack that was to a depth of more than 1 mil. A slight amount of inter- granular void formation was noted, but this was neither dense nor deep. Measurements of wall thickness showed no variations attributable to cotrosion, The loop was examined carefully for effects of temperature variations between the 182 inside and outside walls of tubing at bends, but no effects of overheating were observed. Samples from the inlet, the center, and the exit side of the loop showed less than 1 mil of corrosion. The low corrosion is credited to careful temperature control of the salt-metal interface and to the maximum wall temperature being below 1500°F at all times. Future loops will be operated at higher fission powers and therefore greater temperature differ- entials, The dilution factors will be kept low. New fuels aond new alloys are being considered for testing in future loops. EXAMINATION OF THE DISASSEMBLED MTR IN-PILE LOOP NO, 3 M. J. Feldman C. Ellis R. N. Ramsey E. J. Manthos A. E, Richt W. B. Parsely E. D, Sims Solid State Division The sectioning of MTR in-pile loop No. 3 was completed in the General Electric Company hot- laboratory facilities at NRTS, and three carriers containing the different sections of the loop were received at ORNL. The first carrier received contained the nose coil and two 14-in.-long pairs of fuel tubes from behind the nose. The cobalt foil, thermocouples, Calrods, and insulation were removed from the nose coil in the ORNL hot- . PERIOD ENDING MARCH 10, 1956 laboratory facilities. Five of the original six cobaToils were found and will be analyzed. The stripped nose coil and part of the heat exchanger assembly are shown in Fig. 8.1. The nose coil was sectioned with a hack saw, and the fuel was melted out of the coil and the two pairs of 14-in.- long fuel tubes in an argon atmosphere at a temperature of approximately 650°C. Thirty-five metallographic samples were obtained from the nose coil, heat exchanger, bellows, and fuel tubes. The locations of the specimens cut from the nose coil are shown in Fig. 8.2, and the locations of the other specimens are described in Table 8.3. The eleven specimens cut from the nose coil were examined metallographically in the as- polished unetched condition, and no evidence of attack was found. Wall-thickness measurements TABLE 8,3, NUMBER DESIGNATIONS AND LOCATIONS OF METALLOGRAPHIC SPECIMENS CUT FROM MTR IN-PILE LOOP NO. 3 Specimen No, L ocation of Specimen 280 281 283 284 285 286 287 288 289 290 291 Radial section 295 Radial section on 296 Radial section on 297 Radial section at 303 Radial section on 304 Radial section on 305 Radial section on 306 Radial section on 307 Radial section on 308 Radial section on 309 Radial section on Section of weld from heat exchanger U-joint Transverse section of bellows on heat exchanger from outlet side Transverse section of bellows on heat exchanger from inlet side Radial section at location of thermocouples 5, 6, and 7 on outlet side Transverse section of heat exchanger weld on outlet side Transverse section of end of helium sniffer line on outlet side Radial section at location of thermoccuple 21 Same as specimen 284 except on inlet side Same as 285 except on inlet side Radial section ot location of thermocouple 18 on cutlet side ]0% in, from end of helium sniffer outlet side 151/2 in. from end of helium sniffer outlet side 20 in. from end of helium sniffer ocation of thermocouple 23 on outlet side outlet side 30 in. from end of helium sniffer inlet side 7 in. from end of helium sniffer inlet side 12 in. from end of helium sniffer inlet side 17 in. from end of helium sniffer inlet side 271 in. from end of helium sniffer inlet side 25 in. from end of helium sniffer inlet side 30 in. from end of helium sniffer 183 ANP PROJECT PROGRESS REPORT ey PHOTO 16072 Fig, 8.1, Stripped Nose Coil and Part of Heat Exchanger Assembly of MTR In-Pile Loop No, 3, were taken on four of the samples from the coil and on the original tubing. Statistical variations in the wall thickness of the original tubing and variations resulting from bending of the tubing were established, Measurements taoken on the four samples from the irradiated nose coil fell 184 UNCLASSIFIED ORNL~LR-DWG 13033 264 1455 OF 263 261 262 269 1435 oF 270 1490 °F 271 273 1415 °F — 1390-1490 °F 274 ¢ OUTLET Fig, 8.2, Locations on Nose Coil from Which Metallographic Specimens were Taken. The number designations of the samples and the operating temperatures at various points are shown, within the established variations of the unir- radiated coil. The results of attempts to etch these specimens chemically and electrolytically have not been satisfactory. The etched specimens show what appears to be moderate intergranular attack to a depth of | mil, but this may be an etching effect rather than true corrosive attack. Specimens 264 and 271 are shown in Figs. 8.3 through 8.6 in the polished and etched conditions. Two etching techniques are being studied that show promise of providing more definable micro- structures, Specimens from the pump and the two straight sections of fuel tubes will be examined as soon PERIOD ENDING MARCH 10, 1956 Fig. 8.3. As-Polished Specimen 264 from Nose Coil of MTR In-Pile Loop No. 3. 50X. Fig. 8.4, Etched Specimen 264 from Nose Coil UNCLASSIFIED S RMG 311 of MTR In-Pile Loop No, 3. 250X. 185 ANP PROJECT PROGRESS REPORT Fig, 8.5. As-Polished Specimen 271 from Nose of MTR In-Pile Loop No, 3, 50X. Fig, 8,6, Etched Specimen 271 from Nose Coil of MTR In-Pile Loop No. 3, 250X. 186 as two cells can be decontaminated for cell- equipment servicing. The fuel from the two sections of fuel tubing will be analyzed chemically, and specimens of the tubing will be examined metallographically. EFFECTS OF RADIATION ON THE MECHANICAL PROPERTIES OF STRUCTURAL MATERIALS J. C., Wilson Solid State Division Stress-Corrosion Tests W. W, Davis J. C, Zukas J. C. Wilson Solid State Division Eight Inconel, helium-pressurized, tube-burst, stress-corrosion specimens were exposed to radi- ation in hole HB-3 of the LITR in a helium atmos- phere. The circumferential stress on each specimen was 2000 psi, ond various test temper- atures were used, Similar specimens are being tested out-of-pile to obtain control data. A plot of the log of the time-to-rupture vs the operating temperature during irradiation, Fig. 8.7, —_—— ORNL-LR-DWG {1772 T T T 0.040-in. WALL, 0.191-in. ID 3 | 1600 _ INCONEL TUBING IN LITR ! . | | | STRESSED BY INTERNAL GAS \ | PRESSURE TO A CIRCUMFERENTIAL L STRESS OF 2000 psi { } ~ o | L : ] | 2. 1550 OO ; 1 l uJ | ' : o« 5 | = i g o L . o E I © : 1500 : A A4 A — N w o 1430 o ; L ‘ o | L 10 20 50 100 200 500 1000 2000 TIME TO RUPTURE (hr) Fig., 8,7, Results of Tube-Burst Stress-Corrosion Tests of Internally Pressurized Inconel Tubing Exposed to Radiation in the LITR at a Stress Level of 2000 psi in a Helium Atmosphere, PERIOD ENDING MARCH 10, 1956 revealed an unusual degree of scatter in the data and led to an investigation of the condition of the specimen tubing prior to testing. Extensive non- destructive tests are being conducted on tubing for tube-burst experiments. All but four of 28 specimen tubes examined by Zyglo inspection were rejected because of pinholes and spongy regions. Metallographic inspection of 25 sections from the 24 rejected tubes resulted in the location of only one appreciable defect, Fig. 8.8, on the inside of the tubing. The types of defects that show up in nondestructive testing may have been partially responsible for the unusual scatter of the data obtained in the experiments described above. Several specimens that pass the non- destructive tests will be irradiated, after the out-of-pile stress-corrosion tests are completed, to study the effect of irradiation on rupture time in these inspected tubes. It is believed that the data will aid in estimating the seriousness of defects in the tubing in terms of creep life. The better quality tubing stock obtained for radiator fabrication will probably be used in subsequent experiments if the radiation effects indicate the need for more accurate determinations. The typical structure near the fracture of one of the in-pile tube-burst specimens is shown in Fig, 8.9. This specimen ruptured after 84 hr at 1500°F. The structure near the fracture of the only out-of-pile specimen available is shown in Fig. 8.10. This specimen fractured after 356 hr at 1500°F. A cantilever stress-corrosion rig similor to that used in the LITR? has been in operation out-of-pile for over 250 hr. Two earlier out-of-pile tests were not completed because of a weld failure in one case and a thermocouple failure in the other. Static sodium is used as a heat-transfer medium on the outside of the Inconel tube being tested in this experiment. The Inconel tube is filled with the molten fuel mixture NOF-ZFF4-UF4 (50-46-4 mole %). Preparations have been made for opening the similar apparatus irradiated previously in the LITR and for examining the Inconel specimen metallographically, Two more rigs of this design have been filled with fuel and will be operated in the future. Metallographic examination of a similar canti- lever apparatus that had been operated out-of-pile with helium on both sides of the tubes (rather than 2W. W. Davis et al., ANP Quar. Prog. Rep. Sept. 10, 1954, ORNL-1771, p 142. 187 ANP PROJECT PROGRESS REPORT UNCLASSIFIED RMG 1314 Fig. 8.8. Defect Found by Metallographic Examination on inside of Inconel Tubing from Stock Used in Tube-Burst Specimens, sodium or fuel) showed that some surface oxidation had taken place. There was some oxidation pene- tration into the grain boundaries, but this was only noticeable on the tension sides of the specimens. Two water-cooled, finned, tube-burst rigs for stress-corrosion testing of Inconel tubing, de- scribed previously,? have been filled with the fuel mixture NqF-ZrF4-UF4 (63-25-12 mole %). One of these will be operated in the LITF to check heat-transfer design calculations. Alternate Stress-Corrosion Apparatus C. D. Baumann W. E. Brundage Solid State Division A design of an alternate apparatus for stress- corrosion testing was described previously.4 With this apparatus the maximum temperature of the fuel 3J. C. Wilson et al., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL.-1947, p 165. 4w, E. Brundage and C. D, Baumann, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL.2012, p 184. 188 and of the specimen would be limited to 1500°F, The preliminary design was based on calcu- lations of the expected fission power, heat flow, and the limitations of fabrication. Since the thermal-neutron flux in hole HB-3 of the LITR was not known accurately, a mockup of the fuel and container was constructed, as shown in Fig. 8.11, and was irradiated for a short time at a position about 2 in. from the inner end of the hole. A cadmium-magnesium alloy (9 wt % Cd, 91 wt % Mg) was used to simulate the cross section of the fuel, and cobalt foils were used as the monitors., After irradiation the foils were removed, and the activity was measured in a 100% geometry counter, The measurements indicated a flux of 9 x 1012 neutrons/cm2.sec at the outer surface of the fuel and 5 x 1012 neutrons/cm2.sec at the inner surface of the annulus. Calculations based on these measurements indicate an unperturbed thermal- neutron flux at this position of 2 x 1013 npeu- trons/cm?.sec, The flux measurements permitted a refinement of the design, and the equipment now being fabricated PERIOD ENDING MARCH 10, 1956 Fig. 8.9. Structure Near Fracture in Inconel Tube-Burst Specimen That Was Internally Pressurized with Helium to a Fiber Stress of 2000 psi and Ruptured After 84 hr at 1500°F in Hole HB-3 of the LITR. 300X. Fig. 8.10, Structure Near Fracture in Inconel Qut-of-Pile Tube-Burst Specimen That Was Internally Pressurized with Helium to a Fiber Stress of 2000 psi and Ruptured After 356 hr at 1500°F, 300X. 189 ANP PROJECT PROGRESS REPORT ALUMINUM (LTI T ’ UNCLASSIFIED ORNL-LR-DWG 11773 POCKETS FOR COBALT FOIL T A T NTTT i - O.2821in. ,-{ — ~— 0.360in. i \\\ \\ \\\\\\\\\\\\\\\\ \\\\\\ \\\\\\\ \\ \\\ ‘-4—*4OOOIH —-———"1 Figo 8-]]. 9% Cd-Mg ALLOY Design of Capsule Used for Measuring Thermal-Neutron Flux in Hole HB-3 of LITR and the Flux Attenuation To Be Expected in the Stress-Corrosion Apparatus, is shown in Fig. 8.12. The platinum heater surrounding the fuel container will be used to supplement the fission heating to maintain the operating temperature, the equipment will act as a safety container in case of a leak or rupture of the specimen. Power generation in the fuel will be about 750 w/cm3. Postirradiation measurements will be made on the outside of the specimen tube, and the specimen will be examined metallographically for corrosion. It is expected that this apparatus can be used in the MTR with little modification. The outer container for MTR Tensile Creep Tests W. W. Davis N. E. Hinkle J. C. Wilson Solid State Division The MTR tensile-creep-test apparatus in which two lnconel tubes were stressed in tension to 1500 psi at 1500°F (in helium) was irradiated in hole HB-3 of the MTR for 33 days at an average power level of 30 Mw, Postirradiation sectioning revealed that both specimens had fractured. Tem- perature measurements taken during the test showed an abnormal temperature distribution after 518 hr for specimen 1, which failed outside the gage length, and after 188 hr for specimen 2, which failed within the gage length. Throughout the high-temperature region, both specimens had dark surface films. The absence of metallic 190 reflection from the fracture surfaces indicated that the specimens had not fractured during disassembly of the apparatus. Metallographic examinations indicated that attack by an unknown contaminant, alone or in combination with temperature ex- cursions, was responsible for the short rupture life. The corresponding out-of-pile control test has been delayed by leaks in the apparatus. Metallographic examination of irradiated specimen 2 revealed an intergranular fracture and necking at the point of failure. Photomicrographs of longitudinal sections showed considerable internal grain-boundary separation. The incidence of grain- boundary separation appeared to be greatest ad- jacent to the fracture and to decrease with increasing distance from the fracture, Surface grain-boundary separation was observed at the inside surface to a maximum depth of 4 mils in o few places. Photomicrographs of sections adjacent to the fracture show the presence of an unidentified film on the specimen surfaces. The film on the inside surface was thicker than that on the outside surface of the specimen tube. Fracture surfaces and surface grain-boundary separations exhibited the film material, but internal grain-boundary sepa- raticns did not, Sulfur attack, because of its well-known effects on nickel alloys, was thought to be a possible cause of the surface film and the intergranular UNCLASSIFIED ORNL-LR-DWG 12000 FILLING TUBE 1 | GAS TUBE ——-| m E O { INCHES oUT 1 p N R Fig., 8,12, Alternate Stress-Corrosion Apparatus, attack observed. It is important that the source of the contaminant be found so that it can be eliminated in future tests, No indications of sulfur were found in the insulation for the wiring or the furnace. An attempt to reproduce the surface film on the irradiated specimens was made by heating a similar Inconel tube in contact with sulfur for a short time at a temperature near 1500°F. Obser- vation of the region contacted by the sulfur revealed a large amount of grain-boundary pene- tration compared with the limited amount of surface film and surface grain-boundary separation in the irradiated specimen. Sulfur printing techniques were applied to the control sample without SUCCess. The Inconel tubing used as the specimen stock was subjected to ultrasonic and eddy-current tests, and indications of defects were obtained. Samples PERIOD ENDING MARCH 10, 1956 were cut from the regions classed as most de- fective, but metallographic examination did not confirm the defects. mental results of the tests will be continued. The analyses of the experi- Pneumatic Stressing Device A. S. Olson J. C. Wilson Solid State Division A pneumatic stressing device that will provide a powerful, compact means of stressing creep specimens with loads as great as 10,000 Ib has been designed and is being constructed. The device can be used for both laboratery and in-pile tests of creep, stress corrosion, and stress relax- ation. Work continues on the design of a suitable recording creep extensometer. Ductility of Nickel Alloys J. C. Wilson Solid State Division T. C. Price Pratt & Whitney Aircraft The effect of temperature on the ductility of nickel alloys in the temperature range 800 fto 1400°F is being studied, with the effect of strain rate vs temperature being measured first, To date, fairly complete data have been obtained for Monel, and work on Nichrome and Inconel is under way. The gage-length portions of the Monel tension specimens were 0.1500 * 0.0005 in. in diameter and 0.060 * 0.05 in. in length. Strain rates of 2 and 0.002 in./min were used, and the test temperatures ranged from 800 to 1400°F. The ultimate and yield strengths at both strain rates appeared to vary linearly with temperature. There was, however, a large nonlinear decregse in elongation and percentage reduction of area in the region 900 to 1000°F at the faster strain rate, which was not observable at the slower strain rate. These results are illustrated in Figs, 8.13 and 8.14. At the slower strain rate the ductility values were lower than those found at the faster strain rates. At about 1000°F the ductility values nearly coincide for both strain rates, and therefore the processes responsible appeared to be independent of strain rate at this temperature. At higher and lower temperatures the curves tend to diverge, with greater ductility being found at the faster strain rate, 191 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL—LR-DWG {14685 l \ STRAIN RATE 2.0 in./min %, ELONGATION % REDUCTION OF AREA | AT FRACTURE % REDUCTION OF AREA " 0.50 in. FROM FRACTURE 70 ELONGATION (%) OR REDUCTION OF AREA (%) o ; 1 J 800 90¢ 1000 400 1200 41300 (400 4500 TEMPERATURE (°F) Fig, 8,13, Effect of Temperature on Tensile Ductility in Monel at a Strain Rate of 2 in,/min, EXPERIMENTAL STUDIES OF REACTOR MATERIALS AND COMPONENTS W. E. Browning Solid State Division Effect of Radiation on Static Corrosion of Structural Materials by Fused Salts W. E. Browning H. L. Hemphiil Solid State Division Irradiations in the MTR of Inconel capsules containing various fused-salt fuels have been con- tinved in order to study the effect on the Inconel of fissioning in the fuel. Since earlier tests® had shown that irradiation for two weeks in the MTR had no significant effect on corrosion, the irradi- ation periods used for the capsules tested during the past year have been of six or nine weeks duration, Seven capsules filled with Nc:F-ZrF4 plus 2 mole % UF4, 2 mole % UF_, or 4 mole % UF4 were irradiated at 1500°F :}or the longer periods and are now being sampled for metallo- graphic and chemical analyses, Additional Inconel capsules filled with NaF-ZrF -UF, (50-46-4 mole %) are being prepared for irradiation in the W. E. Browning, G. W, Keilholtz, and H, L. Hemphill, ANP Quar. Prog. Rep. Dec. 10, 1954, ORNL-1816, p 120. 