MARIETYA ENEQ.E'{l SYSTT-‘S LIBRARIES R HRHA 45t 03L0OLED 9 ARTIN | ! I ; y ORNL -1609 This document consists of 172 pages. Copy Q'J?of 262 copies. Series A, Contract No. W-7405-eng-26 ATRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT For Period Ending September 10, 1953 R. C. Briant, Director A. J. Miller, Assistant Director W. B. Cottrell, Editer DATE ISSUED 0T G 13 QAK RIDGE NATIONAL LABORATORY Operated by CARBIDE AND CARBCN CHEMICALS COMPANY A Division of Union Carbide and Carbon Corporation Post 0ffice Box P 0ak Ridge, Tennessee (TR 3 4456 03L0OLAO 9 1. A 2. R. G. Agfel 3. C. B. Balgock 4, C, J, Bartan 5. E. S. Bettisi 6. D. S. Billingten 7. F. F. Blankensh# 8. E., P. Blizard 9. M. A. Bredig 10. R, C. Briant 11. F. R. Bruce 12. A. D, Callibhan 13. D. W, Cardwell 14. J. V. Cathcart 15. C. E. Center 16. R. A. Charpie 17. J. M, Cisar 18. G. H. Clewettd 19. C. E. Cliffg¥d 920. W. B. Cottgéll 21. R. G. ochfian 22. D, D, Cow&n 23. F. L. Cgfiler 24. W. K. Egster 25. L. B. Emlet (Y-12) 26. W, K.fErgen 27. A, P;gFraas 28. W. fififGall 29, C. B. Graham 30, W.#. Grigorieff (consultant) 31. WgR. Grimes 32. ti Hollaender 33, ;fi S. Householder 34. T, Howe 35. iW. B. Humes (K-25) 36.‘§{. W. Johnson 37. 2G. W. Keilholecz 38.F P. Keim 39, M. T. Kelley 40. F. Kertesz 41. E. M. King INTERNAL DISTRIBUTION ORNL- 1609 Progress 42. J. A. Lane 43. C. E. Larson 44. R. S. Livingston 45. R. N. Lyon 46, W. D. Manly 47. L. A, Mann 48. W. B. McDonald 4-9. Jo Lo Meem 50. A. J. Miller 51. K. Z. Morgan 52. E. J. Murphy 53, H. F. Poppendiek 54. P. M. BReyling 55. H. W, Savage 56. E. D. Shipley 57. 0. Sisman tg, L., P. Smith (consultant) 59, A. H. Snell 60. C. L. Storrs 61. C. D. Susano 62. J. A. Swartout E£. H. Taylor P. M. Uthe E. R. VanArtsdalen F. C. VonderlLage J. M. Warde A. M. Weinberg J. C., White E. P, Wigner (consultant) G, C. Williams J. C. Wilson C. E. Winters . ANP Library 84. anlogy Library ) L&boratory Records Department 90. Lafinratory Records, OBRNL R.C. 91. 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Metak}urgy Library 93, Reactar Experimental ing Library Central ‘Besearch Library 111 103-114. 115. 116-123, 124. 125. 126-130. 131. 132. 133-134. 135-140. 141. 142, 143. 144. 145. 146-149, 150. 151. 152-154. 155-158. 159, 160. 161-168., 169, 170-177. 178-179. 180-181. 182. 183. 184, 185. 186-188. 189-192. 193. 1v ti Air Force Engineering Office, Oak Ridge % Air Force Plant Representative, Burbank EXTERNAL DISTRIBUTION ®1r Force Plant Representative, Seattle g@f Xfip Force Plant Representative, Wood-R%fi%e ANE, Project Office, Fort Worth xf Argdgne National Laboratory (1 copy, to Kermit Anderson) Armeffiborces Special Weapons Projeg% {Sandia) Atomlcfiynergy Commission, Wash1ng&on Battell® Memorial Institute Bechtel %wrporatlon Brookhavenfiéatlonal Laborator Bureau of A&ronautics (Graqg Bureau of Shfifi Chicago Patent G%_ Chief of Naval Re& duPont Company, Afig duPont Company,dflw Foster Wheelerg b General Electfic Company, EKPP General Flec#ric Company, R?%hland Hanford Opefations Office %% USAF- Headfifdrters Office of A& Idaho Opejat1ons Office (1 copy® Towa Stafe College Knollsfgfiomlc Power Laboratory ;. Lockl fid Area Office % Los ‘-amos Scientific Laboratory ; Massgchusetts Institute of Technology (Benedict) Masghchusetts Institute of Technol ogy * {au fmann) Ma;;rlals Laboratory (WADC) (Col. P. L. igill) Mc?santo Chemical Company % M sund Laboratory ffi mtional Advisory Committee for Aeronautic% Cleveland £ (3 copies to A, Silverstein) 10 Phillips Petroleum Company) {National Advisory Committee for Aeronautics, Washington B 194, 195-196. 197-198. 199-204., 205. 206. 207-216. 217-228. 229-230. 231. 232. 233. 234. 235-236. 237-238. 239. 240. 241-246. 247-261. 262. %@aval Research Laboratory %fiw York Operations Office Ndfigh American Aviation, Inc Nuc¥gar Development Assocj Paten® Branch, Washingtog Pionee®Service and Engfneering Company Pratt ar Whitney Aj#€raft Division (Fox Project) Powerplan% Laborat#fy (WADC) (2 copies to B, Bleaman) Rand Corpofigpion},f copy to V. G. Henning) San Franciscg @erations Office Savannah Rivefg Operations Office, Augusta USAF Headqugf t&gs U, S. Navgh Radi@logical Defense Laboratory Universigt of Cal®%fornia Radiation Laboratory, Berkeley Univerghty of CaliBgrnia Badiation Laboratory, Livermore orporation oMgAmerica Walg#r Kidde Nuclear %aboratories, Inc. Wegfinghouse Electric @prporation #thnical Information SHrvice, Oak Ridge, Tennessee Furtis-Wright Corp., Wri?%t Aeronautical Daivision (K. Campbell) vi Reports previously 1ssued in this ORNL-528 ORNL-629 ORNL-768 OBRNL-858 ORNL-919 ANP-60 ANP-65 OBNT.-1154 ORNL-1170C ORNI.-1227 CRNL-1294 ORNL-1375 ORNL-1439 ORNL-1515 ORNL.-1556 Period Period Period Period Period Period Period Period Period Period Period Period Period Period Period Ending Fnding Ending Ending Ending Fnding Ending Ending Ending Ending Ending Ending Ending Ending Ending series are as follows: November 30, 1949 February 28, 1950 May 31, 1950 August 31, 1950 December 10, 1950 March 10, 1951 June 10, 1951 September 10, 1951 Decemker 10, 1951 March 10, 1952 June 10, 1952 September 10, 1952 December 10, 1952 March 10, 1953 June 10, 1953 CONTENTS FOBEWORD . [ ] . * . L] . . . . * . * . * * * * * PART I. REACTOR THEORY AND DESIGN INTRODUCTION AND SUMMARY . . « « &« « + & ¢« & & 1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT . The Experimental Beactor o+ « « « ¢ + + & Calculation of critical mass . . « . Reactor design .+ + o« s « & o+ o o o« & Reactor control . .+ « ¢ ¢« + o« « « « & Flow characteristics in fuel tubes . External Fuel Circuits . . « « ¢« « . « & Fuel system design .« « o« « « « « ¢ & Sodium system design « 4 o« o s o o Pumps ¢« v & @ 0o 4 o o s s s 4o s s e e Valves o ¢ o v v ¢ o o o o o o & s The NaF-ZrF,-UF, Fuel . « ¢« « « « « « ¢ & ComposSition « o+ o« « o o o s o o o« o » Physical properties « ¢« + o« o« + o + = Fuel solvent production « . « o o« o+ & Fuel concentrate production . + « + Corrosion of Inconel .+« . . ¢« . 4+ « & Instrumentation . « + « o o o o « ¢ ¢ o Auxiliary Systems o+ 4 « o o ¢ o o s o o » Off-gas system . o« o ¢ o o & 2 2 o & Electrical system . » o« « « &« o + o+ & Gas mONitoring SYSLEM « « « o« o o o 2. EXPERIMENTAL REACTOR ENGINEERING . . . . . Pumps for High-Temperature Liquids . . . Frozen-sodium-sealed pump for sodium Gas-sealed sump pump for the ARE . , Combination packed-frozen sealed pump Frozen-lead-sealed pump for fluorides Allis-Chalmers canned-rotor pump . . - o - - . - * . . . - a - . » . - * - » - » * * - » . . . . . . s .« . .« s . » for . . . - L] * » . * - - » . - e s s e u « o on e e fluorides e v e e s . » - * * Rotary-Shaft and Valve-Stem Seals for Fluorides . . . & Spiral-grooved graphite-packed shaft seals Annular-grooved shaft seal . . . . . V"riflg Sea}. - - - » - . - - - - . - - * . . . + a . - - . Graphite—BeF2 paCking « & % 3 & e 4 & ® s mn s s s = Bronze-wool and MoS,-packed frozen seal . . . . . . Bronze-wool, graphite, and MoS,-packed frozen seal Copper-braid and MoS,-packed seal test L * 2 . e * vii Frozen-lead seal . . . . ¢ ¢ v ¢« v v & o Vitreous seals o o ¢ 4 ¢ 4 o o o o o s o o Graphitar-Graphitar shaft seal . . . « . . Packing-penetration tests . . « « & « & + & High-Temperature Bearing Development Program . Instrumentation « « o + o o o s o o o o o o 3+ o Fluoride leak detection . o+ « « « o« o « & Sodium leak detection + « ¢« « ¢ o+ ¢ « « o o Flow measurement . ¢« ¢ ¢ « o « o s o o« »+ & 3. ATIRCRAFT REACTOR DESTIGN STUDIES . . « « « + « & Shield Designs for Reflector-Moderated Reactors Comparative Analyses of Liquid Metal Coolants . Activation of rubidium, potassium, and sodi Effect of secondary coolant on power plant Power Plant Performance and Control . . . « . . Description of power plant . . « « o o o Design performance . . « ¢« « ¢ & & o« + « System characteristics that affect reactor 4., UNIFOBM THERMAL-NEUTRON FLUX REACTOR . . . . . . Critical Mass . & v 4 v v v v o s o o o & o o Neutron Flux . « « ¢ ¢ v o ¢ ¢« o v ¢« ¢« ¢ o o & Importance Fumction . , + ¢« « « ¢ o « v o &+ 4+ & PART I1. MATERIALS RESEARCH INTRODUCTION AND SUMMARY ¢ « v v v ¢ ¢ ¢ ¢ ¢ & o o . 5. CHEMISTRY OF HIGH-TEMPEBATURE LIQUIDS . . . . . Thermal Analysis of Fluoride Fuels ., ., . . . . NaF-ZrF, -UF, + « o o ¢ v v o v v 0 o v s KF-ZrF, UF e e e s e s e e e e Systema contalnlng ThF e e e e e UF,-ZrF, . . . v v v v v v v v v UF -UF, ZrF e e e e e e e e e NaF ZrF UF -UF,. . e e e e e s Thermal Analy51s of Ch10r1de Fuels P e e e e e NaCl-UCl, . . ¢ « ¢ v ¢ v v o v o v KCI-UCL, . . « o ¢ v v v v v v v v v CsCl-UCL, . . v ¢ ¢ v v v v v v o v CaCl,-UCL, « ¢ v v ¢ v v v v v v v v v v Thermal Analysis of Fluoride Coolants . . . . . Quenching Experiments with Fluorides . . . . . viii ight . - - » . . . ntrol » . . * . . - . » - - e 26 26 27 28 28 29 29 29 30 31 31 37 37 37 38 39 43 43 47 47 49 50 53 55 55 56 56 57 57 57 57 57 58 58 59 59 59 59 61 Differential Thermoanalysis of Fluorides . . . + ¢« & ¢« ¢« o & Phase Equilibria of Fused Salts by Filtration . . « « + &« + & X-Ray Diffraction Studies of Fluorides . + ¢« ¢ ¢« o« o ¢ & o & NaF-UF, « « ¢ ¢ 0 v v v v v v i v o v s s s e e e s e v s NaF- ZrF e s e e e e Fundamental Chem1stry of High-Temperature L1qu1ds ¢ e a s s e Spectrophotometry in fused salts .+ & 4 ¢ ¢« & ¢ ¢ « s+ o+ Thermogalvanic potentials in liquids « « + ¢« « o« o o o « EMF measurements in fused salts . . . ¢« & ¢ & &« o & + o+ o Production and Purification of Fluoride Mixtures .+ . .+ + + Treatment of NaZrF, melts with metallic zirconium . . . . Treatment of NaBeF, with metallic beryllium « « « « o « & Reduction of dissolved chromous fluoride by hydrogen. . . Preparation of various fluorides . . . . « ¢« ¢« ¢« « o« « & CORROSTION BRESEARCH . . & v v v v ¢ v ¢ ¢ o o s o o s s o o » Fluoride Corrosion of Inconel in Static and Seesaw Tests ., ., Effect of oxide layer + o+ ¢ o 4 v o o « s o « s s s s o Removal of oxide laver . .+ o ¢ ¢ ¢ 4 o o o o« s o o & o Effect of oxide additives . + ¢« ¢« ¢ + 4 « o o o o o o o o Effect of small zirconium hydride additions . + « « « « & Effect of exposure time . . . . . e e 8 s e 4 s s a4 e Corrosion by fluorides with high UF concentrations . . . Corrosion of high-purity Inconel . . 4 & o o« &+ o o« « o Effect of temperature e e a4 e s e e e s e s e e e e s Fluoride Corrosion of Incomel in Rotating Test . . « ¢ . . & Fluoride Corrosion of Inconel in Thermal Convection Loops . . Zirconium hydride additive . ¢ v 4 & o 4 v « &« o s+ & o Effect of type 316 stainless steel insert in Inconel loop Effect of temperature on Inconel corrosion . . « . . , . Carbon insert in Inconel 1oop ¢« « v ¢ o o o o o o + o s Effect of exposure time o+ « o o o o o o o o « o s s & o » Fluoride Corrosion of Stainless Steels of Varying Purity . . Liquid Metal Corrosion of Structural Metals . « o« & o & + o & Botating tests with sodium . . « ¢« « & v &+ o o ¢ o o o = Static tests with lithiom « « 4 4 ¢ ¢ ¢ « ¢« o ¢« v 2 o o s Dynamic tests with liquid lead in convection loops . . Fundamental Corrosion Research .« « « . ¢ ¢ ¢ ¢« & o o o 4 o Oxidizing power of hydroxide corrosion products . . . . . Equilibrium pressure of hydrogen over sodium hydroxide-metal SYStemS * 2 e s 3 " 2 . e 8 a4 a2 a4 T 8 s s 9 e 4 s e w Chemical equilibria in fused salts . &« ¢« &+ o ¢« ¢« & o « & 61 61 61 61 62 62 62 65 66 69 69 70 70 70 71 72 72 73 73 73 74 75 76 16 16 76 7 17 77 78 79 80 80 80 80 80 83 83 84 85 1X 7. METALLURGY AND CEBRAMICS + 4 ¢« o ¢ ¢ o o o o o s s o & o s o o s o & 87 Welding and Brazing Research .« ¢« ¢ & ¢ o ¢« ¢ ¢ o ¢ o 3 o o o« o s o s 88 Low-melting-point Nicrobraz . . . ¢ « « o ¢ ¢« ¢ o o ¢ o & o o o & 88 Brazing of radiator fins with IMNB . . . ¢ . ¢ « « ¢+ &+ ¢ ¢ o & 88 Nickel-chromium-phosphorous brazing alloys . . ¢« ¢ ¢« « ¢ & « & & 90 Brazing of radiator assemblies with Nicrobraz . « . « « ¢« « « + & 94 Brazing of radiator assemblies with G-E alloy No. 62 . . . « .+ . 95 High Conductivity Metals for Radiator Fins . « ¢« ¢« & ¢« & ¢ « ¢ 4 o+ & 95 Vapor plated chromium and nickel on copper . + « ¢ ¢« « ¢ ¢ « « & 96 Chr()miu!‘l‘l‘plated copper . . . % e o+ & 4 s o s = . . . 0+ v s s w 96 In(:()nel Clad Copper . . e ¢ e e . . * . . . . . » . . . . . « o 96 Type 310 stainless steel-clad copper .« « « « « ¢ « ¢ ¢ « o o o« » 96 Type 446 stainless steel-clad copper .« ¢ ¢« « 4 ¢ ¢« o o & o s o = 96 Copper clad with copper-aluminum alloy . . ¢ + o « « & o ¢ « &« & 96 Mechanical Properties of Inconel . ¢ + o o & ¢ ¢ ¢« o ¢ ¢« o & o & o & 97 Creep and stress-rupture tests . v 4 o 2 o s o s & » s & o s o 97 Tube-burst tests .« .+ ¢ v o &« & 4 o & o o s« 6 s e s s e s s v 97 Fabrication of Pump Seals . o ¢ v ¢« v ¢« v v ¢ ¢ ¢ o o o o o o o s s 98 Hot-pressed pump seals « ¢ ¢ & ¢ ¢ ¢ ¢« & ¢« ¢ ¢ o s 4 4 0 e e o8 Vitreous Seals o ¢ ¢ 4 s 4 4 v 4 s e s v s s e v e s e s s e s 98 Tubular Fuel Elements + o ¢ ¢« v o « o 4 o ¢ & o ¢ o o o s o o o s o 99 Inflammability of Sodium Alloys . ¢ ¢ v ¢ ¢« « & o & « o o« « o« « « « » 100 8. HEAT TRANSFER AND PHYSICAL PROPERTIES . . ¢ ¢ « ¢« ¢ s « « o« s o « « » 101 Enthalpy and Heat Capacity of Halides . . . « ¢« ¢« ¢ v ¢ ¢« « ¢« ¢ « « « 101 Viscosity of Fluorides .+ & o ¢ ¢ v v « o o« o o s o o s s o« o o « « » 103 Density of Fluorides o v v ¢ v v s 4 ¢ o v o 2 s s o + o o o s + « « 104 Electrical Conductivity of Liquids . + ¢« & ¢ « & ¢ « o & o o o« s « » 104 Thermal Conductivity + o o « o o ¢ & o o o o« » o o s 104 Diatomaceous earth . . ¢ & ¢ o v « ¢ « o s s+ o s+ o o & + » o« +« o« 104 Development of thermal conductivity measuring devices 165 Vapor Pressure of Fluorides . . . . . e v s s s s s e e s s e s s . 106 Forced-Convection Heat Transfer with NaF KF-LiF Eutectic . . . . . « 106 Circulating-Fuel Heat Transfer . . . ¢ ¢ ¢ ¢ v ¢ ¢ ¢ ¢ o & o o 2 « « 107 Bifluid Heat Transfer Experiments +« ¢ + o« 4 ¢ « &« o o o + o + s « » » 109 9. PRADIATION DAMAGE . v v v v 4 v 0 v s v s o v o 0 o o s s o o o s« 111 Irradiation of Fused Materials . . ¢ + v o v ¢« ¢ ¢« ¢« ¢ o o o o o « - 111 Analyses of irradiated fuel . . + . ¢ 4 ¢ ¢ ¢ ¢ ¢ v ¢ o 4 o o .o 111 Examination of irradiated fuel contalners « +« + o« + « o o o « « o« 112 In-Pile Circulating Loops « + & ¢« 4+ ¢ ¢ &« & ¢ o o o « s o s o« ¢ +» « « 113 Sodium-Beryllium Oxide Stability Test + & v ¢ ¢ o o « o « o « « o o« + 113 Creep Under Irradiation . o« + & o & « s o o o 2 s s+ s+ s o« o o o« s « & 115 4] - - - - - - & 10. ANALYTICAL STUDIES OF REACTOR MATERIALS . . . . . Analytical Chexr Postrun inspection of the pump showed the interior surfaces to be i1n good condi- tion, although some wear between impeller hubs and mating surfaces was observed and the shaft surface at the seal was roughened. In order to lower the power con- sumption of the seal and yet maintain low leakage, the pump has heen designed to accommodate a 1/2-in.-long freezing gland backing. re- with gas-pressure Gas-pressure backing had (l)w. B. McDonald et al., ANP Quar. Prog. HRep. Jure 10, 1953, OBNL-1556, p. 1l1. been found to be necessary with the 5 13/16-in.-1long seal previously tested 1f low leakage were to be obtained at low powers. The new, short, sealing gland was mounted on the 2 1/2-in.-dia shaft with a 0.015- in. radial clearance between the shaft and the gland. With the helium chamber for ‘‘backing-up” the seal, it is possible to reduce the pressure stress on the seal from 50 to 5 ps1, or less. The helium chamber 1s sealed from the atmosphere by a rubber O-ring seal. A model FP pump, complete except for the impeller, has been run to test the characteristics of a 1/2-in.-long sealing gland. The results obtained to date indicate that this seal re- quires 1/2 to 3/4 hp at 1700 rpm with 75°F water as the coolant. This is about one~tenth the power required for the 5 13/16-in.-long gland under similar conditions. A stable, pre- dictable leak rate has not yet been determined. Leak rates in excess of 100 c¢m® per day and as low as zero have been observed. The average leak rate for 960 hr of operation with a speed of 1700 rpm and a seal pressure differential of 5 psia is 40 cm®per day. A characteristic of the short seal being tested 1s a tendency for leakage rates to increase once any appreciable leakage occurs. Leak- free operation for periods as long as eight days has ocecurred, but periods of operation with high leak rates have followed. Upon removal of the leakage, another period of leak-free operation usually follows, The power and leakage charac- teristics of the 5 13/16-in.