192 UNCLASSIFIED ORNL-LR-DWG {1686 70 T ‘ [ ‘ w I i | STRAIN RATE 0.002 in./min 60 O % ELONGATION ® % REDUCTION OF AREA AT FRACTURE S50 ———— 1 & % REDUCTION OF AREA 0.5Cin. FROM FRACTURE 40 T Ti | 30' 4 \ ‘ 20 ELONGATION (%) OR REDUCTION OF AREA (%) 800 200 1000 HO0 1200 1300 400 1500 TEMPERATURE {°F} Fig., 8.14. Effect of Temperature on Tensile Ductility inMonel at a Strain Rate of 0,002 in,/min, MTR at 1800°F. A duplicate capsule is being tested out-of-pile for comparison, A multicapsule facility has been prepared for exposures in a high-flux position in the MTR. The temperature-control system, de- scribed previously,® has performed satisfactorily for more than a year. Three identical units are now in service, Various unforseen disturbances have been handled, including one in which drops of water were sprayed intermittently on a capsule operating at 1500°F, without the occurrence of excessive temperatures. The improved thermocouples, also described pre- viously,® have also served satisfactorily. Only one of the 25 improved thermocouples that have been used in the MTR has failed. The one that failed was one of three that were sprayed with water while hot after six weeks of irradiation; the other two did not fail. Mockup tests have been run on Hastelloy B capsules under simulated MTR conditions to de- termine whether oxide scaling on the outside can revised be prevented in order to control contamination of the facility., Chrome-nickel plating of the outside ‘w. E. Browning, G. W. Keilholtz, and H. L. Hemphill, ANP Quar. Prog. Rep. March 10, 1955, ORNL-1864, p 146. of the capsules gave an oxide coating in air at high temperatures that was adherent during thermal cycling. The coating was similar to that formed on Inconel in that it did not scale, and it appears to be promising for use with the present capsule designs, Holdup of Fission Gases by Charcoal Traps W. E. Browning Solid State Division C. C. Bolta Pratt & Whitney Aircraft The apparatus described previously? is being used for studying the various factors that affect the performance of charcoal traps for the holdup of fission gases. Both nitrogen and helium are being tested as purge gases, and radiokrypton is being used to simulate the fission gases. The trap is a 13-in.-long, 2-in.-dia, sched-40 stainless steel pipe filled with ¥ |b of 8- to 14-mesh 4 Columbia activated charcoal. For each test the charcoal trap is filled with purge gas, the manifold is evacuated, and radic- krypton is allowed to flow into the krypton chamber, The chamber is then sealed and the rest of the manifold is evacuated. Purge gas is then allowed to push the radiokrypton into the charcoal trap at a constont flow rate of 5 ft3/hr. By measuring the activity of the gas, the concentration of radiokrypton in the effluent gas and therefore its partial pressure can be determined. The relative activity of the effluent gas is measured as it passes through a small cell that has as one wall the end window of a GM counter. The gas activity is registered on a log counting-rate recorder. A theoretical analysis of the adsorption process in a number of theoretical charcoal-filled traps, N, connected in series and occupying the total volume of the trap has been made. The theoretical problem is similar to the continuous-dilution-tank problem in chemical engineering and applies only to systems where dilution processes, and not diffusion or adsorption, are rate limiting. The rate of removal of radickrypton in each trap is first order with respect to its partial pressure and therefore its concentration in that chamber; that is, dP F dt km A series of N differential equations of this type is solved simultaneously for the N chambers, and PERIOD ENDING MARCH 10, 1956 the general equation for the Nth chamber is N 4 p(N=1), (N 1) N7 AF t (]) Pn - e-—NFt/km , (N - D! (km) where P = partial pressure of radiokrypton, atm, A = amount of radickrypton injected into first theoretical chamber, em3.atm, N = number of theoretical chambers, F = flow rate of purge gas, cm3/min, t = time after injection of pulse, min, k = slope of the linear isotherm x/m = kp for radiokrypton in the mixture of inert purge gas and radiokrypton, em® (at STP) per g-atm, m = amount of adsorbent (charcoal) in trap, g, x = amount of gas (radiokrypton) adsorbed iso- thermally, cm® (at STP). The term kim is a measure of the adsorptive capacity of the charcoal trap for radickrypton in the presence of purge gas at a given trap temper- ature. Parameters km and N are chosen to fit Eq. 1 to the experimental data. The experimental results obtained with nitrogen as the purge gas through traps at four different temperatures are compared with analytical data in Figs. 8.15 through 8.18. In these illustrations the use of one trap is compared with the use of two identical traps in series. The trap temper- atures studied were 16, 5, =51, and ~110°C. The deviations between corresponding curves are within experimental error, In each case the activity injected was determined by integration of the activity-vs-time curves; all results were normalized to the same area under the curve. |t may be seen that, for a given trap geometry, the holdup times (i.e., the time required for activity to break through, to peak, and to disappear) increased with decreased trap temperature. This result could be predicted, because the energy of a gas molecule is less at low temperature, and it therefore cannot so readily desorb from the charcoal surface into the moving gas stream. Also, it may be seen that the maximum concentration of radiokrypton in the effluent gas is lower at lower trap temperatures. ’G. W, Keilholtz, W. E. Browning, and C. C. Bolta, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 183 and Fig. 8.1. 193 ANP PROJECT PROGRESS REPORT UNCLASSIFIED 105 ‘ ‘ ORNL-LR-DWG 13046 & EXPERIMENTAL, ONE-TRAP SYSTEM AT16°C = LA ANALYTICAL, ONE-TRAP SYSTEM 51— WN=7.83, 4m=1.487 x 10% cm? - - @ EXPERIMENTAL; TWO-TRAP SYSTEM AT{5°C " — 0 ANALYTICAL; TWO-TRAP SYSTEM — > | N=15,00, /m; 3.29x10%cm® | n o ™ o, Pt RELATIVE EFFLUENT ACTWVITY (counts/min) Q M h_tL+ T:‘_'_“: > 0 10 20 30 40 50 60 70 TIME (min) Fig. 8.15. Holdup of Radiokrypton in Nitrogen- Purged Charcoal Traps as a Function of Temper- ature, One trap held at 16°C; two traps in series held at 15°C, The use of two identical traps in series in- creased the time required for the activity to peak to almost twice that for one trap and increased the break-through time to two to three times as long as for one trap. The time required for the activity to disappear from the effluent gas was also increased, and the maximum radiokrypton concentration was reduced. Similar data were obtained with helium as the purge gas. The shapes of the curves (Fig. 8.19) are somewhat uncertain, however, because the accuracy of the data was limited by the sensitivity of the flowmeter in its lower range when used with helium. Several runs were made at each temper- ature, and the holdup times measured were sig- nificant, even though the exact shape of the curve may be in doubt. The curves show that radio- krypton is held up for much longer periods of time 194 UNCLASSIFIED 5 ORNL-LR-0OWG 13047 s E—— F : J‘ T A EXPERIMENTAL; ONE -TRAP SYSTEM AT 5C | & ANALYTICAL ; ONE-TRAP SYSTEM — - — N=T.27, km =2.05 x 10% em® —— | —— '@ EXPERIMENTAL; TWO-TRAP SYSTEM AT oc 2 -c ANALYTICAL ; TWO-TRAP SYSTEM ——t—— : N=142, km = 4.82 x 10° cm® ‘ 10 5 — : \ ;X— - o - i K’FF’;‘AEP ,Two TRAPS , §x 4%* — RELATIVE EFFLUENT ACTIVITY {counts/min) 0 10 20 30 40 50 60 70 TIME (min) Fig. 8,16, Holdup of Radiokrypton in Nitrogen- Purged Charcoal Traps as a Function of Temper- ature. One trap held at +5°C; two traps in series held at 0°C. in the presence of flowing helium than it is in the presence of flowing nitrogen., Since the nitrogen is adsorbed in large amounts on charcoal, the charcoal surface is not so free to adsorb the krypton. Helium will pass over charcoal with very little adsorption, and hence krypton is retained on the charcoal for longer periods of time in the presence of flowing helium. The experimental curves of Figs. 8.15, 8.17, and 8.18 are plotted together on Figs. 8.20 and 8.21] for comparison. The time required for maximum activity to be reached and the height and duration of the peaks are clearly indicated as a function of temperature. The analytical parameters N and km for various temperatures, with nitrogen as the purge gas, are compared in Fig. 8.22. |t appears that the number of theoretical chambers N is a characteristic of the geometry of the charcoal trap, UNGLASSIFIED 5 ORNL-LR-DWG 13048 & EXPERIMENTAL, CNE-TRAP SYSTEM AT —51°C 4 ANALYTICAL, ONE-TRAP SYSTEM N=7.43, km=1.09 x 10° cm® ® EXPERIMENTAL; TWO-TRAP SYSTEM AT ~50°C o ANALYTICAL, TWO-TRAP SYSTEM N=15.0, km =1.80 x10° ctm ONE TRAP —TWQO TRAPS RELATIVE EFFLUENT ACTIVITY (counts /min) 0 40 80 120 160 200 240 TIME (min) Fig. 8.17. Holdup of Radiokrypton in Nitrogen- Purged Charcoal Traps as a Function of Temper- ature, One trap held at =51°C; two traps in series held at ~50°C. since N does not vary for the different temperatures studied. This should be true if the model used to derive the expression for the curve has physical significance. For two traps in series, N is twice as large as for one trap, which indicates that N may be a first-order function of the length of the trap divided by the diameter, It is indicated in Fig. 8.22 that &m increases exponentially as 1/T increases (decreasing temper- ature); actually, it is & that increases, since the curves are for constant charcoal mass m. |f km, the parameter obtained from holdup data by using Eq. 1, is actually the slope of the linear isotherm multiplied by the amount of charcoal, this straight- line relationship should exist. |t therefore appears that the parameter is proportional to the slope of the isotherm, but whether the proportionality factor is unity has not yet been established. The time, t__ , required for the maximum partial 8. E. Guss, Solid State Div. Semiann. Prog. Rep. Aug. 31, 1955, ORNL-1944, p 18, PERIOD ENDING MARCH 10, 1956 UNGLASSIFIED 5 ORNL-LR-DWG 413049 0 ———— ~— 4 EXPERIMENTAL; ONE-TRAP TRAP SYSTEM AT 10 c: 4 ANALYTICAL; ONE-TRAP SYSTEM T N=8.03. 4m=6.74 x 105 cm ] e EXPERIMENTAL, TWO-TRAP SYSTEM AT —110°C____ o ANALYTICAL, TWO-TRAP SYSTEM al wm=15.1, km=152x10%em® RELATIVE EFFLUENT ACTIVITY {counts/min) 1000 1500 TIME {min) Fig. 8.18. Holdup of Radiokrypton in Nitrogen- Purged Charcoal Traps as a Function of Temper- ature, Traps held at ~110°C. pressure of radiokrypton to be reached in the effluent purge gas can be derived from Eq. 1 (N - 1)km (2) Lo = ————— .+ m NF The maximum partial pressure, P _ oxr Can then also be obtained from Eq. 1: 3) P _N(N — DY-14 -(N=1) max (N = D! km in Eq. 2 the time required for the activity to peak depends on km and only very slightly on N, All traps with the same amount of charcoal, m, at the same temperature should have the same time-to- peak, regardless of the shape of the trap. This conelusion is consistent with the data of Guss.® The height of the activity curve depends more on N in Eq. 3 thon on km; therefore a long thin trap would give a high narrow peak, and a short large- diameter trap would give a broad peak with a low maximum concentration. Work is in progress to determine the relationship between N and trap shape, and further studies of 195 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 43050 & EXPERIMENTAL; ONE - TRAP | | SYSTEM AT —5%C 4 ANALYTICAL; ONE-TRAP — SYSTEM LA =121, | /rm— 9.0 x 404 em? — RELATIVE EFFLUENT ACTIVITY {counts/min) A EXPERIMENTL\L L - — Tt T R STEM N T ‘A . ONE ~ T AP SY TEM 3‘,"‘\ —14.85, ] AT . —/rm = 7.09 x 405 em? A ANALYTICAL | - - " ONE-TRAP SYSTEM : RELATIVE EFFLUENT ACTIVITY (counts/min} o 100 200 300 TIME (min] 500 800 Fig. 8.19. Holdup of Radiokrypton in Helium- Purged Charcoal Traps at -5 and —~50°C, other conditions are planned. The results should provide engineering data for trap design and may confirm further the theoretical relationship given above. LITR Vertical In-Pile Loop W. E. Browning H. E. Robertsor M. F. Osborne R. P. Shields W. R. Willis Solid State Division D. E. Guss, U.S. Air Force Fabrication, assembly, and preirradiation testing of a new forced-circulation Inconel loop for oper- ation in a vertical position in the LITR was completed. This loop is essentially the same as the one operated previously in the LITR,? with 196 UNCLASSIFIED 5 ORNL-L.R—DWG 13054 RELATIVE EFFLUENT ACTIVITY (counts/min) 0 100 200 300 400 500 600 700 TIME {min} Fig. 8,20, Comparison of Experimental Data on Holdup of Radickrypton in Nitrogen-Purged One- Trap Systems at Various Temperatures, minor modifications based on experience with the first loop, The brush-type motor used previously has been replaced with a three-phase induction motor. The speed of the induction motor will be regulated by the use of a variable-frequency power supply. The tip of the loop will extend more than 4]/ in, below the position of maximum flux in the reqctor to obtain a greater temperature differential than that achieved previously. Operation of the loop in the LITR is planned for April. Instrumentation for ART Of-Gas Analysis W. E. Browning Solid State Division Design conditions for the gamma-ray spectrometer for the ART off-gas system have been established. Four scintillation detectors will be used, one at 9G. W. Keitholtz et al., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 164, PERIOD ENDING MARCH 10, 1956 UNCLASSIFIED ORNL-LR—-DWG 13052 S — 10° T - [ ' . I — — 5 —— | ! e | } - 2 — < L ; : E — — — . I 5 T = - £ - — | - ' 3 . 3 o | ] = ?\7 - . E o = '_ Q < _ _ ] _ = - ; _ = - L | 0 . e I L re i - e o Ll N e _ wl o W I —10°C z e ———— i : ~e__ ; w : | X ‘ o e Do e~ e e | g — ] . '\ \'\- = . ] '\'_"'- . e S B S e - ‘ 1 1 i 1 | | | 500 600 700 800 900 1000 GO 1200 1300 1400 TIME (min) Fig. 8,21, Comparison of Experimental Data on Holdup of Radiokrypton in Nitrogen-Purged Two-Trap Systems at Various Temperatures, each of the inlets and outlets of the charcoal traps in the off-gas lines from both the fuel system and the reactor cell. Provision will be made for moving the shielded detector up to 10 ft from the most active off-gas sampling tube, under normal con- ditions, and closer, if necessary. A fourechannel scintillation spectrometer will be used to monitor the detectors. Inconel as a Thermal-Neutron-Flux Meonitor D. E. Guss, U.S. Air Force A study is being made of the feasibility of using Inconel as a thermal-neutron-flux monitor in the vertical in-pile loop and in static corrosion cap- sules. If Inconel can be used as the monitor for Inconel systems, it will not be necessary to add monitors to the experimental system. Furthermore, Inconel has the proper activation and physical properties, whereas the various cobalt monitors currently available either become too radioactive during long-term irradiation or melt at the high temperatures required by the experiments, The cobalt monitors also require extensive activation analysis before they can be used. There are two constituents of Inconel which can be used for flux monitoring: cobalt, about 0.14% in lnconel, and chromium, about 15.5%. Cobalt has the disadvantage that it must be chemically separated from the Inconel after irradiation to remove the Fe3? activity, which has gamma rays of 1.1 and 1.3 Mev. Since cobalt is present in such small quantities, there was, at first, some doubt about its homogeneity, but activation analyses of several pieces of Inconel stock show 197 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 13053 TEMPERATURE (°C} ° ~%0 -100 e 3%10 2 : 2 N FOR TWO TRAPS o ; B ) - N FOR ONE TRAP —— - 5 5 W " 5 2 2 & ~x 5 10 | 5 5 2 2 4 1o - 0.0030 0.0040 0.0050 0.0060 ek Fig. 8,22, Comparison of Analytical Constants N and km for One- and Two-Trap Systems as a Function of Temperature, that, within a given batch of Inconel, the cobalt is homogeneous enough for monitoring purposes. The Cr39 isotope, 4.4% naturally abundant, captures neutrons and goes to Cr3', with a thermal cross section of about 15 barns. The Cr3! isotope has a half life of 26.5 days, and it decays by K-capture, with an associated 0.32-Mev gamma ray in about 9% of the disintegrations. Scanning of irradiated Inconel with a gamma-ray spectrometer has shown that the Cr3! peak is large enough for a good value of the chromium agctivity to be obtained by area analysis. The cadmium ratio of chromium in the lattice of the LITR was measured to be 29, which is consistent with the value to be expected for a 1/v absorber. The cross section is known to be about 15 barns, but a more accurate measurement must be made. The disadvantage in using chromium for long-term irradiations is its relatively short half life. This limits the time of irradiation for which chromium could be effectively used as a flux monitor to one or two months; after this length of time the activity is too near the saturation point for chromium to be 198 effective as an integrated-flux monitor. For longer irradiations, therefore, the cobalt must be used to give a more accurate value for the thermal- neutron flux. Measurement of MTR Flux near Tip of In-Pile Loop No. 3 D. E. Guss, U.S. Air Force The thermal-neutron flux near the tip of MTR in-pile loop No. 3 was measured by means of cobalt foils attached to the nose coil at the positions shown in Fig. 8.23. The values obtained at the various positions are given in Table 8.4. Fast-Flux Measurements in Hole 19 of ORNL Grophite Reactor J. F. Krause Pratt & Whitney Aircraft Two flux measurements were made in hole 19 of the ORNL Graphite Reactor, which has been used extensively for irradiations of organic materials. The first traverse was made with the threshold reaction S$32(n,p)P32 for the flux measurement above 2.9 Mev, and the second was made with Co°?(n,y)Co%? reaction for the thermal-neutron-flux measurement. Each traverse was fairly constant from the center of the reactor to a point 30 in, from the center. Over this range the flux above UNCLASSIFIED ORNL-LR-CWG 13054 NO. 20 NO. 17 NO. 25 NO. 23 Fig., 8.23. Diagram of Nose Coil of MTR In-Pile LLoop No., 3 Showing Positions of Cobalt Foils, PERIOD ENDING MARCH 10, 1956 TABLE 8,4, INTEGRATED AND AVERAGE THERMAL-NEUTRON FLUX AT VARIOUS POSITIONS ON NOSE COIL OF MTR IN-PILE LOOP NO. 3 Foil No. Integrated Thermal-Neutron Flux Average Thermal-Neutron Flux (neutrons/cm2) (neufrons/cmz-sec) 15 3.85 x 1017 4.06 x 1013 17 7.08 x 10'7 7.46 x 102 20 8.20 x 107 8.64 x 1013 23 5.90 x 1017 6.22 x 1013 25 418 x 1019 4.41 x 1013 2.9 Mev was approximately 2.5 x 10° neu- and UFS"'. In all likelihood the bonding in the trons/cmz-sec-w, and the thermal-neutron flux was 2.4 x 10° neutrons/cm?.sec-w. All measurements were made in the ¥-in.-wall aluminum specimen cans ordinarily used in hole 19. These cans were surrounded by % , in. of flowing water, which acted both as a hydraulic fluid for moving the cans and Previous measurements in hole 19 were made during the irradiation of specific ma- terials that may have altered the flux. Similar measurements have been started in hole HB-3 of the LITR, where the stress-corrosion tests are being run on Inconel in contact with fluoride fuel mixtures. » CHEMICAL EFFECTS OF NUCLEAR REACTIONS M. T. Robinson Solid State Division as the coolant. Effects of Fission Products on Properties of Fluoride Fuels M. T. Robinson Solid State Division J. F. Krause Pratt & Whitney Aircraft Three types of radiation effects can be dis- tinguished in fused fluoride salt fuels such as NoF-ZrF4-UF4. Radiation damage occurs as a result of bombardment of the fuel with high-energy particles and electromagnetic radiation; bulk thermal changes are caused by the large amount of energy dissipated in the fuel; and chemical changes result from the replacement of uranium by a mixture of more than 30 other elements. The fuel consists primarily of ions, some simple, such as Na* and F~, and some complex, such as ZrFs" complex ions is largely ionic. For example, the method of Pauling'® shows a Zr—F bond to be about 75% ionic, In such a system, radiation- damage effects should be very small, since there are neither covalent bonds to sever (as in H,0 or organic compounds) nor a solid lattice to disrupt. Bulk thermal effects, while they cause a great deal of difficulty in the interpretation of continyous in-pile measurements, are not par- ticularly important in corrosion studies, where the fuel-metal interface temperature is controlled, and are even less significant in turbulently flowing systems, The yields of the various chemical elements in thermal-neutron fissioning of U235 gs q function of time have been calculated'? and used to calculate the average valency of the fission-product mixture. The chemical form each element is likely to adopt in a high-temperature fused fluoride fuel is dis- cussed below: Elements in group 0, krypton and xenon, will occur only in the elementary form, Elements in groups |A, |IA, lIIB, and IVB will adopt their usual group valencies: +1: Rb, Cs +2: Sr, Ba +3: Y, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy +4: Zr ]OL. C. Pauling, Nature of the Chemical Bond, p 70, 2d ed., Cornell Univ. Press, Ithaca, 1940. ”M. T. Robinson and J. F. Krause, Solid State Semiann. Prog. Rep. Feb, 29, 1956, ORNL-2051 (to be published). 199 ANP PROJECT PROGRESS REPORT There is a possibility that, if the medium becomes sufficiently oxidizing, some Ce(lV) will be present, particularly since it may be stabilized by the formation of some complex ion, like CeFS‘. Similarly, in a strongly reducing medium, some Eu(ll) is to be expected. However, the low yield of europium makes this a relatively unimportant point, The transition metals beyond zirconium in the periadic table are all relatively noble in com- parison with chromium. Accordingly, it has been observed that ruthenium and niobium deposit from fluoride fuels onto the Inconel container.'? The sparse thermodynamic data'? indicate that chromium will reduce germanium, arsenic, and all the metals from niobium through antimony to the elemental state. Where valencies are required for these metals, the following values have been chosen: +1: Ag, In +2: Ge, Te, Pd, Cd, Sn +3: As, Nb, Mo, Ru, Rh, Sb These are, in general, the lowest known valencies which have been reported for Fluorides that exhibit these valencies are not known in every case. The valencies to be expected for Se, Te, Br, and | depend strongly on conditions in the fluoride each element. ORNL-LR-DWG {4720 IRRADIATION TIME (hr) 003 0102 05 41 2 5 10 20 50 100 500 2000 T T 4.5 e I 4.0 —— ‘ ; : i : O e T uw —T i . : el g 35 b, Ru, AND Mo LOST ] L z N L ‘ ALC"NOBLE’iM TALS LOST]|! : ) Hl : > | sl 3.0 1 L _ 2T LT | ‘ T Ll & : ‘ 2 T ] 1 g 25 = ‘ ] | 1B 20 1A/ —_ | T 1.5 L ‘ A | i : T 1 T o : L 11, L s 02 5 16°2 5 12 5 10 IRRADIATION TIME {sec) 02 5 10° 2 Fig. 8.24. Average Fission-Product Valencies in Irrodiated Inconel Capsules Containing Fused Fluoride Fuel Mixtures in the System NaF-ZF - UfF ‘l 200 melt. Under moderate reducing conditions the expected states are the normal anions: -2: Se, Te -1: Br, 1 If the conditions are mildly oxidizing, the ele- mental forms are to be expected (valency 0). Under stronger oxidizing conditions, positive va- lencies may be anticipated. The average valencies of the fission-product mixture under various conditions are plotted in Figs. 8.24 and 8.25 as functions of irradiation time. The region above a valency of 4 presents reducing conditions, whereas, in the region below valency 4, oxidizing conditions exist, in com- parison with the original uranium valency of +4, The curves marked ‘‘inert container’’ are based on the assumption that in capsules (Fig. 8,24) all krypton and xenon descendants are retained and that in reactors (Fig. 8.25) the descendants of rare-gas isotopes of half life greater than 10 min ]2M. T.Robinson, T. H. Handley, and W. A, Brooksbank, Solid State Semiann. Prog, Rep. Aug. 31, 1955, ORNL.- 1944, p 17. 131, Brewer et al., p 76-192, in The Chemistry and Metallurgy of Miscellaneous Materials, Tbermo:ln_,?/namics, NNES 1V-198, ed. by L. L. Quill, McGraw-Hill, New York, 1950; W. M. Latimer, Oxidation Potentials, Prentice Hall, New York, 1938. "y ORNL-LR-DWG 14712 IRRADIATION TIME (hr) 003 0402 05 1 2 5 40 20 50 1C0 500 2000 a5 |INeRT conTa 4.0 i i No AP Ry 5T 5 | . 235 “|il* Nb, Mo, AND Ru LOST > . 5 30 | — ALL"NOBLEMETALS LOS = E ‘ ! u | : : I _j i 425 . } } _ Kr AND Xe ISOTCPES LOST IF r‘/ =140 min ' | 2 i 5 02 5 1©0°2 5 1f2 5 10 IRRADIATION TIME {sec) 102 5 1002 Fig. 8.25. Average Fission-Product Valencies in Circulating-Fuel Reactors Operated with Fused Fluoride Fuel Mixtures in the System NaF-ZrF,- UF,. are lost completely. The left-hand branch of each curve, up to valency 4, is based on the elemental state of the halogens and chalcogens. The right- hand branch is based on the anionic forms of these elements. In the intermediate branch, the halogens and chalcogens are progressively reduced from the elemental forms to the normal anions. The curves marked ‘““Nb and Ru lost’’ are based on the assumption that these two metals deposit on the container walls as elements, Where the valency is below 4, the halogens and chalcogens are taken as the elements; where it is above 4, these nonmetals are taken as the normal anions. The great importance of molybdenum in determining the average valency of the fission products may be seen from the remaining curves, which show the average fission-product valency for loss of Nb, Mo, and Ru and for loss of all the ‘‘noble” transition metals. Curves showing the predicted excess ‘‘fluorine” present in fluoride fuels as a result of the low average valency of the fission products are pre- sented in Figs, 8.26 and 8.27. The right-hand axis is marked with the concentration of excess Cr*? jon required to balance the excess oxidizing - ORNL-LR-DWG 11718 IRRADIATION TIME (hr) 03 4 35 10 20 50 100 500 2000 5000 2000 1000 ‘ 3 , 000 c ol D § "0 _TuERMALFLUX=2x10" 500 = o n/cm®-sec T —~ s O f E 20 : 200 E_ o 100 L S 100 1 T 50 5 : 50 o o o 20 L & 20 9 5 10 S 10 02 5 1%z 5 132 5 W2 5 0 IRRADIATION TIME {sec) Fig. 8.26. Theoretical Prediction of Excess “’Fluorine’" Present in Fused Fluoride Fuel Mixtures in the System NaF.ZrF,-UF, os o Function of f{rradiation Time in Inconel Capsules in the MTR, PERIOD ENDING MARCH 10, 1956 power of the fuel. It is very evident from these curves that a substantial increase in the corrosive attack of the fuel on the container may be expected under irradiation, as compared with the ordinary thermal corrosion, and that the attack will become more serious as the fission rate (w/cm?) is increased. This results from the instability, in contact with Inconel, of ions of molybdenum and ruthenium. Other things being equal, the effect is expected to be more serious in systems in which the fission gases are removed with high efficiency than in systems where they are retained. Contrary to these theoretical predictions, cor- rosion in irradiated fluoride fuel-Inconel systems has never been observed to be increased in com- parison with corrosion in unirradiated controls. No increased penetration of the irradiated metal had been observed. The amounts of container metals (Cr, Fe, Ni) found in irradiated fuels are, in general, much the same as, or even less than, the amounts found in unirradiated controls. It is evident that the excess oxidizing power of the irradiated fuel, which has been expressed here in terms of ‘‘fluorine,’’ is accommodated without increased attack on the container. Some of the - ORNL-LR-DWG 11717 IRRADIATION TIME (hr) 5 10 20 50 100 200 500 (000 2000 1000 1000 500 FUEL: 5.5mole % UF, POWER: 60 Mw 500 200 200 100 50 20 EXCESS “FLUORINE” {ppm) EXCESS Cr*t {ppm) w* 2 5 0% 2 5 w0f 2 5 10 IRRADIATION TIME (sec) Fig. 8,27, Theoretical Prediction of Excess ““Fluorine’”” Present in the System NaF-ZrF,-UF, as a Function of {rradiation Time in a Circulating- Fuel Reactor. 201 ANP PROJECT PROGRESS REPORT ways in which this excess oxidizing power could possibly be absorbed are discussed below. MoF,. ~ If only a small part of the total mo- lybdenum remained in solution as Mo(VI), it could absorb the entire excess oxidizing power. The required amount of molybdenum in solution would, however, be several hundred parts per million and would lead to far more dissolved MoF , than could be expected to be stable in the presence of Inconel. Fluorides of Se, Te, Br, and I. — Since Se, Te, Br, and | all form volatile fluorides, such as SeFé, TeF,, BrF,, and ”:5’ a portion, perhaps two- thirds, of the oxidizing power of the fuel could be absorbed in such compounds. These com- pounds, like MoF,, are strong oxidizing agents, and their existence in the fuels can be only transitory, unless they escape as gases. The results of capsule irradiations argue against this possibility. CeF,. — As was suggested above, at least some of the excess oxidizing power might well be absorbed as Ce(lV), in the form of a complex ion, such as CeF_~. Such an ion would be present in comparatively small amounts and might well serve as a ‘‘tracer’’ for the very similar ZtF_-— ion. The yield of cerium is such that about one-fourth of the oxidizing power could be thus absorbed in systems from which the rare gases were lost and almost one-half in systems in which they were retained, UF.. — Another way of absorbing the oxidizing power of the fuel would be the oxidation of U(IV) to U(V): UF.~ + F = UF, + F- In the fuel of the ART, this would require oxidarion of up to 2% of the uranium after 500 hr of operation, Solid Fluorides. — It is perhaps possible that the ‘“‘noble’’ metal deposit is not metallic and that it contains polymerized subfluorides, (MoF)x. These materials would reduce the oxi- dizing power of the fuel aobout one-half in the reactor case. such as No one of the above suggestions can account for all the excess ‘‘fluorine’” released by the fissioning uranium. In every case, a substantial 202 oxidizing power would remain. That the excess oxidizing power causes no increase in corrosion argues convincingly that the deposited fission- product layer effectively inhibits corrosion of Inconel by the fuel. The effect of the fission-product mixture on the physical properties of the fuel has also been studied. The well-known theory of Frenkell4 concerning the viscosity and electrical conductivity of fused salts was used as a starting point, The viscosities of the fused salts are determined by the larger of the ions present, which in the case of the fuels being considered are species such as UFS' and ZrFs". The fission process de- creases the number of these large ions by replacing them with the small ions Rb*, Ba**, La**?, etc., none of which are likely to be complexed. There- fore a small decrease is expected in the viscosity of the fuel with increasing time of irradiation. This decrease in viscosity will be exceedingly small, however, because of the relative rarity of UFS" ions (about one complex anion in four), the low burnups envisioned in aircraft reactors, and the high fission yield of zirconium, There will be a similar small increase in the electrical con- ductivity of the fuels. The effect of the fission products on, thermal conductivity cannot be esti- mated because of the lack of a suitable theory, but any effect will certainly be very small, The molar heat capacity of the fuel may be taken as independent of composition, to a first approxi- mation. The only effect of fission will be to increase slightly the effective molecular weight of the fuel through loss of the noble gases, mo- lybdenum, ruthenium, etc. Thus the specific heat of the fuel will decrease very slightly as irradi- ation proceeds. The eftect on corrosien is, therefore, considered to be the only important effect of the fission process on the properties of the fused fluoride fuels, Theory predicts a substantial increase in the corrosion rate as irradiation continues, but the absence of such an effect in all experiments to date leads to the suggestion that deposition of niobium, ruthenium, and (presumably) molybdenum on the container effectively inhibits the attack of the fuel on Inconel. ]4.]. Frenkel, Kinetic Theory of Liquids, Clarendon Press, Oxford, 1946. p 439, Use of Natural Lithium in Fluoride Fuels Circulated in In-Pile Loops M. T. Robinson Solid State Division J. F. Krause Pratt & Whitney Aircraft The use of lithium in the fluoride fuel system NaF-KF-LiF-UF4 presents a problem created by the nuclear reaction 6Li* + n—> 4He + 3Tt The concentration of the tritium ion in the fuel mixture NcF-KF-LiF-UF4 (11.2-41-45.3-2.5 mole %) as a function of time and thermal-neutron flux is shown in Fig. 8.28. The tritium-ion concentration is expressed as the pT of the solution and is defined by the following equation: pT = ~log (gram-ions of T* per liter of solution) . PERIOD ENDING MARCH 10, 1956 On the right-hand axis is shown the equivalent concentration of chromium in the fuel, which would be produced by the reaction 27 + Cr (soln) {tnconel) —> Crtt + T (soln) 2(905) [t is evident that such a fuel will be much more corrosive under irradiation than out-of-pile, es- peciaflly when irradiated in capsules in the MTR, unless at least part of the Li% is removed from the fuel. Because of the large neutron absorption cross sections of fuels containing natural lithium, there would be substantial decreases in the effective neutron flux in such materials in comparison with similar fuels containing pure Li7. The decreased effective neutron flux and the large dilution factors normally used in in-pile loops will make the chemical effects of tritium production less serious than they would be in capsules. Nevertheless, gy, ORNL-LR-DWG 11824 6 ] [ 1 T ~— <] 1T — T T TTTH: I ] w7 = —log (g of T per liter of solution) i [ \\1\ | "-...__._.\ fx ot P 115 T Ney;, | f | 5 \\L\ i Zx 1ot ons/cme_ _ ; ; - 10 . \‘\\L:\\\\\ mi\&e\j\\ T 2 , T g 4 ‘-\\\ ; h\:s% \\h“\ \‘\\\ ! g 5] \\\\ X f0f2 \\ ""'--.\ : \ . "‘"--..,_._._\ ] o g T~ \\\\W\ \\\\\\\ \\\ 1 E fi 3 \ M — Tt 2x 013 __\ -l | "'--\ -\\-.._ - "'-~..._._< 10 % 8 ) \"""'--. 5 \\\\""- | \ T~ T~ - = \ \\ ~Sx 1513 — ~L ~ ‘\\ ~L z < e — — N ™~ | | N~ 5 O E \ \\ M~ X014 \ \\\ — \ \\ t & 2 I~ ] 2x 1014 T~ "‘\._\ e~ T~ 107 7S ~ ng \\\ \\\ 5 \\ \h\\ \ \__\ \\\ I E % \\ \\"‘-\ x4 015 \ \\\\.\\ \ N - 5 %[ ; \l._ —— P _‘8 51073 \ \\\x \ "-~>...___= 103 B > f T \ _ \-.\ | :\ : \* N ': \\.‘\:‘"\ 5 X fo 15 . . i 1 \ | h.__‘* | | T [— "'--‘I_: 4 ¢ ] | \"‘—x \ -\-_""-—«-_.___ —-fi‘_\_‘! : 10 0 — o ] 1T e | o I | | ' 5 } L T | 1% ! 2 5 10 20 50 00 200 500 1000 IRRADIATION TIME {hr) Fig, 8.28. Production of Tritium in the Fluoride Fuel Mixture NaF-KF-LiF-UF‘ (11.2-41-45,3-2.5 mole %) Containing Natural Lithium as a Function of Time and Thermal-Neutron Flux, 203 ANP PROJECT PROGRESS REPORT in order to obtain satisfactorily high fission rates in the in-pile loops now being used, removal of some of the Li¢ will be required. In order to keep that portion of the corrosion attributable to tritium to below 100 ppm chromium in 300 hr, any lithium in fuel used in MTR capsule irradiations must contain less than 0,2% Li6. In in-pile loops, up to 1% Li% may be allowed without excessive depression of the fission power at the tip of the loop. This amount of Li% will also restrict tritium corrosion to less than 100 ppm chromium per 300 hr, There seems to be no reason to expect the apparent fission-product protection of Inconel found in NaF-ZrF4-UF fuel systems to reduce the attack of T* on :he metal surface. Furthermore, it is not evident that this protective mechanism will be operative in the lithium-containing systems. The exact distribution of tritium among the various species, T1, TF2', TF (soln), ond TF (gas), is believed to be unimportant, w.k.«"m RADIATION DAMAGE TO BORON CARBIDE 0. Sisman J. G. Morgan M. T. Morgan R. M. Carroll Solid State Division A study of radiation effects on B,C and related thermal-neutron shield materials is under way. The first irradiations were made on uncoated hot-pressed B4C (Norbide) and slip~cast B,C bonded with SiC (The Carborundum Company) at low temperatures {~200°F). The samples were sealed under helium in quartz ampoules and were irradiated in hole C-48 of the LITR for an exposure of 6 x 10'? nut (as measured by cobalt-foil acti- vation at the surface of the specimen). The samples, the quartz ampoules, and the perforated aluminum containers in which they were irradiated are shown in Fig. 8,29, The samples and their container were cooled by the reactor cooling water. Final burnup of the samples will be obtained by analysis of the B10.to-B11 ratio. After irradiation the ampoules were broken, and the gas released from the sample was measured, The gas-release apparatus is shown in Fig. 8,30, The ram is struck to breck the ampoule and release the gas into a calibrated volume. A manometer with an adjustable head is used to measure the volume of the additional gas formed as the result of irradiation, 204 HNCLASS|IFTED PHOTO 15584 HOT-PRESSED " COLD-PRESSED SINTERED Fig, 8,29. Samples of B,C and the Containers for lrradiation in the LITR. UNCLASSIFIED ORNL-LR-DWG 1306t RAM ! ‘ AMPOULE - | o /j ] | SPECIMEN — 7% | “ " TO MANCMETER Fig. 8.30. Apparatus for Releasing Gas from Quartz Ampoules in Which B4C Samples Were Irradiated, indicated that most of the helium produced from the n,a reaction with B10 in hot-pressed B C would be released at temper- atures above 1500°F, and therefore no helium evolution was expected as a result of these tests at about 200°F. The recent results confirmed the prediction; none of the approximately 1 cm3 of helium produced as the result of irradiation was detected by the apparatus, which could detect 0.2 cm3, The hot-pressed samples also retained their physical dimensions and bulk structure, as shown in Fig. 8.31. At higher burnups, however, the hot-pressed material may be expected to crumble, With the cold-pressed and sintered B ,C Previous work!3 Sw. . Yalovage, Effect of Irradiation on Hot-Pressed Boron Carbide, KAPL-1403 (Nov. 15, 1955). COLD-PRESSED SINTERED B,C IRRADIATED UNIRRADIATED PERIOD ENDING MARCH 10, 1956 (bonded with SiC), a different effect was noted. While the material appeared to retain its bulk stability, there was gas release (under irradiation) which far exceeded the total helium generation. In addition, the walls of the quartz ampoule were coated with a heavy black deposit, as shown in Fig. 8.32, and the sample weight decreased 4%. Unirradiated specimens heated for 72 hr ot 1800°F did not show these effects, It is thought that some foreign material (organic binder or pitch) introduced during fabrication sample, may have remained in the in order to determine whether the hot-pressed samples will retain helium at elevated temper- atures, gas release from samples irradiated at temperatures up to 2150°F will be measured. In - UNCL ASSIFIED PHOTO 16273 HOT-PRESSED B4C INCH Fig. 8,31, Irradiated and Unirradiated B,C Specimens, 3X. Reduced 7%. 205 ANP PROJECT PROGRESS REPORT UNCL ASSIFIED PHOTO 16275 Fig, 8.32. Coating Found Inside Quartz Ampoule in Which Cold-Pressed Sintered B4C Was lrradi- ated, 3X. Reduced 38%. 