-long seal could be described as analogous to those of a viscous film model, as reported previously, but this has not been possible for the 1/2-in.-long seal running under a 5-psl pressure differential. A re-evaluation of the sealing mechanism, along with an analysis of test results, suggests PERIOD ENDING SEPTEMBER 10, 1953 that wetting or nonwetting of the shaft by seal sodium is an important factor for the short seal. Since wetting is essentially an end effect, it would be considerablyless important for a long seal operating under a 50-psi pressure differential. This is a likely explanation for reasonable success of the viscous film model in predicting performance of the longer seal. Gas-Sealed Sump Pump for the ARE. The gas-sealed sump-type pump described and 1llustrated in the previous report(z) has been fabricated and partially tested., This pump requires a vertical shaft to maintain a liquid- gas interface, and only a gasseal is necessary. In additien, the sump tank for the pump suction bell serves as an expansion and degassing tank for the system fluid. The construction of the pump and test loop was completed in June, and test runs with NaF-ZrF -UF, (50-46-4mole %) were started promptly. The pump has now operated a total of 546 hr, 410 hr at temperatures between 1200 and 1300°F and 136 hr at tempera- tures between 1450 and 1500°F. Pump operation during this testing period was suspended on three occasions. However, inspection of the hardened- tool-steel vs. silver-impregnated- graphite gas seal, after each of the three terminations, revealed pgood sealing surfaces. A dry, hot, shake- down run was terminated when noise indicated that the shaft was rubbing against a stationary surface. The pump was reworked so that the clearance between the labyrinth seal and the shaft was increased from approximately 0.010 in. on the diameter to approxi- mately 0.040 inch. The second termination, high-temperature test with the fluoride fuel, was to permit investigation of power pulses of 200 to 400 watts superimposed on the normal power after a (2) Ibid., p. 11 and Fig. 2.1, p. 12. 21 ANP QUARTERLY PROGRESS REPORT curve. No evidence of metal scoring found; frozen fluorides were found clinging to heat radiation shielding just above the labyrinth. This layer of frozen fluorides was pressing against the shaft i1n brake fashion over an arc of about 120 degrees. The presence of the fluorides in this region was due to accidental overfilling of the surge tank. The third termination for an investigation of bearing housing noise. The noise appeared to be due mostly to the races slipping on the shaft, accompanied by a small amount of wear caused by material suspended in the bearing lubricant. At present, the bearing fits, bothon the shaft and in the bearing housing, are being improved. After reassembly, high- temperature tests will be resumed. The level of fluid in the pump tank was raised and lowered to simu-~ late ARE operation. The pump primed immediately with the impeller sub- merged. With pumping flow established, the liquid level was lowered to 1 1in. below the inlet eye of the impeller (which level also provided 1-in, sub- mergence of the inlet to the suction hell), and satisfactory behavior of the pump was noted, that is, steady flow, steady head, steady power, and no evidence of entrapment of system fluid. The pump was subjected to many stop-and-start tests with the fluoride in the system; 13 of the tests were logged. Most of the tests logged were made to verify a minimum priming level located approximately one-half the way up on the leading edges of the impeller vanes. The pump primed successfully at this minimum priming level for all tests in which the shaft speed was 300 rpm or greater. The slower the speed, the greater was the time required to obtain definite priming, that is, resumption of sub- stantially the head and flow for the test shaft speed. During these tests, the power trace evidence of wa s however, was inner gave 22 nearly complete degassing of system fluid in less than 10 minutes. The excellent performance of these pumps to date has resulted in their selection for use in the Aircraft Reactor Experiment, not only in the fluoride system but also in the sodium (moderator-coolant) system. Combination Packed-¥Frozen Sealed Pump for Fluorides. Three different packed-frozen seals were tested on the ARE-si1ze fluoride pump described previously.(3) The first seal was made upof alternate layers of silicon- bronze wool and graphite powder, the second contained a 1:1 mixture of BeF, and graphite powder fused in place under temperature and pressure, and the third was composed of NaBel, powder in small annular cavities between close-fitting machined rings placed around the shaft. An maintained on inert gas was the frozen end of the seal, and a water-cooling sleeve was added. The first seal operated for a period of 160 hr 1in circulating NaF-ZrF,-UF, (50-46-4 mole %). The fluid temperature was 1250°F, the shaft speed was 1250 rpm, the flow rate was 50 gpm, the developed head was 50 psi, and the motor power was 5.7 kw., Heat removed by the water was 5000 Btu/hr. The estimated power absorbed in the seal was 0.8 hp, while the average leakage rate was 100 g per day. The test was terminated because of excessive leakage. OStartup was very difficult with the second seal, and excesslve heating was required to begin shaft rotation. Gross leakage occurred when the pump loop was filled with the circulating fluid. The third seal arrangement was inoperable be- cause the closely fitting metal rings seized the pump shaft during initial dry runs, The work on beryllium-type seals has been discontinuwed, and the punp (3rpid., p. 15. is now being assembled with a short (1/4-in.-long) frozen-fluoride seal with no packing. Frozen-l.ead-Sealed Pump for Fiuo- rides. An ARE size pump was modified to incorperate the frozen-lead seal described previously.(®? A lead- fluoride interfaceis maintained within the pump, and the seal is on the bottom side of the pump in the denser lead. A test of a pump with this seal was terminated after 120 hr of opera- tion when i1t was found that lead from the seal reservoir had been carried out into the loop. A combination of the mixing actionof the rotating shaft at the lead-fluoride interface and of the high rate of fuel circulation over this region because of the drilled impeller seems to have caused this lead transfer. Operation of the seal appears to have been satisfactory; the power in- put to the seal is low and it is relatively insensitive to speed and pressure changes. Ten stop-start tests that ranged in duratien from 1 min to 1 1/2 hr were made during the run. A subsequent test with a 2 1/2- in.-dia shaft rotating in a pot con- taining both lead and fluorides at 1200°F showed that the fluorides con- tained approximately 5% lead during operation and that the lead immediately settled to the bottom when shaft rotation ceased. The mixing during operation makes this type of seal impractical for sealing a fluoride punmp. Allis~Chalmers Canned-Rotor Pump. The Allis-Chalmers canned-rotor pump described previously(B) was operated with NaK at a maximum temperature of 1085°F. Difficulty was encountered with the impeller binding against the pump housing at temperatures above 1000°F. Also, the coolant passages through the motor windings showed a tendency to plug during prolonged operation. This plugging was caused PERIOD ENDING SEPTEMBER 10, 1953 by deposits of zinc transported from the galvanized pipe in the distilled- water coolant circuit. RBeplacement of the galvanized pipe with copper pipe apparently corrected this condition, Pump performance data were taken at various temperatures. The data 1in Table 2.1, which were taken at 800°F, are typical. Several attempts to operate the pump at temperatures substantially above 1000°F resulted in mechanical binding; therefore the pump was returned to Allis-Chalmers, at their request, for modifications to eliminate the binding. TABLE 2.1. PERFORMANCE DATA ON THE ALLIS-CHALMERS CANNED-ROTOR PUMP HEAD (psi) FLOW (gpm) POWER (kw) 13.3 12.1 0.93 14.3 17.9 0.91 15.3 16.0 0.9 16.1 14.0 0.87 16.9 11.9 0.83 17.4 10.0 0.8 17.8 8.0 0.79 18.1 5.9 0.75 18.5 3.8 0.73 19.1 1.5 0.71 ROTARY -SHAFT AND VALVE-STEM SEALS FOR FLUORIDES J. Cisar W, B, McDonald W. R. Huntley G. Petersen L. A. Mann Vv, C. Tunnell B. N. Mason P. G. Smith D. R. Ward ANF Division W. K. Stair, Consultant Spiral-Grooved Graphite-Packed Shaft Seals. The previously mentioned investigation(*) of the nonwetting of graphite by molten fluorides was con- tinued. A test was conducted with a (4)W. B. McDonald et al., ANP Quar. Prog. Rep. June 10, 19532, OBNL-1556, p. 17. 23 ANP QUARTERLY PROGRESS REPORT 1 3/16-1n.-dia vertical shaft that rotated clockwise and had a left-hand V-thread groove machined on the shaft in the seal area, The powdered, artificial-graphite packing was com- pressed by a gland in the packing area, and the spiral groove further compressed the packing at the fluoride graphite packing interface. There was no fluoride leakage, stops and starts could be made without the addition of heat, and there were no signs that freezing was occurring within the seal region after rotation was stopped. After the seal had operated for 158 hr, the pressure to the gland for compressing the packing was in- advertently left off overnight, and the seal began to leak and to act as a frozen seal. Operation continued for another 675 hr before termination, with the seal operated as a frozen seal; the leakage rate was low. This test was duplicated and i1s still run- ning with 700 hours. ma de zero leakage after over Stops and starts have been without the characteristic freezing that is associated with frozen seals. The fuel temperature 1s about 1250 °F and the seal temperature gradient 1s from 500 to 900°F, with power to the seal reasonably constant at less than 150 watts. Annular-Grooved Shaft Seal!. Another test was made with the annular-grooved shaft seal reported previously.(*) The packing material used was bronze wool and a mixture of Asbury graphite and MoS,. Only the mixture of graphite and MoS, was packed in the grooved region of the seal., The test was operated for a period of 106 hr, during which time the leakage rate was higher than in previous tests. The test was terminated becaunse of leakage of molten fuel which resulted from overheating of the seal by friction in the packing. V-Ring Seal. The V-ring seal described previously(®’ was tested with )rpid., p. 19 and Fig. 2.5, p. 20. 24 NaF-ZrF, -UF, (50-46-4 mole %) at about 1200°F and 5-psi pressure. A leakage rate of about 0.5 g/hr occurred for 100 hr of operation; the power require- The coldest above 1050°F ment of the seal was low,. part of the seal during this period, and therefore there was no freezing. The test was terminated for inspection after 100 hr of operation, and no damage was ap- parent. During disassembly, the seal region was heated to remove a thermo- couple, and some oxidation apparently was occurred. As a result, in retesting the same seal there was gross leakage. The same seal arrangement was set up for a test of a 2 1/2-in.-d1ia horizontal shaft. Since all the heat 1s supplied external to the seal region and shaft, the seal clearances temperature sensiltive. tivity 1in this test was much more severe than that 1n the previous test for which a 1 3/16-in.-di1a shaft was used. arc The sensiti- There were periods of apparent zero leakage and low power require- ments, but slight temperature changes had cumulative effects; that 1is, a temperature increase caused greater friction and resulted in an even greater temperature 1ncrease and binding, and, conversely, a temperature drop i1ncreased the clearance in the annu lus and resulted in leakage. The total operating time was 114 hr with the whole seal region abave 1050°F, This seal arrangement will be retested with a heater the shaft to provide more stable temperature con- trol. inside Graphite -BeF, Packing, A seal con- sisting of a mixture of artificial graphite and 20% BeF, with a bronze- wool retainer was tested with a 1 3/16-in.-dia vertical shaft. Since BeF, is glass-like, it should act as a high-temperature lubricant, a binder for the packing material. heated, and com- pressed 1n place, then heated again and compressed to permit the melted as well as The material was mixed, BeF, to fill the voids in the graphite. The seal 1is being tested with NaF- Zr¥F, -UF, (50-46-4 mole %) at about 1200°F and 10-psi pressure, The seal temperature gradient is 500 to 1000°F, and the power requirement is lower than that with straight graphite at shaft speeds to 2400 rpm. The total operating to date, 1s over 1300 hr, and the seal leakage rate wasa little over 4 g per day for the first 1000 hours. time, Bronze-Wool and MoS, -Packed Frozen Seal. The testof the seal with bronze- wool and MoS, packing on a 1 3/16-in.- dia shaft was reported previously.(%) The seal was tested with NaF-ZrF, -UF, (50-46-4 mole %), and 1t continued to operate smoothly until the test was terminated at the end of 1652 hr to make the equipment available for other tests. The average leakage rate of the fuel from the seal was less than 2.5 em® per day. The same seal was then set up with a packing material of bronze wool and Asbury graphite. An attempt was made to pack the seal so that the packing would be a uniform mixture of the two materials. The test was operated for a period of 338 hr, and the results were much the same as those in the above test, but the leakage rate was slightly higher. This test was termi- nated to release the equipment for other seal tests. Bronze-Wool, Graphite, and MOSz- Packed Freozen Seal. A packed-frozen seal of bronze wool with graphite and MoS, as a lubricant has been tested on a 2 1/2-in.-dia shaft. The seal was operated for a period of 76.5 hr, and the test was terminated because of excessive leakage - about 8 cm®/hr at the conclusion of the test., The test was made with the shaft rotating at 1150 rpm to seal NaF-ZrF,-UF, against a pressure of 5 psi at 1100°F, (6) 1 pid., p. 18. PERIOD ENDING SEPTEMBER 10, 1953 Another test, similar to is being made with a cooling coil added to the seal and packing of 1/4- in., layers of bronze wool sandwiched between 1/4~in., layers of powdered Asbury graphite. This test was first operated for 242 hr with the shaft in a vertical position, then with the shaft in a horizontal position. This change from vertical to horizontal operation was made without disturbing the seal. the above, There 1is no apparent difference in operation in these two positions. While the shaft was operated in the vertical position, the average leakage rate of fuel from the seal wasas great as 2.3 cm® per day, but for most of the time, the leakage was less than 1 cm® per day. The test was made at shaft speeds from 800 to 1400 rpm to seal fuel against a pressure of 10 psi at 1175°F, The average leakage rateof the fuel from the seal in the horizontal test has been about 7.5 c¢m® per day, but the test operated for a considerable time with leakage rates of less than 1 e¢m?