206 addition, other boron-containing materials will be investigated, such as a 2% B-98% Fe alloy, o 6.6% B,C body bonded with copper and flame- coated with copper and stainless steel, and an 89% B,C body bonded with SiC and flame-coated with A|20 . Coating adherence, dimensional sta- bility, cng helium evolution will be studied at elevated temperatures under irradiation. Future irradiations will probably be made at the MTR at temperatures up to 2000°F with radiation exposures of 1020 nut and greater. PERIOD ENDING MARCH 10, 1956 9. ANALYTICAL CHEMISTRY OF REACTOR MATERIALS J. C, White Analytical Chemistry Division DETECTION OF TRACES OF NoK IN AIR A. S. Meyer, Jr. J. P, Young Analytical Chemistry Division R. G. Affel F. W. Manning Instrumentation and Controls Division An investigation was initiated in order to de- velop an instrument for monitoring the concentra- tion of alkali metais in the exhaust gases from NaK-to-air radiators, This instrument is to be used for detecting NaK leaks in the ART radiators, as well as in the radiators now being used in test installations, Tentatively, the sensitivity of the detector to be used on the ART radiators has been specified to correspond to an incremental increase of 0.01 ppm (10 parts per billion) of alkali metals in the air while it is crossing the bank of radiators. An instrument of lower sensitivity, 10 ppm, will be satisfactory for monitoring the exit gases from the test stands, |t has also been specified that the detector must have a rapid response time and that the instrument for use on the ART must be highly dependable, since no maintenance can be performed during the period of the reactor experi- ment. Effort is being directed toward the im- mediate development of an instrument to menitor the test-stand radiators as a step in the develop- ment of the monitor for the ART, Detection of Microgram Quantities The proposed leak-detection method is based on the measurement of the hydroxyl ion which is formed when the alkali metal oxides and hydroxides are absorbed from a sample of the contaminated air in an aqueous solution. The pH of the aqueous solution will be maintained at a constant value by the coulometric generation of hydrogen ions in o quantity equivalent to the quantity of hydroxyl ion being absorbed from the sample of air, The coulometric current required to maintain the pH of the solution will be proportional to the rate of addition of alkali metals, and, if a continuous air sample is taken, it will be proportional to the concentration of alkali metals in the sample. Both photometric and electrometric methods for the detection of changes of pH are being studied., To eliminate the buffering effects of carbon di- oxide, the pH of the absorber solution will be maintained between 4.5 and 5.0, In this region the colored indicator bromcresol green (tetrabromo- m-cresolsulfon phthalein) can be used as a sensi- tive detector for increases in the alkalinity of the solution, If 1 pg of NaK is added to 1 mi of the solution, its transmittancy is decreased to 50% of its original value. A gas-scrubbing absorption cell of approximately 30-ml capacity has been fabricated (Fig. 9.1). The gases enter the mixing chamber through a quariz tube, which can be heated to temperatures as high UNCLASSIFIED ORNL-LR—-DWG 12608 LIQUID I LEVEL LIGHT *'- BAFFLE \! i CATHODE | COMPARTMENTQ\% PHOTOMETRIC ABSORPTION CELL \\ LIGHT BAFFLE MIXING CHAMBER ANOCDE 1 O 1 INCHES Fig. 9.1. Gas-Scrubbing Absorption Cell for Detecting Atkali Metals in Air. 207 ANP PROJECT PROGRESS REPORT as 1000°C, in order to prevent deposition of the metal oxides on the walls of the vessel. When bubbles of the gas pass through the helical ab- sorption tube, this section functions as a gas lift to produce circulation of the aqueous solution through the photometric absorption cell. Air flow rates of 600 cm®/min have been found to be prac- tical. In 1 min, if @ concentration of 10 ppm of NaK in the air is assumed, 6 pg of NaK will be infroduced into the absorption cell. Since it has been shown that 30 pg of NaK in the cell will effect a decrease of greater than 50% in trans- mittancy, 6 pug would cause a change of approxi- mately 15%. Since the optical device planned for use with this apparatus can easily detect changes in transmittancy of the order of 5%, a predicted response time of 1 min to a 10 ppm concentration of NaK in the dir is conservative. The time re- quired for the transfer of the solution from the mixing chamber to the photometric absorption cell is less than 3 sec, and it is thus well within the above response period. Photometric absorption cells for both inlet and exhaust air will be required in the instrument for the detection of NaK leaks from the test-stand radiators, because other operations in the vicinity of the test stands may occasionally contaminate the inlet air with high and variable concentrations of alkali metals. The instrument will be designed to respond to the incremental increase of alkali metal conceniration in the air while it is crossing the radiator, For maximum stability, o double- beam optical system will be used to compare the absorbance of the solutions in the two cells, A block diagram of the proposed instrument is shown in Fig. 9.2, Light of wavelength 630 my is separated into two beams by a mechanical chopper. After passage through the inlet and exhaust sample cells, re- spectively, the two beams of light are recombined, at phototube A. A portion of the light transmitted through the inlet-air monitoring cell is separated by a beam splitter and directed to phototube B. When alkali metals are present in the sample of inlet air, this light will be attenuated, and the decrease in the response of phototube B will be amplified and fed into a servomechanism which will pass a coulometric current through the inlet sample cell, This current will be regulated to generate a sufficient quantity of hydrogen ions fo maintain the bromcresol green indicator at the original absorbance. Since the coulometric circuits UNCLASSIFIED ORNL-LR-DWG 12609 I, —» MECHANICAL CHOPPEF?/ i EXHAUSTAIR PHOTOMETRIC 7 ABSORPTION CELL ) e Ty | -] | SERVO 2 INTERFERENCE | OPTICAL\ I, +1, (M )— | CURRENT FILTER | WEDGE — | CONTROL ll | 11 | PHOTOMETRIC | | ABSORPTION CELL | 4 — T NN . TUNED MIRRORA % AMPLIFIER INLET AIR + 'h |y~ PHOTOTUBE I‘I 5 SERVO CURRENT TUNED CONTROL AMPLIFIER Fig. 9.2, Diagram of Instrument for Detecting Microgram Quantities of Alkali Metals in Air. 208 for both the inlet-air and exhaust-air monitoring cells are in series, the same current will flow through both cells. A separate mechanical or electronic device will be incorporated at the exhaust-air cell to compensate for differences in gas flow through the two cells. Any imbalance of the light beams transmitted through the two absorption cells will generate an a-c signal at phototube A. The amplitude of this signal will be a measure of the incremental increase of the concentration of alkali metals in the air while it is crossing the radiator, The output of photo- tube A will be amplified and fed into a servo- mechanism which will pass an additional coulo- metric current into the exhaust-air monitoring cell, The amount of additional current required to hold constant the absorbance of the absorber solution in the exhaust-air cell will be monitored, and if this current exceeds a predetermined rate that is indicative of a NaK leak, an alarm will be acti- vated, The preliminary investigations of electrometric methods for determining the change of pH of the absorber solution have indicated that glass elec- trodes are somewhat more sensitive to changes in pH of the absorber solution than are the photo- metric detectors. The glass electrodes may be less dependable, however, than the photometric detectors because of variations in the asymmetry of the potential of the electrodes. Tests are now being made to determine whether the instability of the potential of the glass electrodes is sufficient to introduce false signals. Since highly stable pH meters are available commercially, the instru- mentation can be simplified if the electrometric detector can be used. The over-all design of the electrometric instrument would be functionally similar to that of the photometric instrument, Detection of Submicrogram Quantities The measurement of the absorbance of light of the frequency of a resonance line has been demon- strated to be one of the most sensitive methods for the detection of traces of vapors of certain elements. A simple optical instrument has been designed to detect concentrations of mercury vapor of less than 1 ppm by the measurement of the absorption of the mercury 2537-A resonance line. ). R. McNally, personal communication te A, S. Meyer, Jr. PERIOD ENDING MARCH 10, 1956 The absorption coefficient of the sodium doublet at 5890 A is larger than that of the mercury 2537-R resonance line, and, on the basis of the transition probabilities and half lives of excited states, cited by Mitchell and Zemansky,? the absorption coefficient,? k, at the center of each of the lines of the sodium doublet falls between that of the mercury 2537-A line and that of the cadmium 2288-A line and thus corresponds to about 10=12 N, It can be calculated from the absorption coefficient that a concentration of sodium vapor corresponding to 0.01 ppm of sodium in air would produce a 50% attenuation in a beam of sodium-doublet radiation which traversed a 50-cm optical path, Since sodium in compound form does not absorb the resonance radiation, it is necessary to heat the air containing the sodium to a temperature sufficient to cause thermal dissociation of the Na,O to sodium and oxygen ions. On the basis of thermodynamic constants derived from measure- ments of the volatility of Na,O at reduced pres- sures, Brewer? has calculated that Na,O in air would be appreciably dissociated at temperatures near 1000°C, An instrument based on these considerations has been proposed for the detection of the sodium which would be liberated by a NaK leak in the radiators of the ART, A block diagram of the instrument is presented in Fig. 9.3. The beam of sodium-doublet light from a sodium-vapor lamp or resonance bulb is divided inte two pulsed beams of equal intensity by a mechanical chopper. The beams are passed through heated absorption cells which contain samples of the inlet and exhaust air from the radiators and are recombined at the phototube. When the intensities of the transmitted beams are equal, a null a-c signal is received at the phototube. If unequal concentra- tions of sodium are present in the absorption cells, an imbalance is produced in the transmitted 2A. C. G. Mitchell and M. W. Zemansky, Resonance Radiation and Excited Atoms, Chap. I, p 92, The University Press, Cambridge, 1934. 3The absorption coefficient, &k, is defined by the relationship I = Ioe—kN, where [ and [ are the intensities of the incident and transmitted light, respectively, and N is the density of sodium in the light path in atoms per square centimeter. 4L. Brewer and J. Margrave, . Phys. Chem. 59, 421 (1955). 209 ANP PROJECT PROGRESS REPORT SODIUM VAPOR LAMP INLET AlR SYNCHRONOUS ¢ CHOPPER UNCLASSIFIED ORNL—-LR-DWG 12610 ABSCRPTION CELLS ] —2m-- OPTICAL WEDGE % MIRROR ___________ 'é—_*—-fl— PHOTOTUBE /W FURNACE ;F / —_—— —_ - — _"l _____ INTERFERENCE FILTER Y MIRROR f EXHAUST AIR METER AND ALARM SYNCHRONIZED AMPLIFIER Fig. 9.3. Diagram of Instrument for Detecting Submicrogram Quantities of Sodium in Air, beams and a proportional a-c¢ signal is generated. For low absorbancies, the output from the photo- tube, at a given temperature of the samples in the absorption cells, is proportional to the differ- ence in sodium concentration in the two absorption cells, An instrument of this type would thus give a measure of the incremental increase of the concentration of sodium in the air while it is crossing the radiators, The proposed double-beam instrument is readily adaptable to stable instrumentation and does not require careful regulation of sample flow rates. If further investigation indicates that Na,O is sufficiently dissociated at the temperature of the exhaust gases, the absorption beam can be passed directly through the stack gases to obtain a more rapid response., It will be necessary to calibrate the instrument empirically because its sensitivity will vary with the temperature of the absorbing gases and with the spectral distribution of the incident radiation, A test model of the instrument is being assembled. This apparatus contains two quartz absorption cells, 24 in. in length, which can be heated to a temperature of 1050°C, for the introduction of gaseous samples to each tube or for the volatilization of sodium compounds of known vapor pressure for calibration of the system. Tests of the effect of temperature on the sensitivity of the instrument, comparison of light Provision is being made 210 sources, and tentative calibration should be possible with this instrument. SPECTROPHOTOMETRIC DETERMINATION OF TITANIUM IN MIXTURES OF FLUORIDE SALTS WITH TIRON J. P. Young J. R. French Analytical Chemistry Division A method has been developed for the use of Tiron (disodium-1,2-dihydroxybenzene-3,5-disulfonate) in the determination of trace amounts of titanium in NaF-ZrF ,-UF, and in other mixtures of fluoride salts, The yellow color of the titanium-Tiron com- plex is a very sensitive test for titanium.> The molar absorbancy index is 13,900, The moximum absorbance of this complex is at a wavelength of 380 mu. The intensity of the color is essen- tially independent of pH over the range 4.3 to 9.6. The uranyl, zirconium, and ferric ions in the fluoride salts have been found to interfere with the development of the titanium-Tiron color. Zirconium ions interfere by forming a colorless complex with Tiron and thereby consuming the reagent. Ferric and uranyl ions form colored com- plexes with Tiron., It has been reported that the interference of the ferric ion may be eliminated by reducing the ferric ion with sodium hydrosulfite °). H. Yoe and A. R. Armstrong, Anal. Chem. 19, 100 (1947). at a pH of 4.7;% the zirconium interference may be eliminated by the addition of a large excess of reagent;> and the interference of the uranyl ion may be eliminated by the addition of iron as a carrier and the precipitation of titanium with sodium hydroxide in the presence of carbonate ion® to complex uranium as the soluble uranyl carbonate anion, In the determination of titanium in NaF-ZrF ,- UF,, quantitative separation of the titanium is achieved by an ammonia precipitation in the pres- ence of carbonate ion., Apparently, the precipita- tion of zirconium, under these conditions, co- precipitates the titanium completely, and it is unnecessary to add a cartier, such as iron. The precipitate is separated by centrifugation and then dissolved in dilute hydrochloric acid. A large excess of Tiron is added, and the solution is buffered at a pH of 4.7, The solution is then allowed to stand for 30 min, since the titanium- Tiron complex develops slowly. Solid sodium hydrosulfite is added to reduce the ferric ion, and the absorbance of the complex is measured at a wavelength of 140 mu. Since the absorbance of a solution of sodium hydrosulfite is significant below 400 myu, the absorbance of the titanium- Tiron complex cannot be determined at its point of maximum absorption (380 my). Determinations of titanium in five samples of fluoride fuels have been made by this method. The coefficient of variation for these determina- tions was 2%. Quantitative recovery of the standard solution of titanium added to several of these samples prior to analysis was obtained. DETERMINATION OF TANTALUM IN FUSED MIXTURES OF FLUORIDE SALTS J. P. Young J. R. French Analytical Chemistry Division The determination of trace amounts of tantalum in NaF-LiF-KF, as described previously,”® was 5p. A. Lee, B. S. Weaver, and J. W. Gates, Colori- metric Determination of Titanium, Part 11 (.001% to 5%), C-1.360.8 (Oct. 25, 1945). 7). c. White, Determination of Small Amounts of Tantalum in NalF-L{F-KF and in NaF-Lz'F-KF-UF4, ORNL CF-56-1-49 (Jan. 10, 1956). 8). P, Young and J. R. French, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 192, PERIOD ENDING MARCH 10, 1956 accomplished by modifying an existing method, described by Dinnin,® for the colorimetric determi- nation of tantalum with pyrogaliol. In the modified procedure the fluoride sample is carefully digested in dilute sulfuric acid in order to hydrolyze TaF, to Ta,04 without loss of tantalum by volatilization of the pentafluoride. The solution is then evapo- rated to dryness, and the residue from the evapo- ration is fused with potassium pyrosulfate. The melt is then dissolved in ammonium oxalate. The tantalum-pyrogallol complex is formed in a solution that is 0.175 M ammonium oxalate and 4 M HCI, after which its absorbance is measured at 330 mp, A satisfactory method for the determination of small amounts of tantalum in NaF-LiF-KF-UF was also developed. Since uranium interferes with the determination of tantalum by the pyrogailol method, it was necessary to develop some means interference, According to Hillebrand et «l, ' tantalum is quantitatively of removing this separated from vranyl ions by cupferron in a solu- tion of sulfuric acid {(5%) which contains tartaric acid. Since the samples were composed of alkali fluorides and uranium tetrafluoride, it was neces- sary to dissolve them carefully to prevent the loss of TaF ¢ by volatilization and to oxidize the tetra- valent uranium to uranyl ions. The samples were dissolved in a mixture of sulfuric acid (5 vol %) and aqua regia at 100°C. After the dissolution of the samples, tartaric acid was added, and a cupferron precipitation was performed at ice-bath temperature. The precipitate was removed and then ignited to obtain Ta,Og, which was fused with potassiumpyrosulfate. The tantalum-pyrogallol color was then developed, as described previously, The coefficient of variation was 2% for the re- sults of the determination of tantalum in both the uranium-containing alkali fluoride mixtures and the alkali fluoride mixtures without uranium. In the separation of tantalum from uranium, a sample which contained at least 1 mg of tantalum was taken. It is expected, however, that the separation is applicable for smaller amounts of tantalum, 9], 1. Dinnin, Anal. Chem. 25, 1803 (1953). 0. F. Hillebrand et al., Applied Inorganic Analysis, p 120, 2d ed., Wiley, New York, 1953. 211 ANP PROJECT PROGRESS REPORT DETERMINATION OF OXYGEN IN ZrF, BY BROMINATION J. P. Young M. A, Marler Analytical Chemistry Division A study of the possibility of applying the bromi- nation method of Codell 't 4o the determination of combined oxygen in metallic oxides and in fluoride salts was continued, The and Norwitz bromination method consists in the high-temperature reaction of bromine vapor, in a carrier gas of helium, with an intimate mixture of the sample and graphite. The products of this reaction are CO and the metal bromide. The excess bromine and the metal bromide are removed by a system of cold traps, while the last traces of bromine are removed by granular zinc maintained at 350°C, The CO is oxidized by hot CuQ, and the resultant CO, is utilized as a measure of the oxygen present in the sample, Attempts to determine oxygen in the presence of fluoride salts have shown that consideration must be given to the possible reaction of volatile fluorides with the glass components of the appa- ratus. In the study of the bromination of pure Zr0,, it was found'? that it was necessary to mix the oxide with a fluoride salt in order to re- move ZrO, completely as a volatile reaction product. It is believed that the fluoride salt undergoes some type of exchange reaction with ZrQ, that results in the formation of a volatile zirconium fluoride compound and an oxide of the metal in the original fluoride salt. The oxide then reacts with bromine, In spite of the com- plete removal of the ZrO,, the recovery of oxygen as CO, was low. The apparatus and the procedure used in the bromination of ZrQO, mixed with FeF, and graphite were described previously,'? In an effort to improve the recovery of oxygen as CO,, mixtures of ZrO, and FeF; were placed in small nickel bombs and heated to temperatures of 550 to 700°C for periods of 1 hr. After the bombs were cooled, the contents were mixed with graphite and allowed to react with bromine, Again, the recovery of the oxygen originally present in the ZrQ, was incomplete, The effects of mixing other fluoride salts with the ZrO, and graphite "M, Codell and G. Norwitz, Anal. Chem. 27, 1083 (1955). 12y, p. Young and M. A. Marler, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 191. 212 were also being studied. The addition of NiF, did not improve the recovery of oxygen, but it was found that the solid products resulting from the bromination of the mixture of ZrO,, NiF,, and graphite are large crystalline masses, whereas the products obtained from the bromination of the mix- ture of Zr0,, FeF,, and graphite are finely divided solids, It is anticipated that analyses of the crystalline products will provide a basis for a better understanding of the reaction mechanism. Spectrographic analysis has shown zirconium {and nickel) in the crystalline reaction products, and petrographic analysis has shown nickel bromide and an unidentifiable phase. No zirconium has been found in the graphite which remains in the sample container after bromination, DETERMINATION OF MICRO AMOUNTS OF BORON IN FUSED FLUORIDE SALT MIXTURES A. S. Meyer, Jr, W. J. Ross Analytical Chemistry Division A method has been developed for the determi- nation of small amounts of boron in fluoride salt mixtures. salts in an acidic solution of aluminum chloride at room temperature, extracting the boron into ether, and determining the boron by the carminic acid method,'® The determination of boron by direct application of the usual procedures was not feasible, because the high concentration of fluoride prevented the formation of the chromogenic com- plexes of boron. Therefore it was necessary to develop a means of separating sufficient boron from the fluoride for the carminic acid method to be used satisfactorily. is volatilized from hot acidic therefore the dissolution of the salts must be effected at a relatively low tem- Boron trifluoride solutions, and perature, unless provision is made to trap the volatile compounds of boron. It was found that the dissolution of NaF-ZrF ,-UF, could be per- formed efficiently by stirring the solid sample in an acidic {(~2 M HCI) solution of 1 M AICI3~6H20 at room temperature., In this manner, boron was retained in solution in the presence of fluoride and metal ions. 133, T. Hatcher and L. V. Wilcox, Anal. Chem. 22, 567 (1950). The method consists in dissolving the ' | he method of Glaze and Finn'4 has been adapted to the separation of boron from the other com- ponents of solutions of these mixtures. This method is based on the relatively high partition coefficient of boron between ethyl ether and an acidic solution of boron in ethanol-water (1:1). When the aqueous solution is equilibrated with an equal volume of ether at a controlled temperature, a reproducible fraction (approximately 50%) of the boron is ex- tracted into the ethereal layer. The presence of fluoride is reported to decrease the extraction coefficient. The original determination of boron in the ether phase is performed titrimetrically on a macro basis, The presence of certain cations, AI3* Fed*, Zn?*, and Ba?”, reduces the accuracy of the titration, The extraction coefficient for micro amounts of boron was established as 0.45, when the parti- tioning was performed at 19°C by shaking the mixture manually for 5 min. Practically all the aluminum and the other metallic ions remained in the aqueous phase, The reproducibility of the extraction coefficient, +0.02, and the absence of fluoride in the ether extract indicate that inter- ference by fluoride is effectively eliminated through the formation of stable complexes of aluminum and Hluoride during the dissolution of the sample. The concentrations of fluoride and metallic ions which accompany the boron into the ether phase are tolerable for the carminic acid method. This method has been used with good precision for the determination of boron in concentrations as low as 10 ppm in NaF-ZrF ,-UF,, and it should also be applicable to the determination of boron in NaF-KF-LiF base fuels and in other fluoride salts. SPECTROPHOTOMETRIC DETERMINATION OF BISMUTH IN FUSED MIXTURES OF FLUORIDE SALTS A. S. Meyer, Jr, B. L. McDowell Analytical Chemistry Division A spectrophotometric method for the determina- tion of bismuth as the tetraiodobismuthate(lll) complex'® has been applied to the determination of bismuth in NaF-ZrF ,-UF,. The absorbance of the complex was measured at 337 mp in a volume 4 W, Glaze and A. N, Finn, J. Research Natl. Bur. Standards 16, 421 (1936). 15N, M. Lisicki and D. F. Boltz, Anal. Chem. 27, 1722 (1955). PERIOD ENDING MARCH 10, 1956 of 50 ml of 1 M H,S0,. Since the cited reference includes a thorough report of variables such as iodide concentration, acid concentration, and time, the optimum conditions as set forth there were used for the determination. Test portions which contained 0.6 to é pg of bismuth in @ solution of sulfuric acid were taken for color development, The color was developed by adding 10 ml of 5 M H, 50, and 20 ml of a solution that was 14% Kl and 0.1% ascorbic acid to the test portion and then diluting to a volume of 50 ml with water, The absorbance of the solu- tion was measured against a reagent blank, The effects of several diverse ions on the ab- sorbance of the tetraiodobismuthate(lll}) complex were reported,'> but the effect of uranium was not mentioned. The effect of uranium was there- fore investigated for uranium-to-bismuth weight ratios of up to 1000:1, and it was found that for weight ratios of less than 30:1, the method was applicable with an error of less than 3% in the absorbance. For larger amounts of uranium a pre- liminary separation is necessary. The coefficient of variation was 2%, based on reproducibility of results for standard samples and duplicate deter- minations on samples which contained both uranium and bismuth, DETERMINATION OF DISSOLVED OXYGEN IN LUBRICATING FLUIDS A. S. Meyer, Jr, B. L. McDowell Analytical Chemistry Division The determination of dissolved oxygen in lubri- cating fluids was carried out in a closed system in which the dissolved or entrained gases in the fluid were swept with CO, into a 25-ml azotometer filled with a solution of KOH. The CO, was sup- plied from a 2-1 Dewar flask filled with dry ice and sealed with a Hershburg valve. The insoluble gases collectedin the azotometer were equilibrated with a solution of potassium pyrogallate, and the loss in volume of gas was attributed to absorption of oxygen by the potassium pyrogallate solution. The loss of volume at standard temperature and pressure was used to calculate the concentration of oxygen in the lubricating fluid. The oxygen content was essentially the same for each of the lubricants studied; it ranged from a maximum of 42 ug of oxygen per milliliter of fluid to a minimum of 15 pg/ml. The coefficient of variation of the results was 7%. 213 ANP PROJECT PROGRESS REPORT DETERMINATION OF ZIRCONIUM BY THE COMPLEXIMETRIC.-YERSENE METHOD J. P. Young J. R. French Analytical Chemistry Division The compleximetric-Versene (disodium ethylene- diaminetetraacetate) method for the determination of zirconium was developed by Manning, Meyer, and White.'® This method consists in adding an excess of Versene to zirconium in a sulfuric acid medium. The complex is formed within a period of 10 min, at room temperature, after the pH of the solution is raised to approximately 6, and the excess Versene is then titrated with ferric ion to a Tiron end point at a pH of 4,8, Very few cations interfere with this determination, since zirconium and iron both form very stable complexes with Versene. This method was used successfully for several months for the routine determination of zirconium before difficulties were encountered — the color change at the end point became sluggish and indistinct, and the accuracy was found to be unsatisfactory. In a further study of the method, it was found that the color change at the end point could be greatly improved by digesting the test solution on a steam bath, after the addition of the Versene, and titrating the solutions while hot. The accuracy of the procedure was improved by raising the pH of the solution to approximately 6 before the ad- dition of the Versene. A comparison of results obtained by a gravimetric (mandelic acid) procedure with the results of 21 determinations by the modi- fied compleximetric-Versene procedure gave an average difference of 0.6%. DETERMINATION OF RARE-EARTH ELEMENTS IN STAINLESS STEELS AND INCONEL A. S. Mevyer, Jr. B. L. McDowell Analytical Chemistry Division J. A. Norris Stable Isotopes Research and Production Division The method previously described'’” for the de- termination of traces of rare-earth elements in Yop, ., Manning, A. S. Meyer, Jr., and J. C. White, The Compleximetric Titration of Zirconium Based on the Use of Ferric lron as the Titrant and Disodium- 1,2-Dibydroxybenzene-3,5-Disulfonate as the Indicator, ORNL.1950 (1955). 7a. s. Meyer, Jr., and B. L, McDowell, ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 189. 214 stainless steel was modified to reduce the time necessary for the deposition of reducible metals at the mercury cathode from 4 to 1 hr by volatili- zation of the chromium before electrolysis. The chromium is removed as the volatile CrO,Cl, by the addition of solid NaCl during the dissolution of the sample and the precipitation of the oxides of silicon, niobium, tantalum, etc. The results obtained for samples of type 347 stainless steel indicate that no loss of rare earths results from this modification of the procedure, The meodified method has been used for the determination of lanthanum, cerium, samarium, europium, gado- linium, and dysprosium in types 304, 316, 321, and 347 stainless steel and in Inconel and inconel X, Various electrolysis periods, from 1 to é hr, were investigated, but no variations were noted in the recovery of the rare earths, The recovery of the rare earths for all types of samples on which the method was used was quan- titative, within the precision of the spectrochemical procedure, There are some indications that the recovery of lanthanum may vary, but the results are inconclusive at this time. The lowest recovery of lanthanum was approximately 85%. DETERMINATION OF OXYGEN IN METALLIC SODIUM A. S. Meyer, Jr. G. Goldberg Analytical Chemistry Division Further study of the distillation method for the determination of oxygen in metallic sodium'® has revealed that excellent reproducibility is obtained in the recovery of added Na,0 when distillations are carried out for periods in excess of 1 hr at temperatures between 800 and 850°F. No differ- ences in oxygen concentration were found when a series of replicate determinations of oxygen were carried out over the extreme distillation con- ditions of 1 hr at 800°F and 4 hr ot 850°F. Ad- ditional distillations at higher temperatures have confirmed the earlier observation that Na, O is volatilized at temperatures above 900°F., The exact mechanism by which Na,O is lost from the system has not yet been determined. Flame- photometric determinations of alkaline-earth metals in the distillation residues have shown that the concentrations of oxides of these elements are 187, s, Meyer, Jr.,, G. Goldberg, and W. J. Ross, ANP Quar. Prog. Rep, Dec. 10, 1955, ORNL-2012, p 186. not sufficiently large to contribute to the alkalinity of the residues. ' Efforts to measure the recovery of a weighed addition of oxide to a batch of sodium metal have been unsuccessful, When a quantity of NaOH, which was calculated to increase the concentration of oxygen by 300 ppm, was added to a previously analyzed batch of the metal, an increase of oxygen concentration of only 70 ppm was measured in a sample after the batch was agitated with helium for 12 hr ot 1000°F. Since it appears that the oxygen from solid compounds is difficult to dis- perse in metallic sodium, a test is to be made in which the contamination with oxygen will be carried out by the injection of a measured quantity of elemental oxygen into a stream of helium, which will be bubbled through the molten metal. ANP SERVICE LABORATORY W. F. Yaughan Analytical Chemistry Division Analyses of 39 samples of mixtures of fused fluoride salts were performed for the Wright Air PERIOD ENDING MARCH 10, 1956 Development Center (WADC), The determinations made on the WADC samples included total ura- nium, ftrivalent uranium, iron, nickel, chromium, and niobium, The bulk of the work during this period continued to be the analysis of fused fluoride salt mixtures and alkali metals for the Reactor Chemistry and Experimental Engineering Groups. A total of 1638 samples was analyzed for all ANP sources, and an average of 3.4 results was reported per sample, The backlog consists of 77 samples. A breakdown of the work follows: Number of Number of Samples Reported Results Reactor Chehistry 937 3139 Experimental 420 1904 Engineering WADC 39 202 Miscellaneous 242 282 Total 1638 5527 215 ANP PROJECT PROGRESS REPORT 10. RECOVERY AND REPROCESSING OF REACTOR FUEL H. K. Jackson D. E. Ferguson H. E. Goeller W, K. Eister M. R. Bennett R. L. Jolley F. N. Browder W. E. Lewis G. |, Cathers J. T. Long W. H. Carr R. P, Milford S. H. Stainker Chemical Technology Division PILOT.PLANT DESIGN AND CONSTRUCTION Completion of the engineering design of the pilot plant for recovering fused-salt fuels has been de- layed until March 1956, The remaining major items of equipment, the electrical load center and the instrument panel boards, are to be delivered about mid-April, As of March 10, installation of equip- ment is expected to be 60% complete, with 100% completion scheduled for May 15, 1956. ENGINEERING DEVELOPMENTS Contactor A percolator type of contactor was chosen for the fluorinator to be used in the volatility pilot plant, |In performance tests of the contactor, the fused-salt fuel mixture was entrained in the product gas line at a rate of 4.2 g/hr at 625°C and a sparge gas flow rate of 3.3 scfm (Table 10.1), which is the probable sparge-gas flow rate for pilot-plant operation. Analysis of the entrained TABLE 10.1. ENTRAINMENT OF FUSED SALT MIXTURE IN PRODUCT GAS LINE FROM FLUORINATOR VESSEL Fused salt composition: NaF-Zer (5050 mole %) Duration of run: 2 hr Sparge gas: nitrogen Vessel temperature: 625°C Run Sparge Gas Flow Rate Entrainment Rate No. (scfm) {g/hr) 17 0.0 0 18 1.1 0.75 19 2.2 , 3.13 20 3.3 4.20 216 material indicated that 70 wt % was particles of the fused salt mixture and that 30 wt % was ZrF sublimate. 4 A vapor trap has been designed for eliminating entrainment in the product gas line, but it has not yet been tested. It was also found in these tests that the flange on the fluorinator, which will be at 105°C when the fluorinator is at 650°C, is leaktight against a gas pressure of 12 psig. Freeze Valves A freeze valve capable of sealing a ]/’Q-in.-dia molten-salt transfer line against a gas pressure of 20 psig was designed. A reservoir of molten salt is used in one arm of a U tube to seal against a frozen plug in the other arm. Both the inlet and outlet arms are vented. In 50 melting and freezing cycles no leakage was detected. Several designs of internally cooled freeze valves were tested, however, on the assumption that rapid cooling was not effective, and it is now reasonably certain that failure of earlier designs was the result of minute fissures caused by shrinking of the salt upon cooling from its freezing point to the test temperature, PROCESS DEVELOPMENT Fluorination Studies In further studies on the fluorination step, it was found that the fluorine-to-uranium mole ratio re- quired for volatilization of more than 99% of the UF, could be decreased by the elimination of impurities but that it was not significantly affected by the concentration of the uranium in the fused- salt fuel mixture (Fig. 10.1). The method of intro- duction of fluorine into the melt and the rate of flow of the sparge gas had some effect on the fluorine-to-uranium mole ratio {(Fig. 10.2), but the results could not be correlated with the known PERIOD ENDING MARCH 10, 1956 Sl 108 OCRNL-LR-DWG 12641 ) : /7 / / / 7/ / / / / 80 : / _ / 3 D 60 N 2 e 5 , 2 40 s — w L o 20 0 1 2 3 4 Fo /U MOLE RATIO RUN U CONTENT NO. MOLTEN SALT (mole %) 1 AS-RECEIVED NoF-ZrF,-UF, (50-46-4 mole %) 4 6 NoF - ZrF,-UF, (50-46-4 mole %), HYDROFLUORINATED 4nr AT 600°C 4 7 PREFLUORINATED NcF-ZrF, (50/46 mole RATIO) PLUS UF, (76.2 %" URANIUM) a 8 PREFLUORINATED NaF-ZrF, (50/46 mole RATIO) PLUS UF, (76.2 %% URANIUM), AIR-SPARGED FOR 2hr BEFORE FLUORINATION q 2 PREFLUORINATED NaF-2ZrF, (50/46 mole RATIO) PLUS UF, (72.9%" URANIUM) 4 3 PREFLUORINATED NaF - ZrF, (50/46 mole RATIO) PLUS UFy (72.9 % URANIUM) a 4 PREFLUORINATED NaF-ZrF, {50/46 mole RATIO} PLUS UF, (72.9%" URANIUM) 5 5 PREFLUORINATED NaF-ZrF, (50/46 mole RATIO) PLUS UF, (72.9 %" URANIUM) 1 ¥ THEORETICAL = 75.8% Fig. 10,1, Effect of Uranium Concentration and Impurities of the Fused Salt Fuel Mixtures on the Fluorine-to-Uranium Mole Ratio Required for UF Volatilization. Conditions: 100 ml of F, per minute; lflé-in.-diu sieve plate on dip tube that introduced fluorine to melt, 217 ANP PROJECT PROGRESS REPORT S ORNL-LR-DWG 12612 100 7 77T 77T ;7] / S0 80 ya A UF6 VOLATILIZED (%) Mo D M O O Q AN \ Z‘.& ~ O AN @ N\ N | | | | AN \ \ \ 0 1 | { 2 3 4 F, /U MOLE RATIO RUN FLOW RATE (ml/min) GAS DISPERSION DEVICE ON END OF NO. Fo No 1/,-in.-DIA DIP TUBE 10 100 0 NONE 5 40 200 SIEVE PLATE, 3/, -in.- DIA HOLES 1 100 0 SIEVE PLATE, 34 -in.-DIA HOLES 2 100 200 SIEVE PLATE, %4 -in.-DIA HOLES 3 150 0 SIEVE PLATE, 3/4-in.-DIA HOLES 4 150 150 SIEVE PLATE, 344 -in.~-DIA HOLES 6 300 0 SIEVE PLATE, 3/g4-in.-DIA HOLES 7 100 0 SIEVE PLATE, '/yg ~in.~DIA HOLES 8 100 200 SIEVE PLATE, !/y¢ -in.-DIA HOLES 9 150 0 SIEVE PLATE, Y4 —in.-DIA HOLES {1 100 0 THREE SIEVE PLATES, 3/g4-in.-DIA HOLES, '/, in. APART {2 100 0 PERCOLATOR DRAFT TUBE Fig, 10.2. Effect of Sparge Gas Flow Rate and Method of Introduction of Flyorine into the Melt on UF6 Volatilization. 218 variables. The experiments were performed at 600°C in a 2-in.-dia nickel reactor with a 375-g charge of NaF-ZrF ,-UF,. In some of the tests the salt was made by the addition of UF, to NaF- ZrF, (50/46 mole ratio) that was believed to be relatively free of oxide impurities as a result of previous use in a fluorination run, Uranium tetra- fluoride concentrations of 1, 2, and 4 mole % were used to study the effect of concentration. In other tests NaF-ZrF ,-UF, (50-46-4 mole %) was used as received. Data were obtained by direct sampling of the salt at intervals during the fluorination. The curves were extrapolated to the 100% volatili- zation point for comparison, but usually a sharp break was observed in the curve between 95 and 100%, which extended the curve to higher fluorine- to-uranium mole ratios for volatilization of the last traces of UF . Volatilization of more than 99% of the UF , from as-received NaF-ZrF ,-UF, required a fluorine-to- uranium mole ratio of about 3.3:1, which was re- duced to about 2.2:1 by sparging with HF for 4 hr before fluorination. In two tests with the fuel mix- ture synthesized by adding UF, with a uranium content of only 72.9% (theoretical, 75.8%) to pre- fluorinated NaF-Zer, the fluerine-to-uranium mole ratio required for more than 99% UF , volatilization was about 2.4:1. When very pure UF,, uranium assay of 76.2%, was used, the fluorine-to-uranium mole ratio was 1.4:1, which represents a fluorine utilization efficiency of about 70%. Vapor Pressure of the UF ;-NaF Complex Vapor-pressure data for the UF -NaF complex were obtained by the transpiration method over the temperature range 80 to 320°C (see Fig. 10.3). The reaction involved is described by the equation (1) UF,-3NaF(s) —> UF(g) + 3NaF(s) The data are fitted with the analytical expression log Py, = 1088 — (5.09 x 10%/T) , where T is the absolute temperature. Use of the Clausius-Clapeyron formula with this equation resulted in a value of +23.2 keal per mole of UF, for the enthalpy change of Eq. 1. The data were obtained by passing nitrogen at a flow rate of 100 ml/min, or less, through a pre- pared bed of the UF ,-NaF complex at any desired temperature, trapping out the UF, in the nitrogen PERIOD ENDING MARCH 10, 1956 oy 3 QORNL-LR-DWG 12613 Np FLOW RATE: © 100 ml/min o ® 50 mi/min 4 20 ml/min " log Fym =10.88 - 508 x 10°/7 T c o € £ E i w @ ) W Wy ul [ag o, e 10 Q a g 102 1o 1074 1.6 1.9 22 2.5 28 103 x 17 (°K) Fig. 10.3. Dependence of UF .NaF Complex Vapor Pressure on Temperature. stream in dilute AI(N03)3 solution, and measuring the total volume of nitrogen with a wet-test meter, The UF, hydrolysis samples were analyzed by colorimetric or fluorimetric methods to an accuracy of better than 15%. Temperature control of the bed was maintained always to within £0,2°C. The UF .-NaF complex was prepared by saturating a 30-g bed of 12-20 mesh Harshaw NaF in a 1-in.