® per day. To date, the seal has operated for 1000 hr in the horizontal position. There have been power fluctuations in both positions of as much as 500 watts, but, at times, the power requirement was constant for periods of up to 12 hours, Copper-Braid and NMoS,-Packed Seal Test. A packed seal employing copper braid lubricated with MoS, has been tested with NaF-ZrF, -UF, (50-46-4 mole %) under the following conditions: fuel temperature, 1250°F; shaft speed, 2000 to 2800 rpm; shaft diameter, 1 3/16 in.; shaft material, Stellite No. 6 (coated); L/D ratio, 4.2; power to seal, 0.1 to 0.35 kw. This test was terminated after 1000 hr of operation. FPower surges to the seal became large during the later stages of the test. Postrun examination showed severe weldingor galling of the copper to the Stellite shaft surface, ANP QUARTERLY PROGRESS REPORT and, 1in addition, the usual scoring of the seal region was quite severe. Frozen-Lead Seal. The second lead seal test has operated for 2100 hours. The equipment for this test differed from that for the i1nitial lead seal test{7? in that a relatively short (approximately 1/2 in.) frozen seal was obtained by providing water cooling, and an inert atmosphere was maintained over the open end of the seal. The Jead used in this test was 99.49% pure, and 1t was hydrogenated before 1t was loaded into the equipment. The shaft used for this test was 1 3/16 in. 1in diameter, the shaft material was type 316 stainless steel, the shaft speed was 1000 to 4000 rpm, and the power to the seal was 0,04 to 0,20 kw, The seal operated reliably over the available speed range, 1000 to 4000 rpm. input to the seal varied with the speed of the shaft from 0.04 to 0.20 kw. Tests of power input to the pressure showed no effect over the pressure range of 0 Power seal vs. M ypia., p. 19. TABLE 2. 2. to 26 psi. Approximately 40 stop-start tests were made during this test, and there were no failures of the frozen seal (cf., “Frozen-l.ead-Sealed Pump for Fluorides,” above). Vitreous Seals. Seven tests of the use of a vitreous substance, such as BeF,, in the packing gland of a shaft seal were described in the previons report.‘’) In addition, two other seals have been tested. A summary of these two seal tests 1is given in Table 2.2, For the first vitreous seal test made during this quarter (number 8 in the series) the seal region around the 1 3/16-in.-dia shaft was divided into fifteen 3/16-by 3/16-in. compartments, and each compartment was packed with granular NaBeF, which had passed a The test was with the shaft 1n the vertical but, after a short period, the test rig was rotated to place the shaft in a horizontal position. The shaft seal operated satisfactorily for Tyler 8-mesh screen. started position, TESTS OF VITREOUS SEALS ON ROTATING SHAFTS OPERATED IN NaF-Zrf, -UF, (50-46-4 mole %) TEST NUMBER 8= g+ Seal Packing Chamber Over-all length, in. 5 4 Length of packing, in. 4 25 Number of compartments 15 5 Thickness of Facking Annulus, in. % s Diameter of Shaft, in. 12 2% Speed of Shaft, rpm 1350 to 4400 640 to 800 Helium Pressure, psi1 5 5 Duration of Operation, hr 6E79%* S5T** *Composition of seal packing: and 9% AlF,. Test No. 8, 100% NaBeFS; Test No, 9, 50% BeFQ, 25% KF, 16% Mng, **Terminated because of system faitlure, not at seal. 26 679 hr, and the test was terminated because of leakage at a flange joint that was independent of the seal. The shaft surface speed was held at approxi- mately 930 fpm during the test to simulate operation of the ABE pump shaft. A seal leakage of not more than 4 cm® of dry powder per day occurred from the fifthto the fifteenth day of the test. During the remainder of the test the leakage was zero. The friction power loss at the seal was less than 200 watts during most of the test; there were, however, intermittent power surges of greater magnitude. The seal temperature ranged from 100°F at the cold end to 1000°F at the hot end. The second vitreous seal tested (number 9 of the series) employed a 2 1/2-in.-dia shaft with the seal annulus that contained five seal com- partments filled with a vitreous precast ring of a mixture of fluorides (by weight: BeF,, 50%; KF, 25%; MgF,, 16%; AlF,, 9%). This mixture of fluorides was chosen because of its high viscosity over a wide temperature range. The mixture was hydrofluorinated to remove all moisture and oxides and then mixed with an equal weight of BeF, and cast (under an argon blanket) into the packing rings with which the seal was packed. The seal operated successfully for 46 hr with zero leakage, but the test was terminated when a heater failed. Examination revealed that the heater failed when 1t was contacted by molten fluorides which leaked from a flange joint that was not part of the seal region. During most of the successful run, the friction power dissipated at the seal was about 1000 watts. The seal temperature ranged from 300°F at the cold end to 1000°F at the hot end. Graphitar-Graphitar Shaft Seal. Graphite has been used as one of the sealing elements in much of the work with packed and ring types of shaft seals for fluorides. One of the property of graphite 1in PERIOD ENDING SEPTEMBER 10, 1953 characteristics of graphite which has prompted 1ts continued use 1s 1its apparent reslstance to wetting by the fluoride fuel. However, all the packed and ring types of seals tested have had one of the sealing elements, usually the shaft, made of Stellite. This provided a wetted surface for the passage of the fuel. Thus the real significance of the nonwetting charac- teristic of the graphite on seal performance was masked by other variables, In an effort to ascertain the importance of the nonwetting property of graphite, a 1 3/l6-in.~dia shaft seal was designed which employed solid Graphitar 14 shaft bushings running against similar bushings in the seal housing. The seal housing was provided with heaters and a cooling medium. 1t was then possible, by varying the temperature difference between the shaft and housing, to control the running clearance of the seal to a small degree. The seal was operated for 195 1/2 hr under pressure differentials of up to 10 psi with the seal temperature above the fuel melting point. The power required was very low and steady, and the maximum leakage rate observed was 4 g per day. During two successful stop-start tests that were made 1t was found that the seal temperature should be maintained above the fuel melting point. Disassembly and i1nspection revealed that (1) the Graphitar 14 wear was slight (small amount of cracking noted), (2) the clearance between the Graphitar 14 and the adjacent stationary metal surfacewas filled with fluoride, and smaller amounts of fluoride were retained between the mating Graphitar rings, and (3) mountingof the Graphitar 14 had been accomplished without excessive crackingat high temperature. In summary, these results confirm the usefulness of the nonwetting seal design., 27 ANP QUARTERLY PROGRESS REPORT Packing-Penetration Tests, Several additional(®) packing-penetration tests were made this gquarter; the results are summarized in Table 2.3. Test 16, which enmployed a mixture of graphite and BeF, and was reported previously‘®? as ““still running,” was terminated after 736 hr of operation with no leakage. Upon examination, there appeared to be no penetration of the packing by the fuel and no oxidation of the graphite at the lower opening where the packing material was exposed to the atmosphere. Five of the tests reported this quarter contained artificial graphite. One of the most promising graphites vyet tested i1is Asbury graphite 805, The test with this material ran for 1660 hr and terminated with no leakage. Examination showed that there wa s was no penetration by the fuel and very little oxidation of the graphite. Botating shaft seals packed with this material show low power dissipation from friction. It 1s believed that a small amount of BaF, added to the graphite will reduce the friction. A penetration test was first run with only BalF, as the packing material. This packing leaked immediately, and the test showed that the BaF, was soluble in the fuel. A subsequent test with a mixture of 10% BaF, and 90% Asbury graphite B05 has been running for 550 hr with no leakage. HIGH-TEMPERATURE BEARING DEVELOPMENT PROGRAM W, C, Tunnell, ANP Division W, K. Stair, Consultant In future work contemplated by the ANP Division, there will be need for a journal bearing that is suitable for operation i1n a fluoride fuel and that can be lubricated by the fuel. The expected applications will impose low radial bearing loads, but the operating temperature will be in the range of 1100 to 1500°F, Except for the problem of material selection, the design of the bearing 1s expected to be more or less routine. The phase of the program now being started involves a study of (8)1pid., p. 22 and Table 2.4, p. 23. the compatibility of various material TABLE 2.3. SUMMARY OF PACKING-PENETRATION TESTS PERFORMED AT 1506°F AND 30-psi PRESSURE DURING THIS QUARTER 1&iT PACKING MATERTAL DURATION JF REMARK S 16 80% graphite from Y-12 Carbon Shop, 20% BeF, 736 Terminated with no leakage 18 National Carbon Company graphite BB-4 1/2 19 Asbury graphite 805 1660 Terminated with no leakage 20 National Carbon Company brush-type graphite 1/2 21 Norton Company graphite BN 1275 Terminated with no leakage 22 Carborundum Company graphite BN 140 23 Bal, 1/2 24 Zx0, 1/2 25 90% Asbury graphite 805, 10% BakF, Had not leaked at 500 hr, test continuing 26 Asbury graphite 805 and Fel-Pro C-5 high- Had not leaked at 250 br, teinperature lubricant test continuing 28 combinations and their resistance to attack by NaF-ZrF,-UF, (50-46-4 mole %). Compatibility in this application implies the ability of bearing materials to move relative to each other while in contact under load in the fluoride without excessive wear or corrosion and without galling or seizing. A com- patibility tester is being fabricated, and 1t will be used to screen from a large group of material combinations those materials which have good com- patibility and merit further exami- nation. This phase of the program will probably include tests on the mechanical face seal, which equally necessitates knowledge of compati- bility., INSTRUMENTATION G. Petersen P. W. Taylor ANP Division Fluoride Leak Detection. A test which simulated a small leak from an ARE fluoride fuel pipe i1into the helium annulus surrounding it was conducted at 1250°F with NaF-ZrF -UF, (50-46-4 mole %). With a gas flow of 4 cfh (80% He, 20% air) passing through the annulus and out into a sight feed bubbler containing a neutral phenol- red indicator solution, approximately 0.5 in.> of fluoride was expelled into the annulus. The resulting hydrogen fluoride rapidly changed the color of the indicateor solution to yellow. Another test was run with dry air instead of with helium in the annulus, and the total pas flow (BO% dry air, 20% air) was increased to 50 cfh. Only a fraction of the gas passing through the annulus section was exhausted through the bubbler. When the fuel expelled into the annulus, the indicator solution turned yellow, as in the previous test. The tests indicate that this leak detection system may be used in the ABE fuel system. In both cases, the line leading to the bubbler was 1/4- was PERIOD ENDING SEPTEMBER 10, 1953 in. copper tubing approximately 50 ft long. Sodium Leak Detection, The apparatus for detecting sodium leaks involves the decrease 1n electrical resistance of sodium-contaminated insulation. As shown in Fig. 2.1, an ohmmeter and a filament lamp were connected between the container wall and the 0.,020-1in. stainless steel sheet stock with a 6-volt battery. The resistance of the circuit was then tabulated against temperature as heat was added to the sodium. When the sodium temperature was increased to 850°F, the resistance of the circuit increased. At this point, the ohmmeter showed a drop in resistance because sodium vapor began to contaminate the spun glass therefore to decrease its resistance. Atv 1200°F the apparatus was tilted and liquid sodium was forced through the leak. The lamp filament was energized immediately and the re- and electrical sistance reading dropped to zero. This test was performed with the sodium leaking througha 0.125-in.-dia opening and a 0.028-2n.,-dia opening. In each case, the same results were obtained. UNCLASSIFIED CWG 21150 LAMP FILAMENT T\Mfi""“" S-volt o BATTERY [ ! -t OHMMETER € __HELIUM ATMOSPHERE TERMINAL ‘~—~TERM|NAL 0.020-in. STAINLESS STEEL SHEET | ~SODIUM LIQUID LEVEL INSULATION o, CONTAINER - WALL TUBULAR HEATER [/ SPUN GLASS -—X Fig. 2.1. System. SodivmLeak-Detection Test 29 ANP QUARTERLY PROGRESS REPORT This method would work well for a small system with perhaps 10 to 20 welds, but for a larger system, such as the ARE, the wiring and instrumenta- tion would be complicated and diffi- cult, Flow Measurement. Venturi meters used in conjunctionwith Mcore Nullmatic pressure transmitters have been demon- strated as being reliable for measuring flows of high-temperature fluids. The outputs from the two Moore trans- mitters that measure the venturi AP are fed inte a differential pressure transmitter. The output gage of this transmitter 1s calibrated to read flow directly in gpm. Such a system has been used successfully for about 550 hr on a pump loop to measure flows of up to 67 gpm. The pressure trans- mitters, rated 0 to 100 psi, are attached to the underside of the venturi section and filled completely with liguid to avoid a liquid-gas interface. The pressure-sensitive element 1n the transmitters is a 3-ply Inconel bellows. The temperature of the transmitters 1s maintained at 1100°F maximum to avoid serious cor- 30 rosion. The liguid in the transmitters on this loop has been frozen and re- melted with no apparent damage to the and the instruments have not required recalibration upon re- starting. The null-balance transmitters are instruments, accurate as long as the zero setting remains fixed. When the transmitters are initially heated to 1100°F, there 1s a substantial zero shift that can easily be adjusted. Subsequent zero shifts are probably caused by sudden decreases in applied pressure. Because of the balancing pressure on the inside of the bellows and the reduced spring rate at the high temperature, the bellows i1s often stretched beyond its elastic limit. This permanent set causes a zero shift., Although 1its occurrence 1s unpredictable, such a zero shift can readily be corrected when the applied pressure 1s brought to zero. When the temperature 1s main tained constant at 1100°F and sudden pressure changes are elimi- nated, the Moore pressure transmitters can be expected to give accurate and dependable service, PERIOD ENDING SEPTEMBER 10, 1953 3. AIRCRAFT REACTOR DESIGN STUDIES A, P. Fraas, Early in 1953, the Air Force re- quested that designs be prepared for 100-, 300-, and600-megawatt reflector- moderated(??® reactors for evaluation. Since the outstanding uncertainty was that of shield weight, to outline a reactor and shield designs to provide a fairly clear picture of the implai- it was decided still wider range of cations of variations in such factors as reactor power density, core diameter, design power output, etc, At the same time, further test data were obtained on the relative activation of several secondary circuit fluids, and the effects of using them instead of NaK were examined, The power plant system outside the reactor shield was also given further attention to produce a more detailed picture of problems in performance and control, SHIELD DESIGNS FOR REFLECTOR-MODERATED REACTORS M. E. LaVerne F. H. Abernathy A, F. Fraas ANP Pivision C. S. RBurtnette USAF R, M. Spencer A fairly complete survey of shield designs to determine the effects of reactor power, core diameter, and division of the shield on shield size and weight, The work was carried out in conjunction with the Shielding Poard, (3) and all details of the calculations were made by using methods recommended by the Board. Each reactor shield was designed on the basis of LLid Tank Faci1lity test was made (I)A.