-dia vertical nickel reactor with UF , at 100°C. Crude adiabatic experiments were made with 100-m} batches of NaF to show that the sorption heat of 23.2 kcal per mole of UF, produces a large temperature rise, approximately 130°C, if total saturation with UF, (preheated to about 100°C) is carried out in a period of a few minutes, Uranium Losses on Desorption of UF ; from NaF Further evidence was obtained that proper control of temperature and flow would prevent excessive uranium losses in the NaF absorption-desorption 219 ANP PROJECT PROGRESS REPORT pro®®@&te developed for decontaminating UF from fission-product activity. Successive use was made of a two-bed NaF system for three complete process cycles with UF .. The over-all uranium loss was 0.04% in the fjrst bed and less than 0.01% in the 220 second bed. In a single cycle test in the same equipment with fresh NaF, the uranium losses were 0.05 and 0.01% in the respective beds, The percentage loss therefore does not appear to de- pend on how many process cycles are used. Part li SHIELDING RESEARCH PERIOD ENDING MARCH 10, 1956 11. SHIELDING ANALYSIS F. H. Murray C. D. Zerby Applied Nuclear Physics Division S, Auslender H. S. Moran Pratt & Whitney Aircraft ENERGY ABSORPTION RESULTING FROM GAMMA RADIATION INCIDENT ON A MULTIREGION SHIELD WITH SLAB GEOMETRY S. Auslender The code of a Monte Carlo calculation of energy penetration and deposition resulting from transport gamma radiation in a shield of slab geometry! has been used in a parametric study of a two-region lead-water shield. The code utilizes straight- forward sampling techniques, except for a doubling technique operating on the unscattered flux. For the parametric study the radiation was 1-Mev gamma rays incident on the slab at 0 deg, 60 deg, 70 deg 32 min, and 75 deg 31 min. The first region of the slab was composed of water 1,5 mean free paths (mfp) thick at the initial gamma-ray energy, Ic. b. Zerby and S, Auslender, ANP Quar. Prog. Rep. Dec, 16, 1955, ORNL-2012, p 203, UNCLASSIFIED 2-01-059-47 {28 I ] | FRACTION OF THE TGTAL ENERGY = > © . REFLECTED —0.0345 o 112 ——— -———1— ABSORBED —0.7240 5 TRANSMITTED —C.2415 > £ oeel — — | - _ - T e L WATER — = LEAD:‘ < o = a = 3 i aso — —_— x L \ = 83 \ Syess | — —1- — = o _l._i z 4 EQo4s —0u — —t = m o - — \ — e & § -‘-_"'_‘_'T___ — ; \\ = . o 0.32 . == o e S D P = | o LL_ Q46— “_ - - > | O L | 0 0.5 10 1.5 20 Ho s NORMAL THICKNESS OF COMPOSITE SLAB IN MEAN FREE PATHS AT INITIAL ENERGY Fig. 11.1. Fraction of Energy from 0-deg Inci- dent Gamma Rays Absorbed in a Water-Lead Slab. and the second region was composed of lead 0.5 mfp thick. Preliminary resuvits of the calculation are shown in Figs. 11.1 through 11.4. The dashed lines are a fit by eye to the data. The large rate of change of heat deposition near the front surface of the lead can be explained by the rapid absorption in the lead of the low-energy degraded gamma rays. The coding has been extended to calculate energy flux and tissue dose rate, as well as energy depo- sition. Plots of the buildup factor for the dose rate, from the gamma rays that penetrate the slab, as a function of the composition of the slab are presented in Figs. 11.5 through 11.9. The buildup factor, B,, is defined as: D B r = 5 ' = Hpr sec D_e ¢ UNCLASSIFIED 2—0i—-059-48 T 144 ‘ & \ im b ‘ . i wl i E oo | _ FRACTION OF THE TOTAL ENERGY = | ; ' REFLECTED - Q.0910 g ABSORBED — 0.8578 24| | TRANSMITTED-0OSt2 | Lel < ot WATER - *#—"L'F LEAD -] @ N £ 0.96 _ S RN a SO I L | i o 0.80 N I ‘ L) . g N\ 5 0.64 N\ T '_‘—'i — 1 ! 8 1 g 048 _ ~ L - N i 3 N \ g o032 3 = - 5 . \ 2 |~ N 5 S 2 _ N\ —— 016 = —T o \~___‘ & i 5 o | ‘ x 0 05 1.0 1.5 2.0 ~ po7» NORMAL THICKNESS OF COMPOSITE SLAB IN MEAN FREE PATHS AT INITIAL ENERGY. Fig. 11.2. Fraction of Energy from 60-deg Inci- dent Gamma Rays Absorbed in a Water-Lead Slab. 223 ANP PROJECT PROGRESS REPORT UNCLASSIFIED 2-01-059-49 = g 1.60 | L W ; w4449 WATER tea— | EAD ——mf a \ ! Ll \ : = \ | L r 1.28 - ul X - W \ \ O o A 2 o2 { | , ; - S FRACTION OF THE TOTAL ENERGY = @ ! REFLECTED-0.1338 " > 098 \ ABSORBED ~0.84143 o \ TRANSMITTED —0.0249 ¥ w —\ > 080 |- - \\ = ) =\ \ - 0.96 - ~»~ \ xr 016 —_— e T Y ———— - ‘ N\ - L. \ \ ._; —.h--.___W ;\\ 1 - 0 0 05 1.0 15 20 Hery NORMAL THICKNESS OF COMPOSITE SLAB IN MEAN FREE PATHS AT INITIAL ENERGY Fig. 11.4. Fraction of Energy from 75-deg 31- min Incident Gamma Rays Absorbed in a Water- Lead Slab. The plots of Figs. 11.5, 117, and 11.9 are for shields with the lead in front of the water, and the plots of Figs. 11.6 and 11.8 are for water in front of lead. The sources for Figs. 11.5 through 11.8 are plane monodirectional, The source for Fig, 11.9 is a surface source having a cos™ 6 angular distribution, where 6 is measured from the normal to the surface. This source is normalized to one gamma ray per unit area of surface per 27 steradians (in the forward direction): 1 = f S(Q) 4Q 2T n + 1 27 cos” @ S(Q) = When n = 0, the source is isotropic. For n = o, the source is plane monodirectional and normal to the surface. It is obvious from the plots that lead is more effective in stratified slab shields when it is placed behind a good scatterer, such as water, This is especially true for the lower source ener- gies. The explanation for this is fundamentally the same as for the large energy deposition rate UNCLASSIFIED 2-01-059-%4 3.5 ‘ | | A 8,, DOSE BUILDUP FACTOR 1.00 0.75 0.50 0.25 0 Fig. 11.5. Buildup Factor for Dose Rate in a Lead-Water Shield Resulting from 3-Mev Incident Photons from a Plane, Monodirectional Source. PERIOD ENDING MARCH 10, 1956 in the front edge of the lead when it is behind water. The scatter in the answers is, of course, due to the method of solution. Since the answers are estimates of the correct answer, any one problem may be calculated several times, each If this is done, the average of the several estimates can be accepted as the answer, and a standard deviation can be calculated for it. time with a new set of random numbers. UNCLASSIFIED 2-01-059-52 3.5 Bor =1mfp 2.5 2.0 o 8, , DOSE BUILDUP FACTOR 1.0 -9 2 /Pb Py r o0 7‘4 o 0.5 0 0.25 0.50 0.75 1,00 P = (pgrpy /iror AT INITIAL ENERGY Fig. 11.6. Buildup Factor for Dose Rate in a Watet-Lead Shield Resulting from 3-Mev Incident Photons from a Plane, Monodirectional Source. 225 ANP PROJECT PROGRESS REFORT UNCLASSIFIED UNCLASSIFIED 2-01-059-53 2-04-059-54 4.0 T 4.0 f b 3 | | Kor =1imfp Hor =imfp | 3.5 35 . . | | | | 3.0 | - . e » L - o a = o 2 o o < 2.5 a. o po ) 2 S b3 5 = @ @ § & & 8 - - o @ ’ % HZO—F- - H Ob%/ljb// 1.0 i 10— 4 2 /‘ t Y P#o’fi; Pral / s I //#io 'ljof | 0.5 0.5 | l 1.00 0.75 0.50 0.25 C 0 0.25 0.50 0.75 1.00 P=Apgrlp, /tigr AT INITIAL ENERGY P=(pgrlpy/Hor AT INITIAL ENERGY Fig. 11.7. Buildup Factor for Dose Rate in o Fig. 11.8. Buildup Factor for Dose Rate in a Lead-Water Shield Resulting from 1-Mev Incident Water-Lead Shield Resulting from 1-Mev Incident Photons from a Plaone, Monodirectional Source. Photons from a Plane, Monodirectional Source. UNCLASSIFIED ~ 2-01-059-55 e 7= 0 - x g A=t 5 n=2 E ‘fl =4 — o n =8 8 '1 n== 0 ! > m - Ll w o 4 O -~ n+1 a T, 3 2, C088 ANGULAR SURFAGE |7 (2 _ . SOURCE DISTRIBUTION-" |77 , , 1.1 1.00 075 0.50 P={por)py/ror AT INITIAL ENERGY Fig. 11.9. Buildup Factor for Dose Rate in a Lead-Water Shield Resulting from 3-Mev Incident Photons from a Plane cos” 9 Surface Source. 226 S. K. Penny PERIOD ENDING MARCH 10, 1956 12. REACTOR SHIELD DESIGN F. L. Keller D. K. Trubey L. B. Holland Applied Nuclear Physics Division C. A. Goetz H. C. Woodsum Pratt & Whitney Aircraft R. M. Davis, Glenn L. Martin Co. GAMMA-RAY HEATING IN A 300-Mw CIRCULATING-FUEL REACTOR R. M. Davis C. A. Goetz A study of the gamma-ray heating in the lead and alkylbenzene shield of a 300-Mw circulating- fuel reflector-moderated reactor (Table 12.1} was completed. Throughout the calculation, the latest experimental data from the LTSF mockup of the circulating-fuel reflector-moderated reactor and shield (RMR-shield) were incorporated insofar as possible. The heating in the shield was resolved TABLE 12,1. PARAMETERS OF A 300-Mw CIRCULATING-FUEL REACTOR USED IN GAMMA-RAY HEATING CALCULATION Reactor or Shield Region Thickness {in.) Radius (in.) ane Beryllivm island e 6.700 Sodium passage 0.187 6.887 Inconel-X cladding 0.125 7.012 Core fuel region 5.988 13.000 Inconel-X cladding 0.156 13.156 Sodium passage 0.187 13.343 Beryllium reflector 11.887 25,230 Inconel-X cladding 0.010 25.240 Sodium passage 0.066 25.306 Inconel-X cladding 0.010 25,316 Boron-10 0.200 25.516 Inconel-X cladding 0.010 25.526 Sodium passage 0.066 25,592 Inconel-X cladding 0.250 25.842 Heat exchanger 7.030 32.872 Inconel-X cladding 0.125 32,997 Thermal shield 1.035 34.032 Pressure shell 1.000 35.032 Insulation 1.000 36.032 Insulation Inconel-X shell 0.032 36.064 (91.60 cm) Alkylbenzene passage 0.375 36.439 (92.56 em) Lead 1.000 37.439 (95.10 cm) Alkyibenzene passage 0.375 37.814 (96.05 cm) Lead 1,76 (rear) 39.574 (100.52 cm) 4.56 (front) 42.374 (107.63 cm) Alkylbenzene 7.25 {rear) 46.824 (118.93 cm) 14.10 (front) 56.474 (143.44 cm) 227 ANP PROJECT PROGRESS REPORT (1) heating by primary gamma rays otiginating in or near the reactor core, (2) heating by fission-product-decay gamma rays from the heat exchanger region, and (3) heating from thermal-neutron captures in the lead and borated (2% boron) alkylbenzene. The third component was subdivided to take into ac- count the secondary gamma rays from lead, hydro- gen, and boron captures. Alpha particles from thermal-neutron captures by boron were also con- sidered and were the only source of heating other than gamma rays included in the calculation. The results of the calculation are plotted in Figs. 12.1 through 12.6. The methods of calculation are described below. into three principal components: Primary Gamma-Ray Heating The primary gamma-ray source includes both prompt and fission-product gamma rays from the 2-01-059-556 TOTAL HEATING IN SHIELD (watts/g) N 95 99 103 107 11 15 419 REACTOR RADIUS {cm) Fig. 12.1. Total Heating in the Rear Portion of the Lead and Alkylbenzene Shield of o 300-Mw Circulating-Fuel Reactor, 228 core, capture gamma rays from the Inconel core shell cladding, and capture gamma rays from the beryllium. The heating in the alkylbenzene and lead shield that results from these gamma rays was determined by the procedure outlined below. The heating in the alkylbenzene was obtained by use of a recent analysis' of LTSF measure- ments in which the important source contributions to the gammo-ray dose rate in water beyond an RMR-shield mockup were determined for various lead and water thicknesses. The usual material and geometry transformations were applied to the resulting curves to obtain the dose rate in alkyl- benzene behind various lead thicknesses. The dose rate was then converted to heat, with the TR. W, Peelle e al., ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 210. 2-04-059-57 i I — — 1 PRIMARY GAMMA-RAY HEATING. 5| ~——=— 2 HEAT EXCHANGER FISSION-PRODUCT U GAMMA -RAY HEATING. “‘“'fl’—_ 3 HEATING BY THERMAL-NEUTRON | CAPTURE IN SHIELD., ———A— | - HEATING IN SHIELD (watts /g) I 9 95 99 10 107 444 s { REACTOR RADIUS (cm) - 9 Fig, 12,2, Components of Total Heating in the Rear Portion of the Lead and Alkylbenzene Shield of a 300-Mw Circulating-Fuel Reactor. following expression: where Hokyt, Plaliyrrtpy) = Ky — Dp ; pla,zitp,) Hia,z) — alkyl, P tqlkyl PERIOD ENDING MARCH 10, 1956 Ay 2-01-059-58 5 CAPTURES IN LEAD.” HYDROGEN CAPTURE GAMMA-RAYS FROM ALKYLBENZENE. BORON CAPTURE GAMMAS FRCM ALKYLBENZENE. 107! ALPHA PARTICLES FROM BORON CAPTURE. — 3 n )] M 2 N 3a+3b + 3¢ HEATING IN SHIELD (watts /g) n 3 o [ 40—5 3a Toptd 9 95 99 103 107 M 15 149 REACTOR RACIUS (cm) Fig. 12.3. Components of Heating by Thermal. Neutron Captures in the Rear Portion of the Lead and Alkylbenzene Shield of a 300-Mw Circulating- Fuel Reactor, r o c r . 19263 heating by primary gamma-ray sources {w/g), thickness of alkylbenzene (cm), thickness of lead (cm), tactor to convert from dose rate to heat 2.58 x 10~7 (w/g)/(r/hy), outside radius of reactor core (33.0 cm), radius to a point in the reactor at which heating is computed (em), 229 ANP PROJECT PROGRESS REPORT 2-01-059-59 n TOTAL HEATING IN SHIELD (watts/g} w ~ 1o~ 87 gt 95 99 103 107 " "5 12 123 127 134 135 139 143 REACTOR RADIUS (cm) Fig. 12.4. Total Heating inthe Forward Portion of the Lead and Alkylbenzene Shield of a 300-Mw Circu- lating-Fuel Reactor, D = primary dose rate (r/hr) in the LTSF at a distance z from the LTSF source plate p,LT(@zilpp) = with a radius @ (@ = 35.5 cm), H(a,z) = transformation from the dose rate of a finite disk (radius a) source to that from an - infinite plane source E (uz) . Eip2) - E (2)(1 + (a/2)7]'/2 O’R/O'LT = ratio of reactor to LTSF source plate equivalent surface source strength for neutrons, o °r R = ‘ 4171'3_ P, = reactor power (3 x 108 w), ’ o PLT T ma? ' - 230 PERIOD ENDING MARCH 10, 1956 aid - 2-01-059-60C PRIMARY GAMMA-RAY HEATING HEAT EXCHANGER FISSION-PRODUCT GAMMA-RAY HEATING HEATING BY THERMAL - NEUTRON CAPTURE IN SHIELD HEATING IN SHIELD (watts/g) 87 ot 95 9% 103 107 1 115 "o 123 127 134 135 139 143 REACTOR RADIUS (cm) Fig, 12,5. Components of Total Heating in the Forward Portion of the Lead and Alkylbenzene Shield of a 300-Mw Circulating-Fuel Reactor, P,r = LTSF source plate power (3.5 w), 2 c, = change in attenuation that results from replacement of the 4-in.-thick test heat exchanger mockup with a particular reactor heat exchanger composition and thick- ness (based on exponential attenuation at 6.8 Mev), ¢, = change in attenuation that results from the addition of small claddings, correct pressure shell thickness and composition, etc., c, x c, = 0.460, c, = change in attenuation that results from the substitution of alkylbenzene-350 at 330°F, assuming that p ), ). = 0.8 HH 0 based on electron density. 2The latest calibration of the new LTSF source plate has yielded a value of 3.5 w for the old source plate (see Sec. 13). 231 ANP PROJECT PROGRESS REPORT 2-01-059-#61 30 CAPTURES IN LEAD. 3b HYDROGEN CAPTURE GAMMA-RAYS FROM ALKYLBENZENE 3c BORON CAPTURE GAMMAS FROM ALKYLBENZENE. 3d ALPHA PARTICLES FROM BORON CAPTURE. 30+3b+ 3¢ 3c HEATING IN SHIELD {(watts /g) 9% 105 115 125 135 185 REACTOR RADIUS {cm) Fig. 12,6, Components of Heating by Thermal-Neutron Captures in the Forward Portion of the Lead and Alkylbenzene Shield of a 300-Mw Circulating-Fuel Reactor, The determination of the primary gamma-ray heating in the lead was made in a slightly different manner. Since there were no LTSF curves of primary gamma-ray dose rate as a function of lead thick- ness, it was necessary to use the transformed LTSF curves mentioned above, that is, plots of the primary gamma-ray heating in the reactor as a function of alkylbenzene thickness for various lead thick- nesses. 1he lead thicknesses were 0, 1.5, 3, 4.5, and 6 in. By integrating over r from the lead surface to = infinity, the total heating in the alkylbenzene behind 1.5 in. of lead was evaluated and subtracted from that behind no lead to obtain the total amount of energy deposited in the first 1.5-in.-thick lead layer in the reactor, as follows: Hpp pl0 < fpy < 1.5 in.) > M L2 > _ : 2 470 4111 ";o Hateyt, platiytifpy = 0) r%dr — dap -£0+1.5 - Haleyt, PUalkyitpp = 1.5 in) r%dr dmepy 3 _ 3 3 [(ro + 105 ino) - ro] 232 PERIOD ENDING MARCH 10, 1956 where Hpp p(0 £ tp, £ 1.51in.) = average heating by primary gamma-ray sources in the first 1.5-in.-thick lead layers (w/q), ro = inside radius of lead (cm), Pp, = lead density (9/cm®), Palkyl = alkylbenzene density (g/cm?), qukyl,P(tulkyl’th) = heating in the alkylbenzene at a distance tolkyl behind a thickness tpy of lead (w/g). Repetitions of this procedure for the other lead layers resulted in a histogram of primary gamma.ray heating in the lead. A smooth exponential curve was then fitted to the histogram. Fission-Product Gamma-Ray Heating In the absence of experimental data for the contribution of fission-product gamma rays from the heat exchanger region, it was necessary to calculate the heating from the fission-product gamma rays. The results of recent spectrometer measurements of fission-product photons in the rotating-belt experiment at the LTSF were used in the calculation.! The general procedure foliowed was to evaluate the heating contribution from each of several ener- gies and then to perform a numerical integration over the energy range. The energies chosen were 0,10, 0.25, 0.50, 1.0, 2.0, 3.0, 4.0, and 6.0 Mev. The surface source strength (at the outer surface of the heat exchanger region) of photons of energy E was determined first. The energy deposited at o distance from an infinite plane with this source strength was then computed, after which a geometrical transformation was applied to convert to spherical geometry. The choice of a buildup factor in determining the dose rate from an infinite plane was critical. The preponderance of Inconel between the heat exchanger and the lead suggested the use of the energy absorption buildup factor for nickel when the energy deposition at the inner surface of the lead was computed. (In the actual calculation, the buildup factor for iron was employed, since that for nickel was not available.}) When the heating at a point a few mean free paths inside the lead was computed, however, the buildup factor for lead was used. The buildup factor for lead was also applied for the heating in the alkylbenzene at the last lead-alkylbenzene interface. Water buildup factors were employed for results at greater distances into the alkylbenzene. The following expression was used in the calculation of the heating from the fission-product gamma rays in the heat exchanger: o) = fEHHX(r,E) dE where b (E) . SE) O HHX(T:E) = o r '—2_ Ba [iznui(E)ti} E] (?Fi(E)ti} ' ., (E) = mass energy absorption coefficient for photons of energy E, p Tux = outside radius of reactor heat exchanger (32,872 in.), 0 PO S(E) = K(E) L(E) , w(E) K(E) = EN,(E) 3.1 x 10'° (fissions/sec)/w, 233 ANP PROJECT PROGRESS REPORT pr(E) = number of fission-product photons emitted per fission per unit energy interval centered about the energy E, p R = v o ;I Vi P, = reactor power (3 x 108 w), V, = total fuel volume (20.11 %), / v, = volume fraction of fuel in heat exchanger (0.274), ) t(E} = total gamma-ray absorption coefficient for photons of energy E, -~ E)¢t r HX HX. ~ulE)t e - LE) = 1 = %1 ;~HBhx , T x pS(EY s o O Tyx. = inside radius of reactor heat exchanger (25.842 in.), I HX heat exchanger thickness (7.03 in.), enérgy absorption buildup factor for energy E and 2 (E)t . mean free paths, 7 f dx x 2 E)t, z o 8 - M &= ™ k= ~, I t — M & ™ = — Ik Heating by Thermal-Neutron Captures in the Shield Heating in Lead by Capture Gamma Rays Produced in Lead. — The heating in lead by secondary gamma rays from thermal-neutron capture in the lead was calculated by using curves of the thermal- neuvtron flux in lead which were derived from LTSF curves of thermal flux in water behind various thick- nesses of lead. The general method of calculation was to construct a series of spherical shells that were concentric about the point at which the heating was being determined and to sum up the contribution from each, calculated by means of the following expression: #a x2 62 , ea—,u,x ) Hpp,s () = KpF 2 |— f f B(r7) Blpx) ——— 2m® sin 0 d6 dx \p Pb x.l 8] A where HPb.Sl(r) = heating in the lead by secondary gamma rays produced in the lead (w/g), K, = 1.6 x 10- 13 w/Mev/ sec, E = energy of lead capture gamma ray (7.4 Mev), % = lead capture cross section, u,/p = mass energy absorption coefficient, ¢(r’) = thermal-neutron flux at point r* where capture occurs, B{ux) = point isotropic energy absorption buildup factor for lead, x = distance from point 7 “ to point r where the capture gamma ray is absorbed. Heating in Lead by Capture Gomma Rays Produced in Alkylbenzene. — A rough determination was made for the heating in the lead from absorption of hydrogen and boron capture gamma rays produced in the borated alkylbenzene. Heating contributions from several angles 6 about the point of absorption 234 PERIOD ENDING MARCH 10, 1956 were computed so that a curve of heating per unit angle vs angle could be drawn. An integration over 6 then yielded the desired result. The equation employed was K, /2 X1 o= HX HPb,Sz(r) = KyE Ec 0 f f H(r ") B(ux) ——2217962 sin 8 d0 dx , pp O X4 Arx where HPb,SZ(r) = heating in lead by secondary gamma rays produced in alkylbenzene (w/g), E = energy of boron or hydrogen capture gamma ray (0.5 or 2.2 Mev, respectively), %_ = capture cross section for hydrogen or boron in alkylbenzene. Heating in Alkylbenzene by Capture Gamma Rays Produced in Lead and Alkylbenzene. — The analy- ' of the LTSF measurements which gave the secondary gamma-ray dose rate in water beyond an sis RMR-shield mockup was employed in calculating the secondary gamma-ray heating in the alkylbenzene. These secondary gamma rays consisted of capture gamma rays from lead and from the boron and hydrogen in the alkylbenzene. The heating was determined by means of the following expression: r c —-tl/)ll Halkyl.53 = K1f5f6f7cm_r— ¢ Hla,zpp) , where Hc!kyl ¢ = heating in alkylbenzene by secondary gamma rays produced in