[h Fraas et el., ANP Quar. Prog. Rep. Mar. 10, 1953, ORNL-1515, p. 41-84, (2)A. P. Fraas, ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p. 24-25, ) Report of the 1953 Summer Shiclding Session, ORNL-1575 (to be published). ANP Division results(?) to give one-eighth of the radiation dose in neutrons and seven- eighths in gammas, Crew shield weights were calculated for two repre- sentative crew compartment configu- rations; the right-circular cylinder in one configuration was 8 ft in diameter and 12 ft long, and was 5 ft in diameter and 10 ft long in the other configuration. Designs were prepared for attenuation factors of 1¢, 100, 1,000, and 10,000, re- spectively, Total reactor and crew shield weights can be readily obtained by combining the proper reactor and shield combination with the proper crew compartment., This was done to give a crew dose rate of 1,0 rem/hr. Tables 3.1 to 3.4 the stgnificant weight and dimensional data calculated, Typical curves were prepared to show the effects of major parameters on shield weight, A careful scrutiny of these leads to the following conclusions: , 1. A divided shield makes possible a large saving in total shield weight as the full-power radiation dose 50 ft from the center of the reactor 1is increased from 1 to 100 r/hr, Further division of the little, 1f anvy, except for small summarize shield results 1in reduction in weight, large-diameter cores and crew compartments. It does permit substantial reductions 1in shield diameters, See Figs. 3.1 and 3.2, 2. The shield weight for a given however. power output is not very sensitive to reactor core diameter for cores less than 30 in. in diameter. The shield welght increases rapidly for larger cores, See Figs. 3.1 and 3.3. (4)p, N. Watson, A. P. Frass, M. F. LaVerne, and F. H. Abernathy, Lid Tank Shielding Tests of the Reflector-Moderated Reactor, QORNL-1615 {to be published). flu! '!1’51 ey T | | ,,,,,,,,,,,,,,,,, I , el s L _ TN \\q\ I i 1 1 L _____Qalwafrs o ‘ 1 d 2 \\ . | e L e 22.7-in.-dia ‘CORE- y = 80 <~ 1‘ ‘ I | ‘ e e -3 2.00 megawatts ! I i T "3 ‘ ] - \H\T\NNM L T T T e \ o i ! -0-in.-dia CORE; 200 megawatts = \ i 18.0 -in-dia CORE; 100 megawatts - | ! : 1 ; dhmmcromacn s e e el 2 ~— | | | N R A I—O— \ V fi‘_ e | 13.0~‘in,-dlfl LCOR‘E; b!O r'neigmnfi - | '““‘-f-m~nl_m 8.0 -in-dia CORE; 25 megawatts ] | | | 40 - — - e I O e I I i \ . | ; ! . Lo | i o ‘ B SN | | o i i e ,jjm, i _ - %Tfi,_i ! | i ! | o \ NI i . | | ! , | L i * 0 [_ 11 J ,,,,,,,,,,,, . N‘, ] e 1 2 5 10 2 5 100 2 5 1000 2 DESIGN DOSE 50ft FRCM REACTOR SHIELD (r/hr) Fig. 3.1. Effect of Design Dose from Reactor Shield om Total Reactor and Shield Weight for 1 r/brin Crew Compaviment 8 ft in Diameter and 12 ft Long. 3. to reactor power output. A reduction in the reactor power makes possible a Shield weight is quite sensitive substantial saving in weight. See Figs, 3.1 and 3.3, 4., A reduction of the dose in the crew compartment by a factor of 10 imposes a weight penalty of from 5,000 to 15,000 1b, depending primarily on crew compartment size., See Figs, 3.1 and 3.4. 5. sidered, For most of the designs con- the dose from the activated 32 secondary coolant can be kept quite small. Although activation of the secondary coolant circuit (cf., ‘“Activation of Bubidium, Potassium, and Sodtum,” this section) was not included in the calculations, 1in no case will i1t affect the designs for which the dose at 50 ft is 10 r/hr or Its only effect would be to the weight of the crew compartments with very nearly unit shields for reactor autputs of greater than 200 megawatts, See Fig. 3.5. greater, increase TABLE 3. 1. FROM CENTER OF REACTOR WITH DOSE SEVEN-EIGHTHS GANMAS AND ONE-EIGHTH NEUTRONS REACTOR-INTERVEDIATE HEAT EXCHANGER-SEIELDP WEIGHT FOR A DOSE OF 1 rem/hr 50 ft LEAD THICKNESS {in.} REACTOR AND . LEAD LEAD WATER ‘ WEIGHT TOTAL BREACTOR CORE Lid Tank Dose . . WATER PRESSURE BASIC SPHERE WEIGHT OF POWER INSIDE OUTSIDE OUTSIDE LEAD WEIGHT ) i OF . AND REACTOR DIAMETER Z {em) ————————— 1 {Power} For Core Gammas For Core Plus WEIGHT SHELL WEIGHT STRUCTURE . - (ifl- (megawatts} (Full-Scale DDSQ) (Assuming 0.8 Heat Exchanger DI(A‘MET;EB DI(A:fIEI];ER DI(AI.'L:_IETEQ (lb) (1b) ASSEMBLY (1b) P(AIT};CF (lb} bHIEL(leW)}:.IGHT r/hr} Gammas An. ' ' WEIGHT (1b) 14,3 25 133.5 1.55 6.25 65.18 44 .4 56.8 119.3 20, 500 28,650 5,088 54,2490 1,200 1,080 56,500 50 138.0 1.47 7.10 7.03 46.3 60.4 122.9 26,000 30,920 5,852 62,770 1,700 1, 260 65,700 100 145.0 1,36 8.10 §.09 49,4 65.6 128.5 34,000 34,770 7,398 76,770 2,500 1,540 80, 800 18 25 130.0 2,16 6.00 5.94 47.8 59.8 120.2 22,100 28,810 6,100 57,000 1,300 1,140 56,400 50 135.5 2,04 6.70 6.59 49.5 62,7 124.6 26,600 31,880 7,000 65, 480 1,800 1,319 68,600 100 141.5 1.990 7.70 7.65 52.3 67.6 129.2 35,800 34,920 8, 300 79,020 2,500 1,580 83, 100 200 149.0 1.75 8.80 8.82 57.0 74.6 135.3 49, 200 . 38,950 11,400 99, 550 3,750 1,99¢ 165, 300 22.7 50 133.0 2,68 6.55 6.48 53.7 66,7 127.3 30, 500 33,370 8,480 72,450 1,850 1,450 75,700 100 139.0 2.50 7.33 7.26 56.90 70.5 132.1 37,800 36,920 9,900 84,620 2,650 1,690 88,900 200 146.0 2.32 8.33 §.31 60,2 76.8 137.7 50,900 40,760 13,600 104, 660 4,150 2,100 110,900 400 153.5 2,15 9.70 9.71 67.0 86.4 143.7 73,800 43, 860 18, 500 136, 160 6,400 2,720 145, 300 28.5 200 143.5 3.06 g.00 7.97 64.4 80.3 141.5 54, 300 43,750 15, 300 113,350 4,700 2,260 120, 300 400 151.0 2.84 9.15 9.15 70.4 8g.17 147.2 76,000 47,060 20, 500 143,560 7,450 2,870 153,900 800 158.0 2,65 10.6 10.66 79.7 101,90 152.8 114,000 48,21¢ 31, 360 163,570 11,2090 3,870 208, 600 36 200 141.0 4,20 7.55 7.51 70.4 85.4 147.0 59,500 48,520 18,700 128,720 5,300 2,570 136, 600 400 148.0 3.90 8.65 8.65 74,2 91.5 152.5 76,600 52,830 25,200 154,630 8, 800 3,060 166, 500 800 155.90 3.64 10.05 10.10 84.0 104.2 158.0 115,500 53, 445 34,400 203, 440 19,000 4,070 226,500 45.3 200 138.5 5.60 7.15 7.11 78. 4 92.6 154,3 67,000 04,720 24, 800 146,520 6,300 2,930 155,700 400 146.0 3.20 8.10 8.08 82.4 98.6 160.2 85,500 59,910 246,400 174,810 12,000 3,500 199, 300 800 152,90 4,90 9.50 g.53 89.7 108.8 164.9 121,100 60,730 39,900 221,730 20,000 4,440 246, 200 33 34 TABLE 3. 2. REACTOR- INTERMEDIATE HEAT EXCHANGER-SHIELD WEIGHT FOR A DOSE OF 10 rem/hr 50 ft FROM CENTER OF REACTOR WITH DOSE SEVEN-EIGHTHS GAMMAS AND ONE-EIGHTH NEUTRONS LEAD THICKNESS (in.) , REACTOR AND CORE / Lid Tank Dose : LEAD LEAD WATER WATER PRESSURE Basic SPHERE | "EIGHT % wpygyr or | TOTAL REACTOR DIAMETER POWER Z (em) |l———————) (Power) | For Core Gammas | For Core Plus | INSIDE 4 OUISIDE 3 OUTSIDE ~i LEAD NELGRT | we1gpr SHELL WEIGHT patey | STRUCTURE | AYD BEACTOR (in.} (megawatts) \Full-Scale Dose {Assuming 6 Heat Exchanger 1A (J ) (in {1b} ASSEMBLY {1b} (1b) (1b) (1b) r/hr) Gammas LD R s WEIGHT (1b) ’ 14.3 25 111 2.60 4.70 4.58 43.9 53.1 101.9 13, 600 17, 260 5,088 35,948 1,100 719 37,767 50 116 2,42 5.30 5.23 45.8 1 56.3 105.5 17,900 18,471 3,852 42,223 1,600 844 44,667 100 122 2,23 5.90 5.81 48.9 60.5 110.3 22,500 21, 389 7,398 o1, 287 2,100 1,026 54,413 i8.0 25 108.5 3.60 4,50 4.34 47.3 56.0 103.9 14,900 17,828 6,100 38,828 1,250 776 40,854 50 114 3.33 5.1 5.03 49.0 59.1 108.0 19,000 19,894 7,000 45,894 1,700 918 48,500 100 119.6 3.08 5.7 5.65 51.8 63.1 112,90 24,800 21,789 8,300 54, 889 2,375 1,098 58,375 200 125.5 2,86 6.25 6.19 56.5 68.9 117.0 31,600 24,298 11, 400 67,298 3,200 1,346 71,850 22.7 50 112 4.37 4.9 4,78 53.7 63.3 110.7 21,400 20,800 8,480 50,680 1,400 1,014 53, 400 100 117.5 4.05 5. 45 5. 40 55.5 66 .4 113.5 26,200 921,661 9,900 57,761 2,300 1,156 61,250 200 123.3 3.76 6,05 6.02 59.7 T1.7 119.9 33,800 25,781 13,000 72,581 3,500 1,452 77,550 400 129.5 3.50 6.60 6.57 66.5 79.6 124.7 44,800 27,816 18,500 91,116 5,100 1,822 98,020 28.5 200 121.3 4,62 5.80 5.77 63.9 75.4 124.3 36,000 28, 647 15, 300 79,947 4,000 1,598 85, 500 400 127 4,56 6.5 6.46 69.9 82.8 128.5 49,100 29,547 20, 500 99, 147 6,000 1,982 107,080 800 134 4,25 7. 7.13 79,2 g3.5 134.3 69,000 30,883 31, 360 131,243 9,900 2,624 143,740 36.0 200 119 6. 5.55 5,49 69.9 20.9 129.8 44, 300 30,845 18,700 89,845 4,600 1,796 96,250 400 125.3 6. 6,20 6.13 73.7 86.0 135.0 50, 800 34,501 25,200 1190, 5601 7,500 2,210 120,200 8GO 131 5. 6.9 6,87 83.5 87.2 139.2 73,300 34,188 34,400 141,888 11,000 2,840 155,800 45.3 200 117 8.95 5.35 5.23 77.9 8E.4 137.5 47,300 35, 397 24,800 107,497 5,600 2,150 115,250 4060 123.3 8.3 5.55 5.48 81.9 92.9 143.5 53, 800 39,731 29400 123,931 9,000 2,480 135, 400 800 129 7.8 6.60 6.54 89,2 102.3 146.9 76, 600 40,573 39,900 157,073 16, 000 3,140 176,200 TABLE 3.3. REACTOR-INTERMEDIATE HEAT EXCHANGER-SHIELD WEIGHT FOR A DOSE OF 100 rem/hr 50 ft FROM CENTER OF REACTOR WITH DOSE SEVEN-EIGHTHS GAMMAS AND ONE-EIGHTH NEUTRONS CORE Lid Tank Dose LEAD THICKNESS (in.) LEAD LEAD WATER REACTOR AND WELGHT TOTAL REACTOR POWER INSIDE | OUTSIDE | OUTSIDE - WATER PRESSURE BASIC SPHERE | ' WEIGHT OF DIAMETER | (megawatts) Z (em) (%;II?;::}:‘B::?) (Power) F%;;?E:hfaflflf“ }il; %;21:;222 DIAMETER | DIAMETER | DIAWETER | ““A0 MEIGHT 4 yprent SHELL WEIGHT o I stRrcTuRE | AND HEACTOR in. g - {in.) {in.) {in.) {ib} ASSEMBLY (ib} (ib} r/hkr) Gammas ippeigl {1k} {1b) WEIGHT {(ib) 14.3 25 95 2.64 2.95 2.74 44.4 49.9 89.0 7,800 10,984 5,088 23,870 1,000 476 25, 300 50 100 2.44 3.55 3.40 46.3 53.1 93.1 10,800 12,439 5,852 29,080 1, 300 580 31,000 100 105.5 2.24 4.15 4,900 49. 4 57.4 97.5 14,800 13,943 7,398 36, 140 1,750 720 38,600 18.0 25 93 3.6 2.7 9,38 47.8 52.6 . 91.2 8,180 11,599 6, 160 925,880 1, 200 516 27,600 50 97.5 3.35 3.33 3,19 49.5 55.9 94.8 11,500 12, 860 7,000 31, 360 1, 600 626 33,600 100 103 3,10 3.95 3,83 52.3 60.0 99,1 15,800 14,323 8,300 38, 420 2,100 768 41,300 200 108.5 2.86 4.50 4.38 57.0 65.8 103. 4 21, 600 ¥5,519 11, 400 48, 500 2. 800 970 52,300 92.7 50 96 4,40 3.1 2,91 53.7 59.5 98.3 12,000 13,970 8, 480 34, 450 1,650 688 36, 800 100 101 4.06 3,7 3,59 56.0 63.9 103.0 16, 000 15, 890 9,900 41,790 2,250 836 44,900 200 106.5 3.77 4,25 4,18 60,2 68.5 106.6 22,200 16,814 13,000 52,010 3,000 1,040 56,000 400 112 3.48 4.95 4.83 67.0 76.7 111.0 32, 300 7,341 18, 560 68, 140 4,900 1,362 74, 400 28.5 9200 104 4.90 4.10 4. GO 64.4 72.4 110.4 24, 200 18, 276 15, 300 7,780 3,800 1,156 62,700 409 110 4,55 4.70 4.62 70.4 79.6 115.1 32, 500 19,298 2¢, 500 72,300 5,600 1,446 79, 300 800 116 9.22 5,95 5,20 79,7 91.1 120.0 53,900 18,372 31, 360 103,630 8,000 2,072 113,700 36 200 102.5 6.61 3.90 3.74 70.4 77.9 116.7 26, 600 21,117 18,7060 66,420 4, 000 1,128 71,500 400 108.5 6. 10 4.45 4.32 74.9 82.8 121. 4 34, 206 93,114 95, 200 9,510 5,100 1,650 90, 300 800 114 5,70 5.00 5.00 84.0 94.0 125.9 51,000 22 098 34, 400 107, 430 8,300 2,148 117,900 45.3 200 100.5 8.80 3.65 3. 43 78. 4 85.3 124. 4 33, 300 24,664 24,800 82,760 4,800 1,656 89, 200 400 107 8. 10 4.20 4.07 82,4 90.5 129.7 39, 400 97,956 29, 400 96, 060 7,600 1,920 105,600 800 112 7.61 4.85 4,79 89.7 98. 3 133.7 49,000 97, 240 39,900 116, 140 9,000 2,322 127, 500 TABLE 3.4. REACTOR-INTERMEDIATE HEAT EXCHANGER-SHIELD WEIGHT FOR A DOSE OF 1000 rem/hr 3¢ ft FROM CENTER OF REACTOR WITH DOSE SEVEN-EIGHTHS GAMMAS AND ONE-EIGHTH NEUTRONS LEAD THICKNESS (in.) REACTOR AND . LEAD LEAD WATER WEIGHT TOTAL REACTOR DlggggER POWER Z {cm) -Eif—szf—ffff: {Power) For Core Gammas For Core Plus D%Efié?gR g?f;g%g% 饣§¥¥¥% LEA%li%IGHT £¥gg§} p%fifififi?E BAs#gléggERE P;?;H ggé%?%Ugg SQ?ELSEQEEQET in. (megawatts) Full-Scale Dose {(Asguming 500 Heat Exchanger (in.) (in. ) (in.) {1b) ASSEMBLY (1b) (1b) (1b) (1b) e/hr) Gammas rhe Ha . WEIGHT (1b) 14.3 25 81.3 3.30 1.28 1.12 44,4 46.5 78.4 2,785 7,207 5,088 15, 080 700 300 16, 100 50 85.0 3.08 1,85 1.76 46.3 49.8 81.3 5,190 7,939 3,852 18, 880 1,150 376 20, 400 100 89.0 2.90 2.33 2.29 49.4 54.90 84.5 7,900 8,424 7,398 23,720 1,500 474 25,700 18.0 25 80.1 4,45 1,05 ¢.92 47.8 49.6 81.0 2,741 7,723 6,100 16,560 900 330 17,800 50 83.5 4,23 1.65 1.56 49.5 52.6 '83.7 5,190 8,341 7,000 26,530 1,400 410 22, 300 100 87.0 3.98 2.10 2.12 52.3 56.5 86.5 7,940 8,826 8,300 25,060 1,900 500 27, 500 200 91.5 3.71 2.65 2.68 57.0 62.4 90.0 12, 360 9,193 11, 400 32,950 2,650 858 36,500 22.7 50 82.5 5.46 1.40 1.27 53.7 56.2 87.7 4,840 9,402 8,480 22,720 1, 400 454 24,600 100 85.5 5.16 1.93 1.81 56.0 29.6 90.3 7,720 9,914 9,900 27,530 2,000 550 30, 100 200 90.0 4,82 2.40 2.49 60.2 65.2 93.5 12,610 10,2146 13,000 35,830 2,900 716 39, 400 400 94.5 4.46 3.00 3.08 67.0 73.2 96.9 19,570 9,779 18,500 47,840 4,100 956 52,900 28.5 200 88.5 6.20 2.27 2.30 64.4 69.9 98.1 13,120 11,633 15,300 40,050 3,200 800 44,100 400 83.0 5.80 2.80 2.90 70.4 76.2 102.1 20,000 11,753 20,500 52,250 4,900 1,044 58, 200 800 98.0 5.37 3.30 3.48 79.7 86.7 165.5 30,110 9,881 31,360 71,350 7,200 1,426 80,000 36 200 87.0 8.30 2.05 2.03 70.4 T4.5 104.4 13,850 13,528 18,700 46,080 4,200 920 51,200 400 91.0 7.80 2.65 2.65 74,2 79.5 107.6 20, 100 14,055 25,200 59, 350 5,600 1,186 66,100 800 96.0 7.25 3.13 ' 3.33 84.0 90.7 111,86 32,810 12,172 34,400 79, 380 8,400 1,586 89, 400 45,3 200 85.5 10.68 1.90 1,77 78.4 8l.9 112.5 14,400 16,527 24,800 55,730 4,400 1,114 61, 200 400 89.5 10,16 2.40 2.42 82.4 87.2 115.7 22,150 16,317 29, 400 67,870 6,900 1,356 76, 100 800 93.5 9.70 2.95 3.11 89.17 95.9 1i8.9 34,300 15,046 39,900 89, 250 10, 800 1,784 101,800 F)‘*G 21152 ) < | — B o 120 ] | / | e D00 (]/h' / @0 2 POWER DENSITY IN FUEL=2.8 kw/cm DOSE FROM REACTOR = /8 nautrons and 7/8 gammas 1 el 3 | REACTOR SHIELD ASSEMBLY DIAMETER fin o S E < 2 40 L 0 100 200 3GO 400 500 200 F0O 300 2C0 REACTOR POWER [megowotis) Fig. 3.2. Effects of Power Output and Shield Division on the Diameter of the Reactor Shield Assembly. The diameter given is for the basic sphere. If a higher radiation level at the top of the reactor is to be avoided, an extra ‘'pateh” about 5 in. thick will be required there. COMFARATIVE ANALYSES OF LIQUID METAL COOLANTS A, P. Fraas M. E. ANP Division LaVerne Activation of Rubidium, Potassium, and Sodium. A limiting factor in the degree to which a unit shield may be approached is the gamma dose from the secondary coolant outside the main gamma shield, In an effort to determine the extent to which this gamma dose might be reduced, several tests were run 1n which fluoride salts of three of the alkali metals, sodium, potassium, and rubidium, were 1rradiated and their relative activations determined. These data are reported in detail in a separate repurt.(4) From the basic data, gamma doses for potassium and rubidium relative to sodium were calculated for a variety of irradi- decay times, general, and lead the results indicate that potassium is far atiron times, shield thicknesses. In superior to sodium, the relative gamma dose from potassium being from 1 to 5% of that combination of conditions., These data are shown in Fig, 3.6, For rubidium, on the other hand, the dose relative to sodium depends strongly on the combination of conditions considered; the dose ratio ranges from about 10:1 from sodium for almost any for short irradiation and decay times and no shielding to about 0.002:1 for moderate irradiation and decay times and heavy shielding. No tests were run with Iithium, but previous work indicates that 1ts activity would be of the order of 0.1% of that of sedium. Unfortunately, there 1is, as yet, no suitable metal available to contain Iithium in the temperature range of interest, Effect of Secondary Coolant on Power Plant Weight, The effect of the secondary coolant on the over-all power plant weight is an 1mportant consideration, Although many bases for comparison can be used, it appeared that a good indication would be ob- tained by modifying the 200-megawatt power plant previously described(!) to maintain the same pressure and temperature drops with all coolants considered, that is, sodium, potassium, and natural lithium. The resulting effects of the choice of a secondary fluid are given in Table 3,5 in terms of the weight increase or decrease relative to the original design for use with NaK. The lower specific heat and the lower density of potassium require larger lines, larger tubes in the heat exchanger, greater pumping power, etc., and hence intermediate potassium requires a heavier system. The detailed character of the effects is shown more clearly by the data in Table 3.6, which lists the physical properties and related parameters of the alkali metals considered for operation at a mean temperature ot 700°C. 37 ANP QUARTERLY PROGRESS REPORT DWG. 24453 160 140 e | < > 2 420 |- F I s ) = 2 © 400 | [FR} w) w TURBINE Na-TO-NaK REFLECTOR COOLANT HEAT EXCHANGER, TWO REQUIRED CHECK 4*4"b VALVE f b FROM ~ Na REFLECTOR COOLANT — OTHER ENGINES ~— FUEL-TO-NaK INTERMEDIATE HEAT EXCHANGER, FOUR REQUIRED A 4 T FOUR REQUIRED AIR I o TURBINE PUMP T Igh\ 3 i L\ A A «_- BLEED AIR CONTROL VALVE ) A Y AIR —— NaK-~TO-AIR - RADIATOR DIFFUSER \COMPRESSOR TURBINE NOZZLE Fig. 3.7. Schematic Diagram of Power Plant System. 41 ANP QUARTERLY PROGRESS REPORT (56% Na, 44% K) pumped in a closed loop through the heat exchanger and through NaK-to-air radiators located in the engines between the compressor and the The NaK would be pumped by a variable-speed bleed-air and the control of the bleed air might serve as the primary control turbine, turbine, of the system. A variable nozzle area would also be provided on the engines. The bleed air required to pump the fuel and the NaK was determined to be from 1 to 2% of the total engine air flow at full power, For full power operation, control of the nozzle area could be coupled to way that the be kept constant, close to maximum a governor 1in such a engine This efficiency for the turbine and com- pressor, maximum air flow, and thus maximum thrust, To reduce power, the bleed air to the NaK-pump drive turbines could be throttled to reduce the NaK flow rate and, in turnm, the quantity of heat transported by the NakK. In addition, the automatic nozzle area control could be bypassed for any throttle setting below full rpm would would give power, Since the turbojet nozzle area could be the full position (for the operating speed and altitude), the engine rpm would be reduced as the throttle for the pump frozen in open drive turbine was being closed., This feature was found to he necessary because, if constant turbojet rpm were maintained, the temperature of the NaK returning to the intermediate heat exchanger might be below the freezing point of the fuel for operation below about 85% power, Jt 1s expected that the negative temperature coefficient of reactivity of the fluoride fuel in the core will be sufficiently high to maintain the mean temperature of the core at a constant value and to maintain the effective multiplication constant at 1.00, on the reactor. regardless of the power demand 42 The principal parameters that were used in the calculations of the system are given in the following: Air flow rate per engine {at sea- level static, which was estab- Jished as the design point} 220 1b/sec NaK flow rate for full power per engine 480 1b/sec NaK-to-air radiator heat transfer 2 area per engine 10,000 ft NaK-to-air radiator volume per 3 engine 27.7 ft fuel-to-NaK intermediate heat ex- changer heat transfer area, 2 tota 2,400 fc Fuel flow rate 1,250 1b/sec Viaximum fuel temperature 1590°F Minimum fuel temperature 1190°F Maximum NaK temperature 1500°F Minimum full-power NaK tempera- o ture 1100°F Compression ratio of the main engine CcoOmpressor 4:1 It should be emphasized that all these parameters, and others, are completely i1nterrelated and that a change 1n one parameter would, in all probability, change all the others, All values must be considered as preliminary because the fluoride fuel is not definitely specified; typical values were assumed for the characteristics (specific heat, thermal conductivity, etc.) of the fuel., The heat transfer characteristics of all heat exchangers are preliminary because these units are being improved. hence, The flow characteristics of the fuel channels are currently being studied. It 1s believed, however, that this study does outline fairly well the control and performance characteristics of the system, The characteristics of the entire system were found to depend heavily upon the characteristics of the NaK- to-a1r vradiator. The radiator assumed was a finned-tube, flow design mul tipass, cross- , the characteristics of which have been established by test,(7) The NaK {flow rate of 2.5 cfs, as specified in the previous quarterly report,{®) was assumed, while the air flow rate established by the operating speed and altitude. The fluoride fuel flow rate was determined from the maximum and minimum allowable fuel temperatures, the assumed specific heat, and the full-power reactor out- put. Tt was assumed that the fuel flow rate should be maintained constant under all operating conditions, With these parameters, the NaK-to-aar radiator heat transfer area required to transfer the full power output of was the reactor was calculated for sea- level static conditions, If the design point were chosen at a high altitude, a smaller heat transfer area would be required, but sea-level per- formance would be drastically reduced, If{ the design point were chosen at a high sea-level speed, the heat transfer area (and weight) would be increased and the thrust per pound of power plant weight would be reduced, In the course of the investigation, it was found that a reduced NaK flow rate reduced the size of the 1inter- mediate heat exchanger and the total power plant weight without an apprecia- ble increase in the NaK-to-air radiator size or the pressure drop in the fuel system. However, the optimization was not completed, and the initially assumed NaK flow rate was retained. Design Performance, With the established above, the of the power plant av various speeds and altitudes is as shown in Fig. 3.8. The solid lines are the performance at the speeds and altitudes indicated for a maximum fuel temperature of 1590°F (but a maximnm structural metal temperature of 1500°F). The dotted lines, labeled parameters pecxformance (), s, Farmer, A. P. Fraas, H. J. Stampf, and G, D, Whitman, Preliminary Design and Performance Studies of Sodium-to-4ir Rediators, ORNL-1509 (Aug. 3, 1953). PERIOD ENDING SEPTEMBER 10, 1953 “War Emergency QOperation,” are for a maximum fuel temperature of 1690°F (maximum structural metal temperature of 1600°F), The line labeled “*Chemical Augmentation, Sea Level,)” is the esti- mated performance with sufficient chemical augmentation (interburning) to maintain a constant turbine air inlet temperature at 1600°F. The tail cone openings required to maintain constant rpm were found to be reasona- ble., The maximum and minimum fluid temperatures varied with the speed and altitude within the limits indicated above, For part-power cperation, the thrust varied with rpm in a con- ventional manner, Minimum sustained operation under sea-level static conditions was as follows: Per cent normal rated rpm 52 NaK flow rate 50 1lb/sec Minimum NaK temperature 910°F The above engine performance calculations were applied to a typical 200,000-1b, four-engine sea plane. The performance of the aircraft with a 100-megawatt reactor and two combinations of nuclear and chemical power at sea level is given in Fig, 3.9. Top speed could probably be improved in both cases with a cleaner aerodynamic design than that possible for a sea plane, The arrangement with two nuclear-powered engines and two chemically powered engines gives better performance than that with only two nuclear engines, even with chemical The arrangement having two chemically augmented nuclear- powered engines was not considered as practical because of the low take-off thrust and rate of climb indicated. Fuel consumption was very high because both interburning and afterburning had to be provided. System Characteristics That Affect Reactor Control. The most significant part of the control portion of the study was an investigation of the nature and magnitude of perturbations avgmentation, 43 ANP QUARTERLY PROGRESS REPORT DW! 21158 MACH NUMBER 0 0.3 0.6 09 1.2 1.5 1.8 SEA LEVEL | " = g anm—— - - > 40 - oo g — — s ; / b z 000 T 5 20— 530 e et e o - _'_-__,,..fl' /’ L a '_-/ =z L ™ & 20 00 1% —berm Tl = 45,?‘_’_ — / 2 e T o I e :L—) 1 e e T - QO 11 o T ™" | E 1O - 6070 e T A e e e T T 7 | 0 e e oo e —— RS HEAT TRANSFER AREA IN AIR RADIATORS PER ENGINE , 10,000 ft2 NORMAL RATE _ e e WAR EMERGENCY fe \\Y____/” T T — CHEM!CAL. AUGMENTATION SEA LEVEL 10 R L N v O - SEA . x - L S S N w T v, T e =z e p— — | ———— o — pagl = . = N 15,000 ft _ . ] o .—--*""Wd—, it — Lo & ,___50109.9—--"‘"‘?;_'/—/ - N — 3 o I 4 S e T - e T "’ 45,000 1 ::,;"':-———-—*"'._»--f'" e e e e | = T > g | o 2 b e ifi;a_a—'t,‘,,{ ] e e e e v — -'—”__‘__.—._...——-—"" 0 0 0.3 0.6 0.9 1.2 1.5 1.8 MACH NUMBER Fig. 3.8. Power Plant Performance at Various Speeds and Altitudes. 44 DWG. 21160 60 ‘ THRUST AVAILABLE l e TWO CHEMICALLY-PCWERED ENGINES, J53~GE-4, AND TWO MODIFIED, WRIGHT, NUCLEAR-POWERED ENGINES MMAX =095 \\\ 40 | TWO NUCLEAR-POWERED ENGINES WITH CHEMICAL — ”,,f”// AUGMENTATION "W’;>§,f~ / 30 frer—ees el THRUST REQUIRED FOR \ 200,000 b SEAPLANE TOTAL THRUST fib X 40~ ENGINES ONLY 0. 0 0.2 0.4 0.6 0.8 10 MACH NUMBER Fig. 3.9. Sea-Level Performance of a 200,000-1b Pilane with a 100- Megawatt Reactor. that might arise in the system, ex- ternal to the reactor, and affect the reactivity., In circulating-fuel reactors, the only way that such perturbations can affect the reactor is through changes in the temperature of the fluid entering the core. The salient point with regard to reactor stabi1lity and control is the rate at which these changes may occur, If such changes take place very rapidly, that is, in time intervals of the order of 1 sec or less, they might conceivably cause serious power oscil- lations and temperature overshoots. Core inlet temperature changes may take place as a result of changes in control settings and operating con- ditions, or they may result from such accidents as pump or engine failures. In any case, the rates of change will depend upon the properties of the system, that is, the thermal capacities, PERIOD ENDING SEPTEMBER 10, 1953 transit times, etc. Although this analysis was for a particular power plant, the results should be typical. Since no large amount of heat would be added to the system (except in the case of a very severe fire), an abrupt and substavtial fluid temperature change could come about only through changes in either the rate of power in the rate of heat removal. For a circulating-fluoride- fuel reactor with the type of secondary circuit envisioned, a change in the rate of power generation could come about only as a result of a change in the rate of heat removal; hence, the rate of heat rejection to the air flowing through the turbojet engine generation or radiators is critical, and will, in turn, depend upon the engine air flow rate, the NaK temperature, and the Nak flow rate through the radiators, The engine air flow rate will depend mainly upon the engine rpm and the airplane altitude and speed. Variations in engine air flow with changes in aircraft altitude and speed will take place at relatively low rates and, hence are not of serious consequence so far as the fast response of the system 1is concerned. Variations in engine air flow caused by changes in engine rpm will be only moderately rapid; ac- celeration from idling at 40% rated speed to rated speed would require from 5 to 10 sec because of the high inertia of the rotor assembly. Since the air flow will be directly pro- portional to the engine speed, the change in heat rejection rate will be from not less than 8 to 50 megawatts per engine in from 5 to 10 seconds. The heat 1input to the air flowing through the engine under idling con- ditions will depend upon the method of engine control. The method chosen in this particular study was to vary both the jet nozzle area and the Na¥X pump speed, Other means of control were considered, 1including varying the amount of air allowed to bypass the 45 ANP QUARTERLY PROGRESS REPORT radiators or, similarly, allowing a part of the NaK flow to bypass the radiators. If only the jet nozzle area were varied, the engine heat consumption under 1dling conditions might be as much as 50% of the full- power heat consumption. Tt is unlikely that this method of control would be used alone, however, because i1t would be very difficult to maintain full NaK pump speed at low turbojet engine rpm with the air-bleed turbine type of pump drive that seems most promising. There does not appear to be any way 1n which step changes in the temperature of the circulating fuel entering the reactor core can be effected, either accidentally or deliberately, Even if, in some manner, a step change could be introduced into either the air stream entering the NakK radiator or the NaXK circuit, the transit times through the radiator and the heat exchanger, the heat capacities, and the time lags in the lines from the radiator to the heat exchanger and from the heat exchanger to the core would most certainly cause the step change to be ‘““ramped out’ over a period of the order of tenths of a second or longer. Some of the typical mechanical that might be serious perturbations are interesting failures sources of indication of the rates of change that must be considered. Turbo- as an jet engine failure could result from three major the A com- failure of any of the components, the radiator, or the pressor failure could result from the breakdown of a thrust bearing or from the fatigue failure of a blade. A radiator failure could result from compressor, turbine. 46 the connection, rupture of a tube or a welded Turbine failure could result from the fatigue failure of a turbine blade., In any case, failure would act to decrease the power or heat demand on the reactor, As the rate at which heat removal from the fuel passing through the heat ex- an engine the fuel tempera- This, in turn, will act immediately to reduce the reactivity and hence the core power changer 1s reduced, ture will increase, level. A failure of a turbojet engine could not cause a step-type tempera- ture change in the fuel entering the reactor core. Simultanecus failure of all four engines, however, would be drastic because without the engines operating it would not be possible to remove the tremendous heat generated in the core., Even though the creased fuel temperature in the core caused the power to drop radically, to the extent of shutting down in- even the reactor, a large source of after- heat from the fission-product decay activity would still remain. In such a situation, there would be no facilities for heat removal other than the heat capacities of the components, and, in a relatively short time, the vessel containing the fuel would melt. However, circumstances in which all four engines might fail simultaneously are exceedingly rare and would probably result in the loss of the aircraft, Other possible sources of system failure are NaK pump failure, air turbine failure, rupture of bleed air or NaK lines, ete. ‘These would affect the system in the same manner as the primary type of engine failures de- scribed above, but probably at less rapid rates, PERIOD ENDING SEPTEMBER 10, 1953 4. UNIFORM THERMAL-NEUTRON FLUX REACTOR J. W. Morfitt, Development Divisions (Y-12) A. D, Callihan, Physics Division The relationship between minimum critical mass and uniform thermal- neutron core flux has been experimen- tally investigated for a water-moderated water-reflected nuclear reactor employing U?*3® as fuel. The reactor vessel assumed was a right-circular cylinder of aluminum 72 cm in diameter and 91.4 cm long. The core, 30.2 cem in diameter, was divided into five concentric regions by aluminum parti- tions and was surrounded by an ef- fectively infinite water reflector on its lateral surface only. The aluminum matrix of the reactor 1s pilctured 1n Fig. 4.1. Aqueous uranyl fluoride solution, made from uranium containing 93.2% of the U?3% isotope, was added in various concentrations to the several regions to simulate a theoretical fuel distribution having a continuous con-~ centration gradient. This theoretical fuel distribution is that given by a calculation method developed in an analytical treatment of the problem by Goertzel,(x) who demonstrated mathe- matically that the condition of minimum critical mass in a suitably chosen thermal reactor required that the thermal-neutron flux be uniform in the fuel-containing region, - CRITICAL MASS The experimentallymeasured critical height and mass for the theoretically determined fuel loading were within 2.5% of the corresponding calculated parameters, These results, given 1in Table 4.1, clearly establish the validity of the Goertzel theory. _ The concentrations of the solutions in the fuel regions were then altered slightly to establish the critical height at 1ts predicted value, 41.6 (I)G. Goertzel, Reoctor Science and Technology, Vel. 2, Ne. 1, p. 19 (19%52) (TID*?ODI). cm, and thereby reduce the measured critical mass to 1055 g of U?35, The results, both experimental and calcu- lated, are compared in Table 4.2 with the corresponding masses obtained with the nranium uniformly distributed throughout a core of the same length and diameter. TABLE 4.1. CRITICAL PARAMETERS FOR THEORETICALLY DETERMINED FUEL CONCENTRATIONS CRITICAL CRITICAL HEIGHT MASS {(cm) {g) Calculated 41.6 1038 Measured 42.17 1061 TABLE 4.2, COMPARISON OF CRITICAL MASS FOR UNIFORM LOADING AND FOR THE GOERTZEL LOADING CRITICAL MASS CRITICAL Uniform Goertzel MASS loading Loading BED?;?ION {g) {(g) Caleculated 1296 1038 19,9 Measured 1162 1055 9,2 The critical mass for the Goertzel loading was about 9% less than that for a uniformly loaded reactor of identical dimensions. This measured difference is less than that calcu-~ lated from a modified variational solution of the c¢ritical equation 1in integral form by usinga constant trial function. Experience has shown that resul ts obtained by this method are about B% too high when applied to uniformly loaded reactors. An adjust- ment of this magnitude in the above data would bring the values satisfactory agreement. into 47 ANP QUARTERLY PROGRESS REPORT A Rl AR % A S 12426 PHOTO ON ! £G T fonsesit ¥ o O < T < o Ll o O = ONS UEL FLOW REG! — = 1 oo 2 \,w”wwn. .Mn,m.,\a.u..mvv..nrw. e 5 RO e S S 48 Matrix for Uniform-Thermal-Neutron-Flux Reactor. 