lead and alkylbenzene - P W), fs = dose rate (r/hr) in alkylbenzene from an infinite plane source of secondary gamma radiation having the same surface source strength as the LTSF RMR-shield configuration containing 12 in. of beryllium and 4.5 in. of lead = DS,LTHTZ in. Be, 4.5 in. Pb, ta”(yl) H(a”tqikyl)' Ds 1 7B epbrtalkyl) fa - : : , as tabulated in Table 12,2, taken from work by D¢ 7(12in. Be, 4.5 in. Pb, £, ) J. B. Dee and co-workers, f; = correction for the differences in composition and thickness between reactor and LLTSF heat exchangers (2 -2 Hx,R=%Hx, LT 10:16 e-Z . ux, (tHx=10-16) C = correction factor to account for small material differences between the LTSF and reactor, i thickness of alkylbenzene cooling layers in the basic lead shield {(em), TABLE 12,2, CORRECTION FACTOR /g APPLIED TO SECONDARY GAMMA-RAY DOSE RATE FOR YARIATIONS IN LEAD AND BERYLLIUM THICKNESSES ls{tportpy) Lead Thickness (in.) 8 in. of Beryllium 12 in. of Beryllium 16 in. of Beryllium 1.5 4.7 1.6 0.32 3.0 4.3 1.3 0.26 4.5 3.9 1.0 0.21 6.0 3.5 0.76 0.17 235 ANP PROJECT PROGRESS REPORT A, = neutron relaxation length in alkylbenzene cooling layers (determined to be approximately 6.0 cm), H{a,zpy} = transformation from the neutron dose rate (at the lead surface) from a disk source (LTSF source plate) to the neutron dose rate from an infinite plane source, taken to be ~84/ V(1 = e PR, Heating in Alkylbenzene by Alpha Particles Produced in Alkylbenzene. — In the calculation of the heating in the alkylbenzene from 2.3-Mev alpha-particle production by thermal-neutron capture by the boron, it was assumed that the alpha particle gave up all its energy at the point at which it was produced. The alpha heating in the alkylbenzene was determined by means of the following equation: E'c:qSR(r'IF’b) qukyha(rltpb) = (2.3 Mev) (1.60 x 1013 w/Mev/sec) , Palky! where H i alkyl, ot pp) EC = thermal-neutron capture cross section for boron in alkylbenzene, o, T R c —t,/A (;SR(r,th) = QSLT(z;th) H(a,z) > —e ! lCmf7 ’ Lt 7 brrlzilpy) DOSE RATE OUTSIDE THE ART SHIELD J. B. Dee C. A, Goetz D. K. Trubey The gamma-ray and fast-neutron dose rates at a distance of 50 ft from the ART have been calcu- lated as a function of the thickness of the lead shield. The calculations were based on the recently reported shield design procedure® and on recent LTSF RMR-shield mockup tests and their analysis (see Sec. 13, this report). {The analysis of the L TSF data was based on an effective source plate power value of 2.1 w.) The dimensions and compositions used for the ART were taken from previously reported descriptions.4+3 The gamma-ray dose rate was divided into three parts: (1) primary gamma-ray dose rate originating in or near the core, D,; (2) the dose rate from secondary gamma rays originating in the shield, D; and (3) the gamma-ray dose rate from the heat exchanger, D, . The resulting dose rates are plotted in Fig. 12.7. The details of the calcu- lation have been published,® 3). B. Dee et al., ANP Quar. Prog. Rep. Sept. 10, 4A. P. Fraas, ORNL-1947 op. cit., p 15. 5W. L. Scott, Jr., Dimensional Data for the ART, ORNL CF-55-11-148 (Nov. 21, 1955). 4. B. Dee, C. A, Goetz, and D, K. Trubey, Gamma- Ray Dose Rate from the ART, ORNL CF-56-1-181 (Jan. 9, 1956). 236 GAMMA-RAY DOSE RATES AT 50 ft FROM ART (r/hr) Figo 12.7. ART, thermal-neutron flux in LTSF behind a thickness tpy, of lead. heating in alkylbenzene by alpha particles produced in the alkylbenzene (w/g), 2—01—055-39A NEUTRON DOSE RATE (= 5.2 x 1072 rem/hr) IS INSENSITIVE TO LEAD THICKNESS 4 5 LEAD THICKNESS (in.) Op(50 ft) Gamma-Ray Dose Rate 50 ft from PERIOD ENDING MARCH 10, 1956 13. LID TANK SHIELDING FACILITY R. W. Peelle J. M. Miller Applied Nuclear Physics Division W. J. McCool J. Smolen Pratt & Whitney Aircraft D. R. Otis, Consolidated Vultee Aircraft Corp., San Diego W. R. Burrus, U.S. Air Force Analysis of the dynamic source tests in the second series of experiments with mockups of a circulating-fuel reflector-moderated reactor and shield (RMR-shield) in the Lid Tank Shielding Facility (LTSF) was completed. The analysis is based on an effective neutron power of 2.1 w for the old LTSF source plate {since removed). It had previously been assumed that the power of the old source plate was 3.6 w, but a tentative calibration of the new source plate has indicated the 2.1-w value. ANALYSIS OF THE DYNAMIC SOURCE TESTS ON MOCKUPS OF THE REFLECTOR- MODERATED REACTOR AND SHIELD H. C. Woodsum One of the main purposes of the dynamic source tests in the RMR-shield experiments was to meas- ure the dose rate resulting from fission-product ]Shield Design Group, Pratt & Whitney ot ORNL, gamma rays emitted in the mockup of the fuel-to- NaK heat exchanger. sources of radiation in the heat exchanger, a belt of MTR-type fuel plates? was rotated from the ORNL Graphite Reactor core hole (with the ILTSF source plate removed), where the thermal neutrons In order to mock up the induced fissions, to a slot between two heat ex- changer mockup tanks, where the fission-product gamma rays comprised the source to be studied. Neutron and gamma-ray dose rate measurements were made in the water beyond the mockups. The basic mockup (configuration 17) included a 3-in.-thick lead gamma-ray shield (Fig. 13.1). (The dimensions of all the components in the mockup are given in Table 13.1.) Only two changes in this basic configuration were made throughout these tests: ]]/2 in. of the lead was removed (configuration 17A), and ]/2 in. of the boral next 2R. W. Peelle et al., ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL.2012, p 210. 2-01-057-66-258 Fig. 13.1. RMR Shield Mockup with Dynamic Source (Configuration 17). 237 ANF PROJECT PROGRESS REPORT to the beryllium was replaced with 1 in. of poly- ethylene {(configuration 17F). TABLE 13.1. BASIC RMR-SHIELD MOCKUP {CONFIGURATION 17)* USED IN DYNAMIC SOURCE TESTS Thickness Component (cm) Air 2.40 Aluminum window 0.95 Air 2.23 Fuel plates 0.20 Air 2,92 Inconel 0.32 Air 2.99 Beryllium 30.48 Boral 5.08 Ni-MaF tank (heat exchanger mockup) 5.08 Air 2.92 Fuel plates 0.20 Air 1.27 Ni-NaF tank (heat exchanger mockup) 5.08 Baral 5.08 Nickel 2.54 Air (distributed) 1.50 Lead 3.81 Water 3.10 Lead 3.81 Total 81.96 A, *Configurations 17B . through 17E had infinite, 2.5-, 1.5-, and 1.25-sec transit times, respectively. {n con- figuration 17A (infinite transit time) the last 3.81 cm of lead was removed. In configuration 17F (infinite transit time) the 5.08 c¢m of boral adjacent to the beryllium was replaced with 3.81 e¢m of polyethylene. g | J'“’ n N(E D = — S 2 C(E) —— The cycle time, that is, time for the fuel belt to make one complete revelution, was varied as follows: infinite time (belt not rotating) and 2, 1.5, and 1.25 sec, Shorter cycle times were pre- cluded by the danger of damage to the rig. The rotation speed was set by means of o stroboscope, and the constancy of the speed was checked with a tachometer. The results of the experiment were compared with two calculations of the dose rate; for one of the calculations it was assumed that all the fission-product gamma rays were of a single energy, and, for the other, the spectrum of gamma rays from the fuel belt, as previously measured at the LTSF,? was used. The two calculated values agreed, but they differed by about 30% from the measured value. It was found that this difference could be attributed to the dose buildup factor for water having been used in the calculation; that is, the buildup factor was chosen as if the lead were an equivalent thickness, in mean free paths, of water, Substitution of a new buildup factor for the total mean free paths (lead and water), based on Monte Carlo studies of laminated shields, resulted in agreement between the measured and calculated dose rates, Calculated Dose Rate from Fission-Product Gamma Rays from the Heat Exchanger Based on 2.7 Mev/Photon The first calculation of the dose rate behind the lead-water shield was based on a single gomma- ray energy and an overage number of gamma rays per fission. As a first step in arriving at the proper values for the energy aond the average number of gamma rays, the dose rate from an infinite plane, isotropic, surface source with a spectrum expressed as N(E) was calculoted from the following relation: ](zfliti) Br(z‘#i‘ti) dE ’ E=0 where n = number of fissions per second per watt = 3.1 x 1019, N{E) = number of gamma rays of energy E emitted per fission per Mev, C(E) = flux-to-dose conversion factor for gamma rays of energy E, 238 PERIOD ENDING MARCH 10, 1956 E(Zp;t;) = exponential integral for 2y t. mean free paths in the experiment,® tabulated previously? as [ (e Vdy/y), B (Zp;t;) = dose buildup factor for gamma rays of energy E in water through E“iti mean free paths. The dose rate calculated on the basis of a single gamma-ray energy F “ would then be n ['(E’) m =9 CED B Gpg) B Zu,t,) where ['(E ) is defined as the average number of ‘‘penetrating’’ gamma rays of energy F “per fission and all other terms apply to the single energy. Then, from setting D equal to D J‘wN—(EzE(ZzE)B(EtE)dE B0 C(E) 1 y’i i r ‘ui i INE") = 9 E (g E ) B (St ) C(E ") The value of the numerator was obtained by numerical integration over E, using N(E) = 7.0e~ '*2F gamma rays/Mev-fission.® For relatively thick gamma-ray shields the characteristic or average energy of the gamma rays was found to be about 2.7 Mev; hence E” = 2.7 Mev. By substituting the proper values in the equation, a value of I'(E’) = 0.69 was obtained for a 3-in.-thick lead gamma-ray shield. If it is assumed that all the gamma rays emitted by a source of strength S, are of 2.7 Mev energy, the calculated dose rate behind the lead-water shield can be obtained from the following relation: e..zp.z.Rz.(x,y, £) (n Do pelt = So cos O(y) dA(x,y) ————Br[z,uiRi(x,y,t)] 4nR2(x,y,t) Area of belt where cos 0 = a term to take into account the longer period of exposure of the center portions of the belt, than of the outer portions of the belt, to the activating neutron flux, = 1 ~ (y2/4?) (see Fig. 13.2a), dA = element of area on the belt {cm?), .-.Z,u,.R , . e ' ' = exponential attenuation of 2,7-Mev gamma rays from the belt in the heat exchanger through the various materials (nickel, sodium fluoride, boral, lead, and water) of slant thickness R with a gamma-ray absorption coefficient of p. at energy E, 47R? = inverse-square spreading for a point source from point of emission to point of detection, (Zp,R,) = dose buildup factor for 2.7-Mev gamma rays from a point isotropic source through Zp R, mean free paths of water, x,y,t = coordinates,® as shown in Fig. 13.25, 3Acfuo|[y, the value of g used in this calculation was the same as that given inref. 4, and is slightly different from the t for the present experiment, Tables of Sine, Cosine and Exponential Inteials, vol. ll, National Bureau of Standards, prepared by Federal Works Agency, Works Project Administration, New York ]940 5J. B. Dee et al., ANP Ouar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 185. SFor these coordinates, ¢ rather than z is used to avoid confusion with the common usage of z ot the LTSF for representing the distance from the source plate (that is, from the core hole); z is used as the distance from the core hole in Figs. 13.1, 13.3, 13.4, and 13.5. 239 ANP PROJECT PROGRESS REPORT SECRET 2-01-057-66-259 EXPOSED AREA OF BELT :\1_(J,2/g2) o. DERIVATION OF COS 6. T~ FUEL-PLATE BELT 4. BELT GEOMETRY, Fig. 13.2. Geometry Used in Calculations of the Goamma-Ray Dose Rate Beyond the RMR-Shield Mockups with the Dynamic Source. If the length of the belt in the heat exchanger region is L and its height is 2a (Fig. 13.2¥), then Eqg. 1 can be written in terms of x, y, and ¢: (2) D v,belt =L/2 (Lt /t Wrxl+yZ+s2 ' f fx SV ~ (y/a)? e (& Y 1-31[(2;;,1.352./1!)\/x2 + y2 4 tzldx dy dr(x? + y2 4+ £2) where ¢ is the thickness of each slab of material. This integral can be separated as follows: ) Dy belr L/2 =(>pt, W24y 442 A ’ Br{(zpzfi/t)\/xz +y2 4+ t2] " =0 x2 + y2 + tz Then, since = cos ¢ (see Fig. 13.26) , 240 PERIOD ENDING MARCH 10, 1956 x dx /.2 2 d(cos ¢) = Byt = sin ¢ do , (x2 + y2 + t2)3’/2 or dx = sin ¢ m (x2 ¥ y2 +t2) de and, since ' x = sin¢ , \/x2 -+ y2 + £2 x? + y2 + 2 \/yz + 12 dx = ——————————— d¢p =———dod cos ¢ cosqu When terms in ¢ are substituted for x, the second integral of Eq. 3 becomes : -.'I [(L/2)/m] Z,ut /1) (\/y2+t2/cos ®) E‘u.t. dqu< — V2 4y +12> (4) 0 V2 + 12 Now, let 2, ,2 (—b—>f mbsecd 4y | 0 By = Sin and, since f¢° e—bsec® 44 has been tabulated” as Fl,,b], the total integral, in terms of this 0 7). Moteff, Miscellaneous Data for Shielding Calculations, APEX-176 (Dec. 1, 1954). 241 ANP PROJECT PROGRESS REPORT function, becomes (7) D’y,belt = 5, f“ VU= 0@ B[S vy + 2| Flag, Gpz b + 2] ay 7 0 /y2 + 12 The following constants were used for a numerical integration of Eq. 7: a = 14in. = 355cm, t = distance between the belt in the heat exchanger and the detector = 74.22 cm for this case (Table 13.2), 2ut. = 7.74, as shown in Table 13.2, — L/2 by = sin” . (L/D% + y2 + 2 L/2 = 29.5in. = 75 cm. An upper limit of 45 deg for ¢ is obtained by letting ¥ = 0, and a lower limit of 41 deg is obtained by lefting ¥y = a. For these two values of y, E,u-t- - iti /———-—yz L = 7.74 , for y = 0 , = 8.58 , for y = a For ¢o =41 to 45 deg and b = 7.74 to 8.58, Fl¢,,b] is essentially independent of ¢; hence, F[45 deg, &1 was used. (qSG was assumed to be independent of y.) The values of the various terms used in the numerical integration of Eq. 7 are given in Table 13.3. The integration gave a value of D 1.16 x 1074 s, (me/hr) . v, belt = TABLE 13.2, NUMBER OF MEAN FREE PATHS FOR MATERIALS BETWEEN FUEL PLATES IN HEAT EXCHANGER MOCKUP AND DETECTOR ATt = 74.22 em 1/p (2.7 Mev), pit (2.7 Mev), - Component Thickness p, Den sity Mass Ab.sc.orption Number of {cm) (g/cm>) Coefficient Mean Free (cm?/g) Paths - NaF 3.81 1.59 0.0366 0.22 Ni 3.81 8.90 0.0382 1.29 Pb 7.62 11.3 0.0418 3.60 Boral (Al 5.08 2.53 0.0372 0.48 H,0 51.1 1.0 0.0420 2.15 ’ Air 2.8 0.00293 0.04 0.000 2, = 74.22 2pt, = 7.74 242 PERIOD ENDING MARCH 10, 1956 TABLE 13.3, VALUES OF VARIOUS TERMS USED IN THE NUMERICAL INTEGRATION OF EQUATION 7 / 2 1 = (y/a)* B (b) F[45 deg, b] y V1 — (y/a)2 \/y2 + 2 b Fl45 deg, 5] B (b) for H,0 2, 2 0 1.000 74.20 7.740 1.68 x 10~* 7.38 1.671 x 10~3 5 0.9901 74.31 7.751 167 x 10~* 7.28 .64 x 1073 10 0.9595 74.87 7.809 1.55 x 10~* 7.40 1.470 x 103 15 0.9063 75.71 7.897 1.40 x 10~*4 7.50 1.251 x 10~3 20 0.8211 76.85 8.015 1.27 x 10~* 7.55 1.025 x 1073 25 0.7100 78.30 8.167 1.07 x 10™* 7.65 0.742 x 10~3 30 0.5345 80.04 8.348 0.92 x 10~* 7.82 0.481 x 107> 35.5 0 82.26 8580 0.63 x 10~% 8.40 0 Measured Gamma-Ray Dose Rates With and Without Rotation of the Belt The measured gamma-ray dose rate without the belt rotating (which is analogous to the measured dose rate in the static source tests) includes (8) D‘)/,WOR = DP + DC + Dfp,core ’ where D,y wor = total gamma-ray dose rate without rotation of the belt, D, dose rate from prompt gamma rays, i D. dose rate from capture gamma rays produced in the mockup, dose rate from fission-product gamma rays from the core. It /p,core The measured gamma-ray dose rate with the belt rotating includes (9 Dywe =Pp * Dt Pppug + Py core-he + where D,y wr = total gamma-ray dose rate with rotation of the belt, /b HE = dose rate from fission-product gamma rays from the heat exchanger, Dfp core.yg = dose rate from fission-product gamma rays in the core that were not removed by the belt. The term D, is of course the same in both Eqs. 8 and 9, and it is assumed that the term D - is the same, since the small increase in the number of capture gamma rays from the rotating belt is ignored. Therefore, the difference between the two measured dose rates is D - D = ( ¥, WR v, WOR Dfp,HE + Dfp,core-HE) - Dfp,core Since the belt length is much greater than the core-hole diameter, Dfp,core-HE is negligible and can be ignored; thus (10) Dywr = Py, wor = Pppue = Pppcore 243 ANP PROJECT PROGRESS REPORT Calculated Dose Rate from Fission-Product Gamma Rays from the Core Based on 2.7 Mev/Photon Since Dy belt in Eg. 1 represents the calculated dose rate from the fission-product gamma rays from the heat exchanger, it can be substituted for Dfp gg in Eq. 10. Then, if the value of D is calcu- fp,core lated, the difference between the two measured values (i.e., ny,WR - D’y,WOR) should be equivalent to the ditference between the two calculated values (i.e., D%beh - Dfp,core)' In the mockup the core source was a 28-in.-dia disk, as defined by the boral iris in the LTSF core hole. In the calculation of Dfp core’ 1he source was considered to be isotropic. Hence, S'I ar? (1) Dfp,core = 7 El(zflzxi) - E"I 2'uit:!' T+ (_ri:) Br(z'uzii) ' where §, = equivalent isotropic source strength from 2.7-Mev fission-product gamma rays emitted from the activated portion of the belt at the core hole {mr/hr), E,(Zp,t,) = exponential integral for 3yt mean free paths, as defined previously, Zpt, = total number of mean free paths for materials between the fuel plates at the core hole and the detector, 3.14 + 7.74 = 10.88 (see Tables 13.2 and 13.4), a = radius of exposed area of the fuel plates = 35,56 cm, T = total distance between the fuel plates at the core hole and the detector = 49,88 cm + 74.22 ¢m = 124 cm (Tables 13.2 and 13.4), dose buildup factor for 2.7-Mev gamma rays in water for E_pziz. mean free paths = 10. i Br(zyz't i) Self-absorption was neglected, since the source was thin, When the values are substituted in Eq. 11, it is found that Dy core = 3:05 x 10765 (mr/hr) TABLE 13.4. NUMBER OF MEAN FREE PATHS FOR MATERIALS BETWEEN FUEL PLATES AT THE CORE HOLE AND FUEL PLATES IN THE HEAT EXCHANGER 1 p (2.7 Mev), frt (2.7 Mev), Component Thickness P, Dens;fy Mass Ab'scfrption Number of (cm) (g/cm®) Coefficient Mean Free (cm?/g) Paths U235 0.01 18.7 0.0442 0.008 Al 0.09 27 0.0373 0.009 Ni 1.27 8.90 0.0382 0.432 NaF 2.81 1.59 0.0366 0.222 Boral (Al) 5.08 2.53 0.0372 0.478 Be 30.48 1.85 0.0334 1.885 Inconel 0.32 8.5 0.0382 0.103 Air 8.83 2, = 49.89 2yt = 3.14 244 PERIOD ENDING MARCH 10, 1956 W Evaluation of Source Sfrengifl"§"“3”0 and S, Since D can now be written as v,belt — D[p,core (1.16 x 107%5,) - (3.05 x 10~¢5s.) , the source strength S and S, must be evaluated in terms of the gamma-ray dose rate. The source strength S, of the belt of the core hole is 12 . PBnN 1Ak where P, = power of the belt (watts), n = number of fissions per second per watt = 3.1 x 1019, N = number of 2,7-Mev gamma rays emitted per fission = 0.469, A = exposed area of the belt = 3970 e¢m?, K = flux-to-dose conversion factor for gamma rays of 2.7-Mev energy = 2.45 x 102 (photons/cm?2.sec per mr/hr). Thus Eq. 12 becomes §, = 2.2 x 104 Py (mr/hr) . The source strength S, for the center of the belt in the heat exchanger region is L'l (13) S =— S, . LZ where L, = diameter of exposed area of belt = 71 cm, L, = total length of belt = 400 cm. Thus Eq. 13 becomes Se = 3.91 x 10° P (mr/hr) Determinction of the Power of the Fuel Belt The power of the fuel belt P, was determined by comparing the fast-neutron dose rate (Fig. 13.3), the thermal-neutron flux (Fig. 13.4), and the sodium-activation curves for configuration 17 (no rotation) with the corresponding curves for a similar configuration (3-C) used in the static source tests with the old LTSF source plate.2 The ratio of the fast-neutron dose rates was found to be 1.65 and that for the thermal-neutron flux was 1.37. These two values, when averaged with the ratio of the sodium acti- vations2:8 (1.46), give a final average of 1.49. Thus, the power of the belt? is 1.49 times the previously assumed effective LTSF source power of 2.1 w, or Pg = 3.1w . Comparison of Measured and Calculated Gamma-Ray Dose Rates The data accumulated above give D = (1.16 x 10=4(3.91 x 10%)(3.1) (mr/hr) , v,belt 8. 1. Chapman et al., ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 197. 9This new power value was also determined by the LTSF staff by comparing neutron measurements taken in water. 