4.1%. Fig. An exploratory investigation for obtaining experimental verification of a modification of the theory postu- lated by Goertzel and for predicting how the of fuel can be minimized reactor of less than optimum radius met with little success. Al- though some lowering of the critical mass was produced by the theoretically mass in a determined fuel distribution, a dis- crepancy of more than 35% was found to exist between theory and experiment. PERIOD ENDING SEPTEMBER 16, 1953 NEUTRON FLUX The measured thermal and nonthermal components of the neutron flux were in good agreement with those predicted by theory, and the thermal flux, except for deviations produced by a stepwise approximation to the i1deal fuel distribution, was uniform along a radius of the core, as shown in Fig, 4,2, A comparison of the experimental and calculated thermal-neutron fluxes is shown in Fig. 4.3. The longitudinal il DWG. 21182 10,000 90D Lorenene L i e S I 4 o ® ® 800G - @ bl e men e e ] TG0 6000 4000 - 3000 RELATIVE NEUTRON FLUX, ¢ (ARBITRARY UNITS) 2000 |- 1000 |- L« NONTHERMAL FLUX \\-\ | 0 0 o 4 6 8 10 17 16 18 20 RADIAL DISTANCE, &, MEASURED FROM AXIS OF REACTOR (em) Fig. 4.2 Radial Neutron Flux in the Uniform-Thermal-Neutron-Flux Reactor, 49 DWG. 21163 5% ABOVE AVERAGE i J 5% B'lELOW AV'ERAGE 1 0.8 | -t CALCULATED — ‘ | l | EXPERIMENTAL 06— — AVERAGE, MEASURED, THERMAL FLUX IN REGIONS 2, 3, AND 4 T SET EQUAL TO UNITY 0.4 s RELATIVE THERMAL FLUX, ¢ {ARBITRARY UNITS) 02 e L - ——GORE-REFLECTOR BOUNDARY ‘ 3 oL 1 ... ... \ 0 4 g 12 16 20 24 28 RADIAL DISTANCE, X, FROM AXIS OF REACTOR (cm) Fig. 4.3, Comparison of Theoretical and Experimental Values for the Radial Thermal-Neutvron Flux. neutron flux behaved as was expected, Approximately 96% of the fissions were caused by neutrons with energies below the 0.02-in.-thick cadmium cut-off energy; thus, the reactor was essen- tially thermal. This result was ex- 50 pected because the hydrogen molecule density of the fuel was about 99% of that of water., In general, data obtained from the experimental reactor were compatible with the postulates and predictions of the Goertzel theory. IMPORTANCE FUNCTION The importance function of the U%3°% in the minimum-mass reactor was measured by adding an increment of fuel to a localized volume in each of the five concentric fuel regions and by noting in each 1nstance 1ts effect on the over~all reactivity of the system. This fuel importance function was found to be radially uniform to within 5% of its average value. 1In a second test, equal volumes of fuel were ex- changed between the 1nner and outer regions, where the concentrations and the 1mportance functions differ most; the resulting net mass shift of 8 g of U?*% changed the reactivity only slightly. The theoretically predicted fuel concentration distribution for uniform thermal-neutron fluxis, there- fore, very insensitive to small con- centration changes, and the relation between critical mass and concentration distribution has a broad minimum, INTRODUCTION The primary concern in the research on high-temperature liquids has been the determination of the phase diagrams of fluoride and chloride systems with and without uranium (sec. 5). Detailed study of the NaF-ZrF,-UF, system has continued because of the ARE require- ment for higher uranium concentration, The composition containing about 6.5 mole % UF, obtained by the addition of NazUF2 {(the ABE fuel concentrate) to NaZrFs {the ARE fuel carrier) should provide a suitable fuel., [Data are reported from several other systems containing either UF,, UF,, ThF,, or UCl, which have been examined in the search for suitable aircraft fuels. To date, most of the data have been obtained by the direct thermal analysis technique. Since this technique 1s not adequate to completely define the phase equilibria of complex systems, other techniques, including quenching, differential thermal analysis, high- temperature x-ray diffraction, and bigh-temperature phase separation, are also being employed., Several problems associated with the preparation and purification of fluoride mixtures have been investigated. The recent corrosion studies have been devoted almost entirely to the effect of various parameters on the corrosion of Inconel by fluorides, although some work with hydroxides and liquid metals was continued (sec. 6). Studies of the corrosion of Inconel by the fuel NaF-ZrF, -UF, (50-46-4 wole %) as a function of time and of tempera- ture have provided a better picture of the corrosion mechanism. While the corrosion rates at 1500 and 1650°F in runs of 500-hr duration are comparable, the initial corrosion rate is higher at the higher temperature. These initial corrosion rates (~1 mil/day) decrease after several days by a factor of 10. While the initial corrosion mechanism 1s associated with the con- centration of fluoride contaminants AND SUMMARY (NiF, and FeF,) and can be minimized, the secondary corrosion mechanism may be an inherent limitation of Incomnel- fluoride systems. The beneficial effect of adding ZrH, has been demon- strated in additions of guantities as small as 0,1 wt % - an amount consistent with the known concentration of structural metals in the fuel. The corrosiveness of fluorides on Inconel in tests at high fluid velocities 1is apparently no greater than that 1in static tests. This is not generally true when the fluid is liquid sodium, however; its attack on some metals was more severe in rotating tests, Of the several stainless steel convection loops recently tested with circulating lead, only the loop constructed of type 410 stainless steel did not plug. The fabrication of high-conductivity radirator fins and their subsequent assemblage into high-temperature high- performance radiator segments repre- sent the major accomplishment of the metal lurgical research program, Other work in this program included investi- gations of the mechanical properties of Inconel and the fabrication of pump seal and tubular fuel elements (sec. 7). The most satisfactory radiator fins are formed from 10-mil sheets of copper (for high thermal conductivity) clad with type 310 or type 346 stainless steel {for oxidation resistance). Three brazing alloys appear to be suitable with regard to melting point and corrosion resistance for the fabrication of radiator assemblies: a low-melting-point Nicrobraz, G-E No. 62 alloy, or Ni-P alloy. The Ni-P alloy is very promising, since it can readily be preplated (by an “elec- troless’” plating technique) to the to-be-brazed joint, even though the joint must be subsequently chrome plated to attain the desired corrosion Creep-rupture data have and resistance. now been obtained for both coarse- fine-grained Inconel in fluorides at 53 ANP QUARTERLY PROGRESS REPORT 815°C over the stress range 2500 to 7500 psi. Techniques for drawing tubular solid-fuel elements are being investigated, The high-temperature physical properties of several molten fluorides, chlorides, and hydroxides have been measured, the heat transfer characteristics of these liquids are being studied in (sec. 8). The viscosity and density of the latest ARE fuel NaF-ZrF, -UF, (53.5-40-6.5 mole %), measured, and the values properties proved to be very similar to the values for the previous fuel composition, In particular, the viscosity decreased from 16 cp at 580°C to 5.7 cp at 950°C, while the density did not change significantly, The measured fluid-to-wall temperature difference in a simulated circulating- fuel system fell within ¥30% of theoretical values, Additional heat transfer data for circulating NaF-KF- [iF eutectic in Inconel substantiated previous data which showed this entectic to have about one-half the and various systems have been for these heat transfer capacity of the compa- rable fluoride-nickel system. The poorer performance with Inconel was caused by a K,CrF_ film, The film, which was formed by mass transfer in the bimetallic system, was also re- sponsible for the decreased heat transfer capacity in the fluoride-to- NaK heat exchange system. The irradiation damage program included studies of fuel stability, 54 corrosion of beryllium oxide by sodium and of Inconel by fluorides, creep of metals, and the in-pile circulating loop, which 1s now being constructed (sec. 9). PRefined chemical and mass spectrometric techniques of fuel analysis have indicated that there is no gross segregation of the uranium in irradiated fluoride fuels., The corrosion of Inconel capsules 1s more severe when they are i1rradiated; how- it 1is not certain whether the radiation 1s directly or only 1in- With beryllium there was no ever, directly responsible. oxide in sodium, however, effect due to irradiation. The in-pile in the LITR with both Inconel and type 347 stainless steel show no serious effect of i1rradiation creep tests on creep rate, The analytical studies of reactor materials include chemical and petro- graphic analyses of fuel composition or corrosion products (see, 10). The zirconium in fluorides may be determined rapidly and precisely by adifferential spectrophotometric technique which utilizes the zirconium-alizarin red-S complex. Methods for determining the concentrations of the reactants and products of the reaction Cre + 3UF} = 3Uf3 + CrF3 are being investigated. Petrographic examination of about 750 fluoride mixtures, which involved the determi- nation of optical data for 1l unreported fluoride compounds, was completed, PERIOD ENDING SEPTEMBER 10, 1953 5. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS W. R. Grimes, Materials Chemistry Division In the course of the past three vears, the technique of thermal analysis has been applied to a large number of two- and three-component fluoride systems. The data obtained, which have been incorporated into a number of progress reports, have been sufficient to demonstrate which one of the various systems shows promise for reactor application but, especially in systems with complex behavior, have seldom served to completely define the equilibria involved. This “exploratory” study is nearing com- pletion., Among the fluoride systems studied, it appears that NaF-ZrF -UF,, NaF-BefF, - ur,, LiF-BeF,-UF,, and the corres- ponding systems containing thorium are phase of general value as reactor materials., Increasing emphasis will be placed during the next several months on the development of detailed and complete phase diagrams for these complex three-component systems and their associated binary systems. The NaF- Zr¥,-UF, system will be studied fairst because it is to be used in the ARE. In this program, the techniques of differential thermal analysis, high-temperature x-ray diffraction, high-temperature phase separation, and x-ray diffraction will be used simultaneously on the svstem under study. Apparatus has been prepared and tested for each of these methods, and the applicability of the methods to certain aspects of phase equilibria in fluoride systems has been demonstrated. quenching, THERMAL ANALYSIS OF FLUORIDE FUELS C. J. Barton W. C. Whitley L. M. Bratcher J. Truitt Materials Chemistry Division Although the direct thermal analysis method of studying fluoride fuels 1is expected to be supplanted by other technigues during the next year, it has been used considerably during this quarter, In general, the experiments have been confined to exploratory studies of systems containing ThF,, to re -examination of several binary fluoride systems to refine the data so that other techniques can be applied directly, and to examination of some systems for which the data obtained previously were quite meager. No previously unreported fluoride systems containing UF, were investi- gated during this quarter. Further study was applied, however, to several systems for which previous thermal analyses were incomplete or needed to be refined. Early studies of the NaF-UF,, RbF-UF,, KF-UF,, and CsF-UF, binary systems were repeated, 1in large part., Only minor changes 1in the published diagrams seem to be indicated by these experiments. The final diagrams will be published after additional study by x-ray and quenching techniques. Some new data were ob- tained on the KF-ZrF,-UF, system. Study of the complex Nal-ZrF,-UF, system was continued during this quarter. Thermal analysis data for several ‘“jJoins” in this system have been obtained in an effort to improve understanding of the phase relation- ships in this system. One such join, the Na,UF, -NaZrf, system, is of especial importance, since the proposed ABRE fuel makes use of the two com- ponents, Studies of fuel mixtures containing UF, included the systems UF, -ZrF_, UF,-UF,-Z¢rF,, and NaF-ZrF,-UF, -UF4. Some data on the first two of these systems were reported previously,(1) (l)V. 8. Coleman, C. }. Barten, and T. N. McVYay, ANP Quar. Prog. Rep. June 10, 1953, OBNL- 1556, p. 41. 55 ANP QUARTERLY PROGRESS REPORT Study of the Na¥F-ZrF,-UF,-UF, system was initiated in an effort to identify a yellow phase that is present when slight reduction of NaF-ZrF,-UF, mixtures occurs, In connection with these studies, about 1.5 kg of UF, (99% pure by petrographic examination) was synthesized by a previously described method,¢2) Na¥-ZrF, -UF,. Thermal data indicate that the liquidus line for the Na,UF_ - NaZrF, join has the shape shown in Fig. 5.1. No high-melting-point compositions will result from mixtures of these components in any proportion, Recent 1nterest in higher uranium- content fuels has focused attention (2)Vl. C. Whitley and C. J. Barton, ANP Quar. Sept. 10, 1951, ORNL-1154, p. 159. Prog. Rep. 675 on the region of Fig. 5.1 that lies between 77.5 and 82% NaZrF, (7.5 to 6.0 mole % UF,). The data in this region indicate a steady increase in melting point with increasing UF, concentration, Further study has probable existence of a eutectic at about 63.5 mole % NaF, 19 mole % ZrF,, 17.5 mole % UF,, which melts at 600 + 5°C. KF-ZrF,-UF,. Thermal obtained on several mixtures in the KF-ZrF,-UF, system, and special pre- were taken to minimize hydrolysis. The evidence from all the mixtures tested 1s that, in spite of the low melting point (445°C) of KZr¥F., addition of as little as 2 mole % UF, produces high-melting-point indicated the ternary data were cautions DWG. 21103 650 625 e L 600 | i - )_ 1/2 Ni (e) + 1/4 ZrF, (c) , 1/3 FeF, (c) + 1/4 Zr (¢) —2> 1/3 Fe (¢} + 1/4 Zr¥, (o) , 1/3 CrF, (¢) + 1/4 Zr (c) —> 1/3 Cr (c) + 1/4 Zr¥, (c) , The possibility of reducing these and other extraneous materials, such as hydrogen fluoride, with metallic zirconium has been tested in several experiments., In general, metallic zirconium (machined crystal bar) has been suspended in the 800°C melt after hydrofluorination and hydrogenation. Agitation of the melt has been ac- complished in each case by sparging AF° = -31.8 kcal , AF° = ~32.7 kecal , AF® = 22,2 kcal . indicate that the potential varies from 0.5 volt after 2 hr to less than 0.1 volt at the conclusion of the treatment. In every case the weight loss of the zirconium bar is higher by a factor of at least 5 than the loss expected from the reduction of structural metal ions. The melt so obtained has heen shown to be capable of reduction of considerable amounts of added nickel 69 ANP QUARTERLY PROGRESS REPORY fluoride after removal of the zirconium bar. Analytical tests made after filtration of the