245 ANP PROJECT PROGRESS REPORT 2—01-0587-86-260 2—-01-057-66—261 10 T T T 1. { T T ; ‘ '[ CONFIGURATION 174 ] ® CONFIGURATION {7A 7=, 1.5 in OF Pb T=o,15in0F Pb 7 5 0 CONFIGURATION 17B t ==, 30 in. OF Pb _ (IZ_ONFIG%ROA;I%I} iF:be — i = m _ LD - HA CONFIGURATION 17C r = 25sec, 3.0in. OF Pb CONFrE;UHAT\ON e 1 ! A CONFIGURATION 17D v =15 sec, 3.0in. OF Pb - T=25 sec, 30 in OF Pbi;,,,, B a CONFIGURATION 17E 7 = 1.25 sec, 3.0in, OF Pb CONFIGURATION 17D b T 5> |0 CONFIGURATION {7F t =, 3.0 in. OF Ph, fin. OF _ T =15sec,30in OF Pb @ = POLYETHYLENE CONFIGURATION 17E ] T=125sec,30in.OF Pb .| = TRANSIT TiME OF FUEL BELT CONFGURATION 17F 10— - = T=o,30in OF Pb,1in. ———Ho ‘j -] e oo AN OF POLYETHYLENE ‘ T . B N. 5 — T - \ \ - g W ‘T = TRANSIT TIME OF FUEL BELT | 3 - v \ A ‘ - | - \ . _ S 2 g £ A = TN B | e \ER‘—" £ = 2 AR s : % E \ | & b \ ‘ x ! (ZD Lt & t \ - = T > o N\ o = N . = o R A 4 E 5 4 s - > - = o 2 \“{\ 5 ] | h : - _ & \\ 2 . A 1 i/ ! Ao —f 10 SN ~ N 3 AN 5 ——— ' \\ R \ :| 5 l \ 70 20 110 130 150 170 | z, DISTANCE FROM LTSF CORE HOLE (cm) 152 | 80 30 100 110 120 130 . .. 7 DISTANCE FROM LTSF CORE HOLE (om) Fig. 13.4. Thermal-Neutron Flux Beyond RMR- Shield Mockups with Dynamic Source. Fig. 13.3. Fast-Meutron Dose Rates Beyond RMR-Shield Mockups with Dynamic Source. Dfp,care (3.05 x 10=%)(2.2 x 104)(3.1) (mr/hr) . Thus the difference in the calculated values is 1.18 mr/hr D’y,belf Dfp,core The actual difference in the measured gamma-ray dose rates at z = 130 cm (z is the distance from the core hole and z = 130 cm corresponds to T = 124 cm), as may be seen from Fig. 13.5, is 4,55 - 3.65 = 0.9 mr/hr , which is a factor of 1.3 less than the calculated value. Calculated Dose Rate Based on Fission-Product Gamma-Ray Spectrum In order to check the calculations, a separate calculation was made for which the fission-product gamma-ray spectrum measured at the LTSF2 with the same rotating fuel belt system was used. A 246 2-0t-057-66-262 103 — . — — : [ | 777 T ® CONFIGURATION {7A ==, 15in. OF Pb —— 4 o CONFIGURATION 178 7 ==, 3.0in. OF Pb - Sr—— A GONFIGURATION 176 7 =2 sec, 3.0in.OF Pb_ . & CONFIGURATION 17D T =15sec, 3.0in. OF Pb __| m CONFIGURATION 17E T =1.25sec, 3.0in. OF Pb__ 2 "\ y T D CONFIGURATION 17F o | T=wm, 30 in. OF Pb, 1in. OF POLYETHYLENE] 10° TN e e E — - TIT - LTI oo - w w 8 N > ? - < & i ! 10 b——- [ 2 "= —“BACK OF LEAD == = - I ] = - ‘ Q 5 ; T = TRANSIT TIME OF FUEL BELT ~ o 2 f— e ‘ T NG | ! : | i | \ 1 70 BO 90 100 "o 120 130 140 {50 160 2, DISTANCE FROM LTSF CORE HOLE fcm) Fig. 13.5. Gamma-Ray Dose Rate Beyond RMR- Shield Mockups with Dynamic Source. integration was carried out over the The contribution to the dose rate from gamma rays with energies out- side this range is negligible, and, by neglecting numerical spectrum from 1 to 5 Mev. it, only a very small error should be introduced. The method of calculating the gamma-ray ottenu- ation in this case was the same as that used for the 2.7-Mev gamma rays. Atfenuations for discrete energies of 1, 2, 3, 4, and 5 Mev were calculated. If the attenuation of the gamma-ray dose rate at a distance z is represented as A(E,z), the total gamma-ray dose rate at some particular distance may be calculated as 5Mev N(E) A(E,z) dE D(z) = C _ 7, 1 Mev K(E) where N(E) = number of gamma rays of energy E, as given by the fission-product-decay gamma-ray spectrum, K(E} = flux-to-dose conversion factor for gamma- ray energy E, C = constant related to the power per unit area of the belt, The integrals for the dose rate both from fission- product gamma rays emitted in the heat exchanger PERIOD ENDING MARCH 10, 1956 region (rotating belt) and from those emitted in the core region (no rotation) were evaluated by this method, and the difference between the two should be equivalent to the difference in the experimental gamma-ray dose rate with and without the belt rotating, The difference between these more exact integrations was actually 1.17 mr/hr, which was almost the same as the 1,18 mr/hr previously cal- culated on the basis of a single (2.7-Mev) gamma- ray energy and a slightly different spectrum, Discussion of Results in order to resolve the 30% discrepancy between the calculated and measured dose rates, it was necessary to investigate the assumptions involved in the calculation. The most obvious uncertainty was in the choice of a buildup factor through a multiregion shield. Since the dose buildup factors in lead and in water differ by a factor of 2, a considerable error could be introduced by using either one for the total number of mean free paths in the shield. The buildup factor for water for the specified number of mean free paths was used in all the calculations. It was thought that the total buildup would be more characteristic of the water, since about 50 cm of water (2.15 mfp) followed the 3 in, of lead (3.6 mfp). However, some recent Monte Carlo calculations by S. Auslender!® indi- cate that for 3-Mev gamma rays through 8 mfp of lead and water (4 mfp of lead followed by 4 mfp of water) the buildup factor is about 30% lower than the buildup factor through the same number of mean free paths of water alone. Auslender’s calcu- lations indicate that a correction should be applied to the buildup factor that would reduce it to account for the presence of the lead. Such a correction would bring the calculations and the experimental results into closer agreement. Furthermore, if it were the choice of the water buildup factor alone that gave a calculated dose rate that was higher than the experimental dose rate, the discrepancy should decrease if more water were placed be- tween the lead and the detector (or the point of calculation). This theory was tested with another calculation for a z distance of 160 cm, which would place about 3.4 mfp of water behind 3.6 mfp of lead. In this case the calculated gamma-ray ‘OShielding Analysis Group, Pratt & Whitney ot ORNL, 247 ANP PROJECT PROGRESS REPORT dose rate obtained by using a buildup factor charac- teristic of water alone was 0.26 mr/hr, which is only about 20% higher than the experimental dose rate of 0.22 mr/hr. Thus, since the gamma-ray dose from the fission products in the heat exchanger of a circulating- fuel reactor is only about 30 to 40% of the total dose in the crew compartment, the error incurred by using the water buildup factor for the total number of mean free paths is not great for moderate lead thicknesses. Of course, as shown by these dynamic source experiments at the LTSF, it would be better to use buildup factors more characteristic of the lead and water combination in the shield. tn the experiment, the gamma-ray dose rate from fission-product gamma rays emitted in the heat exchanger region did not change (within the limits of experimental error) as the transit time was decreased from 2 to 1.25 sec (Fig. 13.5). This indicates that the fission-product gomma-ray spectrum did not change significantly, which substantiates the conclusions reported previously.? The thermal-neutron flux increased as the transit time of the belt decreased (Fig. 13.4). The in- crease was, of course, due to more of the delayed being emitted in the heat exchanger rather than in the core region. Since the average energy of the delayed neutrons is lower than the average energy of the prompt neutrons, the delayed nevtrons 248 neufrofii would, essentially, all have been lost in the shield if they had been emitted only in the core. However, when they were emitted in the heat exchanger, they penetrated the shield and added to the measured thermal-neutron flux. The maximum increase was about 50% at the outer surface of the lead gamma-ray shield. This per- centage decreased to about 10% beyond about 20 cm of water and to a negligible fraction beyond 50 cm of water. The discontinuity in the thermal-neutron flux curves at z = 90 ¢m is due to a change in instru- ments (from a 3-in. fission counter to a 121/2-in. BF, counter) at this point. Although both instru- ments were normalized to the same value in ploin water, the center of detection of the ]2'/2~En. BF, counter moved toward the front of the chamber (to the left in z) as the slope of the resulting thermal- neutron flux became steeper. It is believed that this shift in the center of detection completely accounts for the difference in measurement by the two instruments. The fast-neutron dose rate (Fig. 13.3) beyond 6 cm, or more, of water was independent of transit time, and hence the delayed neutrons did not contribute significantly to the fast-neutron dose rate. The 1 in. of polyethylene plastic inserted in the reflector region (configuration 17F) decreased the fast-neutron dose rate by 20 or 30%. PERIOD ENDING MARCH 10, 1956 14, BULK SHIELDING FACILITY F. C. Maienschein T. V. Blosser E. B. Johnson G. deSaussure J. D. Kington G. Estabrook T. A. Love K. M. Henry E. G. Silver W. Zobel Applied Nuclear Physics Division A. T. Futterer, Pratt & Whitney Aircraft GAMMA-RAY STREAMING THROUGH THE NaK PIPES THAT PENETRATE THE ART SHIELD T. V. Blosser Preliminary calculations by the ORNL Shield Design Group' have indicated that gamma-ray streaming through the NaK pipes that penetrate the ART shield would increase the dose rate in line with the pipes by a factor of roughly 2 x 103 for the original 7-in.-thick lead shield. For the currently planned 4.3-in.-thick lead shield the dose rate in line with the pipes would be only a factor of 40 more than it would be without the pipe pene- trations. An experimental program has now been initiated to measure the dose rates in a mockup of the ART shield and NaK pipes and to determine the most effective pipe arrangement for reducing this high leakage. The experiment will be performed in the aluminum thermal column tank atop the ORNL Graphite Re- actor, The gamma-ray source, which will have a large energy spread, will consist largely of 4.5- Mev capture gamma rays from the thermal column graphite plus a spectrum of capture gamma rays (0.01 to 7 Mev) from a 0.040-in.-thick (12-in.-dia) cadmium disk placed in the bottom of the tank (Fig. 14.1). A 0.25-in.-thick boral plate, placed over the cadmium, which will cover the bottom of the tank, will prevent a background of capture gamma rays from originating in the lead shield and in the water; however, the boral will produce 0.5- Mev capture gamma rays in the source area, The boral capture gamma rays produced outside the source area will be sufficiently shielded by the 4.3 in. of lead and will not cause a disturbing background. There will be no appreciable attenu- ation by the boral of cadmium capture gamma rays. p. K. Trubey, private communication, to T, V, Blosser, The bottom of the aluminum tank will also produce a spectrum of capture gamma rays in the source area. Various shield and pipe configurations will be inserted in a two-step square opening in the lead at the bottom of the tank. The steps of the opening will be about 2 in. high, the lower section being 12 in. square and the higher section 16 in. square. Aluminum powder (~0.8 g/Cm3) will be used to simulate the NaK in the pipes. The following configurations will be tested: 1. solid lead; 2. 3.8-in.-dia pipe perpendicular to bottom - air- filled; aluminum-filled; 3. 3.8-in.-dia pipe at a 45-deg angle to bottom - air-filled; aluminum-filled; 4. ART mockup (Fig. 14.2). The design of the support for the shield mockup has been completed, and the materials are on order. 2-01-058-0-4 12-in.- SQUARE OPENING FOR DUCT MOCKUP TGP OF REACTOR SHIELD~ CONCRETE i S A e e R oy -in-DIA CADMIUM DISK BOTTOM OF TANK Y4 in. OF BORAL SEPARATIONS ARE INCREASED TO SHOW PARTS \ Fig. 14.1. Thermal-Column Tank Arrangement for the ART Shield Mockup Experiment. 249 ANP PROJECT PROGRESS REPORT PATCH TO BE BUILT UP AS REQUIRED FOR THIS EXPERIMENT 3.625-in.-0D % 0.25-in.-WALL PIPE ofb FLREKS SN QR %fi, o —in - ’ 2-1-058-C-5 <7 SELKERN OSSR 4%%5% SO N CINSULATION > - i ORI OO R KIS SRR B BSOS SN 2Y>-in. SCH 40 PIPE \ (2.875~in. OD x 2.469-in. D) Fig. 14.2. ART Shield and NaK Pipe Mockup Arrangement. DECAY OF FISSION-PRODUCT GAMMA RADIATION T. A. Love R. W. Peelle It was pointed out previously? that information aMéut*the decay characteristics and the photon U235 W. Zobel energy spectrum of the fission products of for short times after fission is essential in the design of an optimum shield for a circulating-fuel type of aircraft reactor, In such a reactor the fuel is circulated through a heat exchanger, located within the reactor shield, which thus constitutes a secondary source of radiation. Preliminary measurements of the time dependence® of the fission-product gamma-ray mixture in the range from 5 to 150 sec, as well as some information on the 2R, W. Peelle et al.,, ANP Quar. Prog. Rep. June 10, 1955, ORNL-1896, p 203. 3R, W, Peelle et al.,, ANP Quar. Prog. Rep. Sept. 10, 1955, ORNL-1947, p 201. 250 energy spectrum of the mixture,* have already been reported., An extension of the first experiment to other time periods is reported here. The equipment used was described previously.® Samples of enriched uranium, which weighed about 2, 7, and 16 mg, respectively, were irradiated in the ORNL Graphite Reactor for periods of 0.8 or 32 sec. Energy calibration of the multiple-crystal spectrometer was carried out in the usual fashion, with Hg293, Na22, Zn85, Na24, F20, 7,90% ThC” and N'6 sources. The time analyzer was calibrated by using a 60-cyele pulser., To determine the efficiency of the spectrometer as a function of photon energy, the absolute strengths of Hg203, Na?2, Cs'37, Zn%5, and Na24 calibration sources were determined with the aid of the calibrated high-pressure ion chamber of the ORNL Radio- isotopes Control Laboratory. A correction was applied to this efficiency to take into account the 4R. W. Peelle et al., ANP Quar. Prog. Rep. Dec. 10, 1955, ORNL-2012, p 223. peak-to-total ratio of the spectrometer. The cor- rection factor is the inverse of the peak-to-total ratio. The data were corrected for counting losses in the electronic equipment, a correction which in some coses amounts to as much as 25%. Since, as before, energy groups were used and since the efficiency of the spectrometer varies considerably over a single group, the average efficiency for an energy group was determined by weighting with the energy spectrum determined in the experiment carried out at the LTSF.4 It is expected that this average efficiency is more correct than that used in the evaluation of the previously reported datgq, but it must be realized that, if subsequent experi- ments show a large variation of the spectrum with time, a renormalization of the data may be neces- sary, The determination of the number of fissions occurring in a sample depends on the length of time the sample was in the reactor and on the thermal-neutron flux available. The error in the length of time amounts to at least 15%. The effective thermal-neutron flux available, as well as the macroscopic cross section of the sample, was taken to be the same as in the previous experiment; that is, the value obtained by Moteff® by using the gold-foil and cadmium-difference technique was used for the effective thermal- neutron flux, and thermal cross sections were used for gold and U233, The decay rate is shown in Fig. 14,3 as a function of time after fission for various energy ranges. It includes a re-evaluation of the previ- ously reported data. Again, the energy group be- tween 1.62 and 2.3 Mev was investigated both with the Compton (two-crystal) and the pair (three- crystal) spectrometer. |t should be noted that the agreement between these curves is reasonable if the large statistical error associated with the pair- spectrometer data is considered. Figure 14,4 is a cross plot of the data presented in Fig. 14.3, and shows the photon energy spectrum 3], Moteff, private communication, to R, W, Peelle. PERIOD ENDING MARCH 10, 1956 as measured 1.5, 10, 100, and 1000 sec after fission. The results obtained by integrating the curves of Fig. 14.3 between 1.25 and 1600 sec are given in Table 14,1. The uncertainty in the total number of photons per fission and the energy per fission is probably about £25%. This work concludes the preliminary analysis of the first part of the investigation of fission- product gamma rays, that is, the detailed time behavior of relatively large energy groups of fis- sion-product gamma rays, The next phase of the experiment will concern the detailed energy spectra at certain time intervals after fission. As was mentioned above, this phase may have an influence on the phase just concluded through the weighting factor used for the efficiency. It is expected that after the conclusion of this second phase of the experiment a more detailed report will be prepared that will cover the entire investigation on fission- product gamma rays. It is hoped that an expression relating the gamma-ray flux to the energy of the emitted gamma rays and the time after fission can be included in the report that will make possible calculations of the flux for any energy and time, TABLE 14.1. MEASURED VALUES OF PHOTON INTENSITY PER FISSION AND TOTAL ENERGY RELEASE PER FISSION INTEGRATED BETWEEN 1.25 AND 1600 sec AFTER FIS5SION Energy Photons per Energy per Range Fission Fission (Mev) Compton Spectrometer 0.28~0.51 0.696 0.275 0.51-1.12 1.103 0.894 1.12-1.62 0.428 0.586 1.62-2.31 0.210 0.412 Pair Spectrometer (1.62-2.31) (0.197)* (0.385)* 2.:?—3.5 0.178 0.515 3.5=-5.0 0.038 0.057 Total 2.67 2.92 *Not included in total. 251 ANP PROJECT PROGRESS REPORT UNCLASSIFIED 2-01-058-0-6 o — 0.28~0.51 Mev (COMPTON) 1 Pl I {12 -1.62 Mev (COMPTON - b e, o.5h—‘1.g2 Mev (COMPTON) '\( | | ] | . X L DECAY RATE (photons/ Mev - sec - fission ) 1.62— 2.3 Mev (PAIR) 1.62-2.3 Mev (COMPTON -2.3~3.5 Mev (PAIR) ——3.5—-5.0 Mev (PAIR) { 2 5 10 2 5 10 2 5 100 2 5 10 2 5 10 T, TIME AFTER FISSION (sec) Fig. 14.3. Fission-Product Gamma-Ray Decay Rates as a Function of Time After Fission for Six Photon Energy Groups. 252 PERIOD ENDING MARCH 10, 1956 UNGLASSIFIED 2-01-058-0-7 t | ’ ——COMPTON SPECTROMETER F———PAIR SPECTROMETER 1o ! l——! [ L _ e | 0 2 -1___._;..__! S | I | c ' —Il | i £ ; , mmm———= - 11.5 sec i f i Lot | S a . L 10 sec 2 = = — e —— ~ | g | 5 10"4 ) Sy o 0.::“_’ | Lo — 10%sec T T | _5 1 1077 — Sommmme I | I L 3 T -1 10" sec 10 € C e 1077 y ; 0 1 2 3 4 5 & ENERGY (Mev) Fig. 14.4. Histogram of the Fission-Product Photon Energy Spectrum for 1.5, 10, 100, and 1000 sec After Fission. 253 THE AIRCRAFT NUGLEAR PROPULSION PROJECT AT THE OAK RIDGE NATIONAL LABORATORY MARCH 1, 1956 ANP PROJECT DIRECTOR W. H. JORDAN RD CO-DIRECTOR S. J. CROMER RD ASSISTANT DIRECTOR A. J, MILLER RD D. HILYER, SEC. RD REPORTS PRATT & WHITNEY AIRCRAFT A. W, SAVOLAINEN ARE E.R.DYTKO, ASST. PROJ. ENG. PWA N. J. HARPER ARE A. GIANGREGORID, ADM. ASST. PWA H. C. GRAY, DESIGN ENGINEER PWA LITERATURE SEARCHES A. L. DAVIS ARE AIRCRAFT REACTOR ENGINEERING DIVISION SUPPORTING RESEARCH S. J. CROMER, DIRECTOR RD ANP STEERING COMMITTEE W. H. JORDAN A. B. LIVINGSTON, SEC. ARE W. H. JORDAN, CHAIRMAN A. J. MILLER E. P. BLIZARD W. F. BOUDREAU G. E. BOYD S. 1. CROMER W. K, ERGEN A. P, FRAAS W, R, GRIMES ASSISTANT TO DIRE CTOR G. W. KEILHOLTZ R. 5. CARLSMITH ARE W. D. MANLY A. J. MILLER A M. PERRY METALLURGY H. F. FOPPENDIEK W. D. MANLY, STAFF ASSISTANT M H. W. SAVAGE E. D. SHIPLEY J. A. SWARTOUT A. M, WEINBERG EXPERIMENTAL ENGINEERING CHEMISTRY H. W. SAVAGE ARE W. R. GRIMES, STAFF ASSISTANT MC NOTE: THIS CHART SHOWS ONLY THE LINES OF TECHNICAL COORDINATION OF THE ANP PROJECT. THE VARIOUS INDIVIDUALS AND GROUPRS OF PEOPLE LISTED IN THIS AND THE FOLLOWING CHARTS ARE EN- GAGED EITHER WHOLLY OR PART TIME ON RESEARCH AND DESIGN WHICH 15 COORDINATED FOR THE BENEFIT OF THE ANP PROJECT IN THE MANNER INDICATED ON THE CHART. EACH GROUP, HOWEVER, POWER PLANT ENGINEERING IS ALSO RESPONSIBLE TO ITS DIVISION DIRECTOR FOR THE DETAILED PROGRESS OF THE RESEARCH SHIELDING A. P. FRAAS ARE AND FOR ADMINISTRATIVE MATTERS, E. P BLIZARD. STAFF ASSISTANT AP THE KEY TO THE ABBREVIATIONS USED IS GIVEN BELOW. AC ANALYTICAL CHEMISTRY DIVISION ~ ORNL INSTRUMENTATION AND CONTROLS AP APPLIED NUCLEAR PHYSICS DIVISION — ORNL RADIATION DAMAGE . ic ARE AIRCRAFT REACTOR ENGINEERING DIVISION — GRNL BMI BATTELLE MEMORIAL INSTITUTE G. W. KEILHOLTZ, STAFF ASSISTANT 55 ENGINEERING DESIGN H. C. GRAY PWA REACTOR PHYSICS A. M. PERRY ENR 7503 W. F. 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