product through sintered nickel indicated that the “reducing power” was still]l present. It appears likely that this reducing action 1s due to soluble di- or tri- valent zirconium fluoride. Further study of this phenomenon 1s planned. Treatment of NaBeF, with Metallic Rerylliam, In connection with a study of the utilization of metallic reducing agents in the purification of fluoride melts, beryllium metal was used in conjunction with hydrogen on a sample of NaBeF; prepared for trials as a frozen seal material, The beryllium metal lost weight in an amount corres- ponding to 710 ppm of beryllium con- sumed by the melt even though the melt had previously received a prolonged hydrogen treatment., The melt was allowed to solidify without filtration, and it was found teo be badly contami- nated with reduced metals and with unidentified reaction products, Reduction of Dissolved Chromous Fluoride by Hydrogen. The raw ma- terials for production of ARE fuel mixtures are, 1n general, low 1in chromium compounds; therefore very little information is available regard- ing possible reduction of such ma- terials by hydrogen. Accordingly, a sample of CrF,, corresponding to 0.1 wt % of Cr*?, was added to a completely hydrofluorinated and hydrogenated bat h of NaZrF, and the resulting melt was treated at 800°C with 166 liters of hydrogen. Through- out the experiment, the hydrogen fluoride concentration was constant at 2 x 1075 mole of hydrogen fluoride per liter of exit gas. Since the back- ground hydrogen fluoride concentration is about 1 x 105 mole of hydrogen 70 fluoride per liter of hydrogen, the hydrogen fluoride generated corresponds to a reduction of abont 1% of the CrF2 added, Chemical analysis of the product showed 830 ppm of chromium, whereas 1000 ppm was added. These results seem to indicate that divalent chromium is reduced by hydrogen very slowly, if at all, Preparation of Various fFiuorides (L. G. Overholser, B, J., Sturm, Materials Chemistry Division). Ad- ditional batches of Fer, FeFS, NiF,, CrF,, Cr¥,;, (NH,),;CrF,, Na,CrF,, and Na,FeF, have been prepared by the methods described previously,(17+18:19) Anhydrous CdF, was prepared by de- hydration under HI' of the precipitate obtained by adding NH,HF to an aqueous solution of Cd(NO;),. Previously, it was found that heating CdCl,-2 1/2 H,0 under HF failed to convert the chloride guantitatively to the fluoride. It was also learned that the precipitate formed by adding NH,HF to an aqueous solution of CdCl, contained equivalent gquantities of chloride and fluoride. This suggested the formation of a douhle salt., Anhydrous Agl was pre- pared by treating Ag,CO, with HF at about 150°C., The product was darkened by a small gquantity of metallic silver, Small batches of anhydrous NiCl2 and FeCl, were prepared for special uses. These anhydrous chlorides were prepared by slowly heating the re- spective hydrates to 600°C while anhydrous HCl was passed through the systeim. (17)F. F. Blankenship and G. J. Nessle, ANP Quar. Prog. Rep. Dec. 10, 1952, ORNL-143%, p. 122. (18)F. F. Blankenship, G. J. Nessle, and H. W. Savage, ANP Quar. Prog. Rep. Mar. 10, 1953, ORNL- 1515, p- 113. (19)8. J. Sturm and L. G. Overholser, ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p. 48. PERIOD ENDING SEPTEMBER 10, 1953 6. CORROSION RESEARCH W. D. Manly Metal lurgy Division W. R. Grimes F. Kertesz Materials Chemistry Division H, W. Savage ANP Division During this quarter, the static, seesaw, and rotating corrosion testing facilities have been used primarily for the study of the corrosion of Inconel by a proposed fluoride fuel mixture, NaF-ZrF, -UF, (50-46-4 mole %). The few rotating tests completed showed no different or greater attack than that experienced in static and seesaw tests. Among the corrosion parameters examined in the latter tests of 100-hr duration at 1500°F were the effect and removal of an oxide layer, exposure time, oxide and zirconium hydride additives, tempera- ture, and uranium content of the melt, As expected, an increase in the uranium concentration 1ncreases corrosion,. The effectiveness of small (0.1%) additions of zirconium hydride in reducing the impurities is consistent with the known metal content of pure fuel after exposure to Inconel, An oxide layer on the Inconel also in- creases corrosion but can be readily removed by pretreatment of the metal with NaZrF,. Corrosion has been cor- related with a decrease 1in the 1iron content of the melt (and a compen- sating increase in the chromium con- tent) during the first 100-o0dd hr of exposure, Although the depth of attack is essentially independent of temperature from 800 to 1400°F, the attack is much greater at temperatures from 1600 to 2000°F. Fluoride corrosion studies in Iinconel thermal convection loops supplement and extend the data obtained above, Although these loops normally operate for 500 hr with a hot-leg temperature of 1500°F, the effects of both time and temperature have been studied. While the corrosion rates after 500 hr at 1500°F and at 1650°F are comparable, the initial corrosion rate is higher at the higher tempera- ture, However, high initial corrosion rates {correlated with NiF, and FeF, fuel contaminants) are superseded around 250 hr by much lower corrosion rates, possibly because of the oxi- dation of chromium by UF,. A type 316 stainless steel insert in the Inconel loop was preferentially attacked by the fluorides, whereas graphite immersed 1in the fluoride causes carburization of the Inconel loop. The hot-leg deposits, previously encountered when zirconium hydride was added to the loops, have been eliminated by adding the hydride to the fuel before it is filtered when the loop is filled, The postulated long-term mechanism for the corrosion of Inconel by fluorides 1s being investigated in a study of chemical egquilibria in fluoride systems, Equilibrium constants for reactions of the type FeF2 + Cr == Cer + Fe and 2UF, + Cr g==== QUF, + CrF, , where the active species are dissolved in NaF-ZrF, melts at high temperatures, are being measured. A limited number of tests were performed with liquid metals, including sodium, lithium, and lead. In general, spinner tests with sodium show greater attack than comparable static sodium tests., Jn static tests with lithium, types 309 and 316 stainless steel 71 ANP QUARTERLY PROGRESS RFPORT exhibited superior corrosion resistance (1 to 2 mils in 400 hr at 1000°C) to that of types 347 and 430 stainless steel. Additional studies on the mass transfer of container materials in lead have been continued with the use of small thermal convection loops. Of the metal loops tested recently, only the type 410 stainless steel loop did not plug, and i1t showed only a small amount of mass transfer; nickel-iron alloy (30-70 wt %), chromium, nichrome V, and nickel loops plugged in 275, 100, 12, and 2 hr, respectively. The corrosion of structural metal by hydroxides at 1500°F has not been sufficiently well controlled to permit consideration of these liquids for use in high-temperature reactors. Hy- droxide corrosion is at a minimum, however, in nickel (and silver and gold) systems in which a hydrogen atmosphere 1s maintained. In recent studies on hydroxide corrosion, attempts have been made to study the oxidizing power of hydroxide corrosion products and to determine the equi- librium pressure of hydrogen over the hydroxide in several metals, FLUORIDE CORROSION OF INCONEL 1IN STATIC AND SEESAW TESTS Ho J- Butt,ram C- f{n Cro.ft R. E. Meadows Materials Chemistry Division D. C. Vreeland E. E. Hoffman Metallurgy Division Both the static and the seesaw tests provide relatively cheap and simple means of investigating the many parameters which affect fluoride Once an effect or mechanism isolated, 1t 1is corrosion, has been fairly well then subjected to moere severe and extensive loop tests (cf.,, ‘“Fluoride Corrosion of Inconel in Thermal Con- vection Loops,” below). The static tests in which the fluoride mixture 1is sealed in an Inconel capsule were operated for 100 hr at 816°C unless 72 otherwise stated., The seesaw tests in which the capsule containing the fluoride 1s rocked in a furnace were operated at 4 cps with the hot end of the capsule at 800°C and the cold end at 650°C, Effect of Oxide Layer. Structural metal oxides are known to be unstable in the presence of ZrF,-bearing mixtures with respect to reactions of the type ZrF, + Ni0 —> ZrOF, + NiF, ZrF4 + 2Ni1Q ——> ZrO2 + 2NiF2 and 3ZrF4 + 2Cr,05 —> 3Zr0, t+ 4CrF, , Since such structural metal fluorides are soluble to a considerable extent ’ in the molten mixture and can attack the Inconel by reaction with the chromium, oxide films on the metal walls are a potential source of in- creased corrosion. The chromium uptake of a pure preparation of NaZrF, in degreased Inconel capsules is compared in Table 6.1 with that observed when the Inconel was subjected to a 24-hr exposure 1in air at 1000°C prior to the test. Both specimens were exposed for 100 hr in the seesaw test apparatus. The un- oxidized specimens revealed, upon metal lographic examination, scattered subsurface void formation to a depth of 0.5 mil. Heavy subsurface void formation to a depth of 2,5 to 3 mils was observed in the oxidized tubes at the hot end., Although the oxide layer still visible in some spots in these specimens, the cold ends of the tubes also showed moderate oxide attack. It 1s of interest to note that when pure fuel is used on Inconel in the as-received condition, a total of 30 to 35 meq/kg of iron and chromium was compounds is found after testing. This lower limit of corrosion products has been observed i1n several experiments (cf., “Effect of Small Zirconium PERIOD ENDING SEPTEMBER 10, 1953 TABLE 6.1. EFFECT OF OXIDE LAYER ON RESISTANCE OF INCONEL TO FLUORIDE MELTS METAL CONTENT AFTER TEST* NO. OF TREATMENT OF TUBE ' meq/kg** TESTS | ppm . - meq/kg Fe Cr © N1 Fe : Cr Ni 3 Oxidized at 1000°C 385 1600 85 13.8 61.6 3.0 2 Degreased 140 725 25 4.9 28.0 0.78 .Metal content before test: Fe, 20 ppm; Cr, 20 ppm; Ni, 20 ppm. **Divalent ions assumed for calculation, Hydride Additions,” below). It is lographic examination of specimens possible that in the handling of the pure powdered NaZrF, mixture, suf- ficient water is picked up to account for the observed concentration of metallic constituents, It also appears possible that as-received Inconel has an oxide layer that is sufficient to account for this attack. Removal of Oxide Layer. The ability of the fluoride NaZrF, to remove the oxide layer on Inconel 1s being in- vestigated because of 1ts possible use in the ARE. : The specific objectives of these latest tests were to determine the lowest temperature at which the fluoride will remove oxide from Inconel and the amount of attack which takes place during descaling. Static tests were run with NaZrF. and Inconel specimens that had been oxidized for 24 hr at 1500°F, Results of these tests can be seen in Fig. 6.1. All specimens were completely descaled, except the one treated at 950°F. From this series of tests, it would appear that a temperature of 1000°F could be recommended for descaling Inconel. However, since previous tests indicated that descaling was not accomplished after 4 hr at 1000°F, a temperature of 1100°F gives the more certain results. In the previous tests the specimens were electropolished, before oxidizing, and possibly the oxide layer on these specimens was more adherent. Metal- which had been oxidized and then descaled showed that only very light attack is to be expected during the usual short time of descaling. Effect of Oxide Additive. As mentioned above, the use of NaZrF5 to remove the oxide laver on Inconel 1is being investigated. Since buildup of zirconium oxide in the fluoride with successive cleanings had been reported, static corrosion tests were made with the NaZrF, to which 0.75 and 1.5% Zr0, had been added. There was possibly a slight increase in depth of attack by the fluorides containing zirconium oxide during the 100-hr test at 816°C; but for the short times used for descaling, no measurable increase 1in attack as a result of the zirconium oxide in the fluoride should be expected, Effect of Small Zirconium Hydride Additions. The beneficial effect of zirconium hydride additions on the corrosion behavior of fluoride melts in Inconel has been discussed in several previous reports. The most recent report in this series(1) indi- cated that for the NaF-ZrF, -UF, mixtures tested the 0.1% addition of ZrH, showed nearly the same effect as larger additions, In additional studies of this problem, amounts of zirconium (I)H. J. Jure 10, Buttram et 2l., ANP Quar. Prog. 1953, OBNL-1536, p. 51. Rep. 73 ANP QUARTERLY PROGRESS REPORT hydride as small as 0,01% were added to NaF-ZrF,-UF, (50-46-4 mole %) mixture and to NaZrF, before exposure to Inconel in a seesaw test for 100 hr. The data obtained are shown in Fig. 6.2. The higher values for soluble chromium in the NaZrF_ when no addition of zirconium hydride was made 1is probably the result of the NaZrF, containing 315 ppm of iron, whereas the NaF-Zr¥F, -UF, mixture contained only 115 ppm. The uranium-bearing mixture 1s the more corrosive, however, at all zirconium hydride concen- trations, probably because of reactions of the type ZUF4 + Cr —> CrfF, + 2UF; . The interesting point 1is that the beneficial effect of zirconium hydride is obtained when as little as 0,06 to 0.1% of the material is added to either specimen, If the reducing power of the hydrogen is neglected, 0.08 wt % ZrH, furnishes 35 meq of reducing agent per kilogram of fuel. This figure is in excellent agreement with the amount of structural metal fluorides known to be present in the fuel (cf,, “Effect of Oxide layer,” above). The guestion is still not answered as to whether the material that causes the corrosion and that i1s reduced by zirconium hydride is an unknown impurity in the fuel, is the hydrogen fluoride intro- duced by hydrolysis during handling, or is the oxide on the Inconel. Eifect of Exposure Time,. report{!) gave some data for the structural metal content of the fluoride melt after exposure times of from 1 to 10 hr in Inconel capsules in the tilting furnace. Additional data for exposure times of from 2 to 128 hr A previous UNCLASSIFIED Y-9553 INCONEL AS RECEIVED ";E*"“ _é f—“"————' oS | 950°F 1000°F 1050°F Fig, 6.1. Descalipopg of Oxidized Inconel by NaZng. Specimens treated for 4 hr at indicated temperatures, 74 Dwe. 2111t 1000 HO0) foenkeeaa R e e aa s T r s an % wia man nam ok aam Ak naa o NaF-ZrFQ--UF4 150-46-4 mole %) - o | - NqF—~ZrF4 (50-50 mole %) CHANGE N CHROMIUM METAL CONTENT {(ppm) ol —~ | e | | -500 ‘ 0 0.1 Q.2 03 ZrH2 ACDED (%) Fig. 6.2. Chromium Uptake of Two Fluceride Mixtures as a Function of ZrH2 Additions. that show the concentration of iron, chromium, and nickel in the melt as a function of exposure time are presented in Table 6.2. It appears that the chromium content of the melt rises immediately in each case, whereas the iron values remain essentially constant during the first few bhours, The chromium content continues to increase steadily, but the iron content drops. In NaF-ZrF, - UF,, the final value for chromium content was considerably higher than that expected from stoichiometric considerations, Corrosion by Fluorides with High UF, Concentrations. There are licttle data available on the behavior of fluoride mixtures with high uranium concentrations and with low concen- trations of structural metal 1i1ons. Since a high-uranium concentration PERIOD ENDING SEPTEMBER 16, 1953 TABLE 6. 2. METAL CONTENT OF FLUORIDE MIXTURES AS A FUNCTION OF EXPOSURE TIME METAL CONTENT AFTER TEST (meq/kg) (%) EXPOSURE | NaF-ZrF,-UF,(P) NaF-2rF, (¢) TIME (hr) Mixture Mixture Fe | € | Ni | Fe | cr | Wi o(d) 4.1} 0.4 1.5|11.1] 0.8 | 1.0 2 6.31 57| 1.4 |11.8] 5.2 | 1.4 4 6.4 1 4.6 1.2 [10.0 5.2 | 1.9 16 2.7 (12 0.8 §.2111.2 1.1 64 1.1 |13 0.7 | 2.6 |14.6 | 1.3 128 1.3 |21 0.7 | 1.3 |14.6 | 0.7 {a)Each element considered to be in the divalent state, (6)50 mole % NaF, 46 mole % ZrF4, 4 mole % UF4~ (€)tp mole % NaF, 50 mole % Z:F,. (d)Each value is the average of duplicate determinations. mixture will be used as the fuel concentrate for the ARE (cf,, sec. 5), a series of seesaw tests with the NaF-ZrF,-UF, mixture (50-25-25 mole %) in Inconel capsuleshas been initiated, The available batch of this mixture contained, initially, about 40 ppm Fe, 25 ppm Cr, and 20 ppm Ni, and, since it had been given a very thorough treatment with hydrogen during prepa- ration, it probably contained a small quantity of UF;. In the standard 100-hr tilting test, this material produced scattered subsurface void formation to a depth of 0,5 mil and slight roughening of the metal at the hot end of the tube, Chemical examination of the melt after test revealed the presence of about 3 meq of Fe'' and 18 meq of Cr*' per kilogram of mixture. This attack can be considered as being slightly less than that commonly observed from more dilute fuel mixtures of somewhat lower purity; this slight improvement probably reflects the presence of UF,, [ ANP QUARTERLY PROGRESS REPORT The addition of 0.7 wt % ZrH, had little, if any, effect on this slight corrosion, However, the chromium content of the melt, as determined after testing, dropped from 18 meq/kg to about 6. Corrosion of High-Purity Iaconel. Static tests were run on several specimens of high-purity Inconel pre- pared by the Metallurgy Division. The nominal analysis of these specimens (by weight) was 15% Cr, 78% Ni, and 7% Fe, but the C, S, Ti, Mn, Al, and Mg contents ranged to a maximum of 1.6, 0.026, 0.25, 0.25, 0,15, and 0,05%, respectively, in the varions Inconels. Specimens of these materials contained in off-the-shelf Inconel tubing were tested in NaF-KF-LiF-UF, (10,9-43,.5- 44,5-1.1 mole %) at 816°C for 100 hours., 1In general, the as-cast, low- carbon specimens had fewer subsurface voids than did the extruded specimens, but attack of the extruded specimens was not excessive, In the tests in which the as-cast and the extruded material from the same ingot were tested, no significant difference in corrosion was observed, The attack in the as-cast specimen containing 1.6% C was the most severe, but depth of attack 1n this specimen was only slightly greater than that in the low- carbon (0.03 wt %) specimens., With the exception of this high-carbon material, the laboratory-melted alloys were not significantly different from commercial alloys with respect to attack by the fluoride, Effect of Temperature. Static cor- rosion tests were made on Inconel im NaF-KF-LiF-UF, (10.9-43.5-44.5-1.1 mole %) at seven different tempera- tures covering the range 800°F (427°C) to 2000°F (1093°C), The duration of each test was 100 hours. Within the range 800 to 1400°F, the depth of attack seemed to be i1ndependent of temperature, being approximately 0.5 mil. Within the range 1600 to 2000°F, there was no systematic variation in 76 depth of attack with temperature; but the depth of attack was much greatex than that at lower temperatures, being from 2 to 4 mils. FLUORIDE CORRGSION OF INCONEL IN ROTATING TEST D. C, Vreeland E. E. Hoffman Metallurgy Division Several tests were completed on the NACA-type rotating apparatus.(?> Witl this apparatus, fluid velocities of ug to 10 fps can be obtained with e maximum temperature of 810°C and temperature drops of from 20 to 75°C. The attack of Inconel by fluorides was no greater than the attack during static tests. Additions of titanium formed a surface layer and inhibited attack, Additions of NiF, and CrF,, as expected,(?) increased attack., Some small globular crystals were found attached to the tube wall in the test of fluorides to which 5% NiF, was added. Chemical analysis of the crystals revealed their composition tc be 92.34% Ni, 6.53% Fe, 0.99% Cr, and 0.14% Mn. FLUORIDE CORROSION OF INCONEL IN THERMAL CONVECTiON LOOPS G. M. Adamson, Metallurgy Division The use of thermal convection loops for determining dynamic corrosion by l1quids has been previocusly described.(?) Unless otherwise stated for the tests described in the following sections, the temperature of the hot leg of the loop was maintained at 1500°F and the temperature of the uninsulated cold leg was approximately 1300°F, With the fluoride salts, this temperature difference results in a fluid velocity of about 6 to 8 fpm. The usual testing period was 500 hours. (Z)D. C. Vreeland et al., ANP Quar. Prog. Rep. Mar. 10, 1953, OPNL-1515, p. 121. 3rpid., p. 119. Zirconiuam Hydride Additive. Previ- ously,{*) when zirconium hydride was added to the fluoride in an Inconel loop, the depth of attack was reduced, but a layer was deposited on the hot- leg wall, An Inconel loop has been run with NaF-ZrF,-UF, (50-46-4 mole %) treated with zirconium hydride 1in the fill pot, After the batch had been held at 1200°F and agitated for 3 hr, it was transferred to the loop through a micrometallic grade-G filter; the loop was then operated for 500 hr at 1500°F. No layer could be found in the hot leg, and the hot-leg attack had been reduced to a maximum pene- tration of 2.5 mils, which is the same as the attack obtained with previous zirconium~hydride additions, Effect of Type 316 Stainless Steel Insert in Inconel Loop. One Inconel loop in which a type 316 stainless steel section 6 in. long had been welded into the upper part of the hot leg was operated with NaF-ZrF,-UF, (50-46-4 mole %). The Inconel showed only very light and widely scattered subsurface void formation to a depth of 3 mils, and the depth of attack decreased near the stainless steel joint. The joining stainless steel showed a very rough surface, with some areas spalling off. Heavy attack extended to a depth of 12 mils and was primarily intergranular in nature, Figure 6.3 shows both the Inconel and the type 316 stainless steel surfaces, In the center of the stainless steel insert, the attack decreased to 8 mils, The Inconel below the stainless steel showed a reduction in attack. 1In the cold leg, a well-diffused surface deposit 0.3 mil thick was found. It appears, then, that stainless steel 1s preferentially attacked by fluorides in the presence of Inconel. Since protection is afforded both above and below the insert, an electro- (4)(}. M. Adamson, ANP Quar. Prog. Rep. Mar. 10, 1953, ORNL-1515, p. 123. PERIOD ENDING SEPTEMBER 10, 1953 chemical reaction is probably the cause of this protection. While it may be possible to protect an Inconel system by using stainless steel, such a combi- nation does not look promising. The major difficulty expected in such a system would be mass transfer, The major benefit derived from this experi- ment is the warning against obtaining corrosion data from a loop in which any stainless steel is present, Effect of Temperature on Inconel Corrosion. As an extension of work previously reported,(5) a loop was operated with a hot-leg temperature of 1250°F. The attack by NaF-ZrF -UF, {50-46-4 mole %) was a moderate-to- heavy subsurface void formation to a depth of 3 mils. These voids were small and evenly distributed, Six inches below the hot leg, no attack was found. Also, since the chromium in the fluoride did not build up during the test, 1t is concluded that little corrosion occurred., Since the attack was low in the upper section of the loop and completely absent in other sections, it seems likely that attack may be eliminated by operatang at aslightly lower maximum temperature, b UNCLASSIFIED T-3695 ¥ TYPE 36 Y% - STAINLESS STEEL &.#! T f -ty e - . - Fig. 6. 3. inconel -Type 316 Stain- less Steel Couple After 500 hr at 1500°F in NaF'ZI’F4~UF4 (50-46-4 mole %). 100X. Reduced 31%. 71 ANP QUARTERLY PROGRESS REPORT As previously mentioned, (%) no difference in maximum depth of attack was found when loops were operated at 1650°F instead of at 1500°F for 500 hours. After 100 hr of operation, however, the maximum attack in a loop operated at 1650°F was 7 mils, while in a loop operated at 1500°F, it was only 4 mils. After 100 hr at 1500°F, the holes were small and evenly distributed, that is, similar to those found at low temperatures in 500 hr of When the loop was operated the holes were larger and operation, at 1650°F, were concentrated in the grain bounda- ries, Carbon Imsert iw Incomel Losp. The use of uncanmned graphite in contact with the fluoride fuel in the core of a circulating-fuel reactor is being considered., In the external system, the fluoride fuel would be circulated (S)G. M. Adamson, ANP Quar. Prog. Rep. June 10, 1953, CRNL-1556, p. 60. Ya 3N e . . - . % .—". - . v 'fié - C ~ ' by Mfl@k @ 6.4. Fig. hr of Circulating NaF~ZrF4-UF4 (56-46-4 mole %) at 1500°F. 250X, leg. 78 "¢’ UNCLASSIFIED T.3818 in Inconel tubes, To determine the compatibility of such a system, an Inconel loop containing a graphite rod inserted in the center of the upper portion of the hot leg was operated with NaF-ZrF4-UF4 (50-46-4 mole %). Hesults obtained with this loop show that under these conditions Inconel may be carburized, 1In the hot leg, opposite the graphite rod, carbides were found in the Inconel grain boundaries to a depth of almost half the pipe wall thickness (Fig. 6.4). The carbides were also found in the cold leg, but they were evenly dis- persed and were to a depth of only 1.5 mils., The hot-leg attack in this loop was heavier than usual and extended to 13 mils, Upon visual inspection, the graphite rod appeared to be covered with a complete metallic coating. Under the microscope it was shown that this layer was actually a collection of non- metallic particles; each particle was 1 - .. UNCLASSIFIED | ¥ - T.3819 - - . - ’ . o' - . e : . . g . . . i 1 ° . : . . . . ! * ® o » . . ; . o . i « ¢ ;( . e . o ! . . » v . s " - ¢ : » .. - fi - + '«w 1 . * - ° ¢ * 4 £ O R Carburization of Inconel Loop Contzinimg Graphite Insert After 500 {a) Hot leg. (&) Cold covered with a thin, metallic skin, These particles had penetrated the graphite to a depth of 17 mils. This deposit is shown in Fig. 6.5. Spectro- graphic analyses showed the presence of chromium, zirconium, and sodium, and only traces of 1iron and nickel. A diffraction pattern revealed the presence of only the fuel, From this, it appears that the layer consisted of clumps of fuel covered with chromium metal. While attack was deeper than usual, the chromium content 1in the fluoride fuel was lower, Effect of Exposure Tiwe. In the previous report,(®) information was presented which indicated that after 250 hr very little, 1f any, increase in the maximum depth of penetration was found., This conclusion was based upon one series of loops filled from the same batch of fuel and operated for various times of up to 250 hr and from several loops filled with fuel from other batches and operated for 500 and 1000 hours., Information 1is now available from the 500-, 1000-, and 3000-hr loops filled from the same batch of fuel as that used for the 8 ypid., p. s8. | UNCLASEIFIED I8 13411 MILS Fig. 6.5. Deposit on Graphite Rod After 500 hr in an Incomnel Loop Circu- lating Na¥F-ZrF, -UF, (50-46-4 mole %) at 1500°F. 250x. Reduced 39%. in Figo 6.69 PERIOD ENDING SEPTEMBER 10, 1953 These data are plotted To show reproducibility, the data from the single-coolant batch are supplemented by ‘comparable data from other batches, The curve in Fig., 6.6 shows that while a change in slope takes place at around 250 hr, attack continues after 250 hours., This continued attack is based upon the single loop operated for 3000 hr and therefore needs to be confirmed. It is thought that the initial steep portion of the curve is caused by a reaction between the chromium in the Inconel and the contaminants in the fuel and the loop. The second phase of the reaction possibly represents the partial re- duction of UF, by chromium metal. An unexpected finding in the loop operated for 3000 hr was the presence of a metallic mass in the cold-leg sump. No deposit was found on the walls of the loop. The mass was identified spectrographically as chromium metal., The method of formation of this metallic chromiumis not known, since the reactions are usually con- sidered as an oxidation of chromium metal to chromium fluoride. One possibility 1s that this reaction 1s temperature sensitive and reverses itself in the cold leg. shorter times. OWG. 22, SR S N e ¥ % o SAME FLUDRIDE BATGH % & COMPARABLE BATCHES _”fim_”fi1w,Tm_.” MAXIMUM DEPTH OF ATTACK {miis) T o GO0 EXPDSURE TIiME (hr) Fig. 6.6. Depthof Attack 9f Inconel by Fluorides as a Function of Exposure Time. 79 ANP QUARTERLY PROGRESS REPORT FLUORIDE CORROSION OF STAYNLESS STEELS OF VARYING PURITY D. C. Vreeland E. E. Hoffman Metallurgy Division In an attempt to determine the effect of various carbon, oxygen, and nitrogen contents of stainless steels on resistance to attack by fluorides, static corrosion tests were made on eight specimens of stainless steels. Two specimens were commercial stain- less steel, types 304 and 305, while the other six specimens were stainless steels cast and extruded at MIT. Three of these special steels contained 18% chromium and 12% nickel, while the other three contained 18% chromium and 8% nickel, Each of the eight specimens was contained in a capsule machined from the same material as the speci- men, with the exception of the speci- men of type 305 stainless steel which was contained in tubing of type 304 stainless steel, The specimens were tested in NaF-KF-LiF-UF, (10.9-43.5- 44.5-1.1 mole %) for 100 hr at 816°C, The compositions of the steels and results of the tests Table 6.3. Insofar termined from these tests, variations of carbon, oxygen, and nitrogen con- tents within the ranges covered by the specimens tested had no measurable are shown 1n as can be de- influence on the resistance to cor- TOS10MN. LIQUID METAL CORROSION OF STRUCTURAL METALS D. C. Vrecland E. E. Hoffman J. V. Cathcart G. P. Swmith W. H. Bridges Metal lurgy Division Roftating Tests with Sodium. Several spinner tests have been completed in sodium at a temperature of 816°C and a speed of 405 fpm. The materials tested were types 310, 410, and 430 Inconel X, and As in the spinner tests stainless nichrome V. steel, conducted previously, attack seemed 80 to be more severe than that encountered in static tests with molten sodium. Surface layers were guite apparent on the Inconel X, nichrome V, and type 310 stainless steel and either less apparent or absent on the types 430 and 410 stainless steel. Weight changes were less on the types 410 and 430 stainless steel. Static Tests with Lithium, Static tests of types 309, 316, 347, and 430 stainless steel were tTun 1n lithium. Extreme care was taken® in rtegard to the details of these tests; for example, an extremely good dry box atmosphere was maintained during the loading of the tubes., These tests were run for A00 hr at 1000°C. Types 309 and 316 stainless steel exhibited corrosion resistance superior to that of the types 347 and 430 stainless steel, Both types 347 and 430 showed evidence of mass transfer on a macreo scale, The type 309 and 316 specimens had small weight losses, very little, if any, mass transfer, and not over 1 to 2 mils of attack. Details of these tests are given in Table 6.4, Dynamic Tests with Liguid Lead in Convection Loops. Studies have bzen made of the extent of mass transfer encountered with type 410 stainless steel, chromium, nickel, nichrome V, and a 30% nickel-iron alloy in small quartz thermal convection loops con- taining liquid lead. Details of the construction and operation of the loops and of the results obtained for Inconel, columbium, molybdenum, types 304, 347, and 446 stainless steel, and Armco iron have been reported previously.(7:8:9) Of the metals tested during the this quarter, type 410 stainless steel gave by far the hest results. The loop was operated for 545 hr with (T)G, P. Swith et al., ANP Quar., Prog. Rep. #ar. 10, 1953, ORNL-1515, p. 128. (8)G. P. Smith, Met. Quar. Prog. Rep. Apr. 10, 1953, ORNL-1551, p. 17. (Q)F. A. Knox et ael., ANP Quar. Prog. Rep. June 10, 1953, ORNL-15536, p. 64. 18 TABLE 6.3. STATIC CORROSION OF STAINLESS STEELS IN NaF~KF-LiF-UF4 (10.9-43.5-44.5-1.1 mole %) AFTER 100 hr AT 816°C COMPOSITION (=t %) ASTM MATERI AL GRAIN DEPTH AND TYPE OF ATTACK PHASE CHANGE Cr Ni C Né 02 SIZE Type;m4 18 8 0.1 G.02 ¢.01 8 to 9 |Subsurface voids 2 to 3 mils deep |Decarburization 1 to 2 stainless steel mils deeper tham voids; some phase transfor- mation Type 305 18 12 0.1 0.02 0.01 7 Intergranular attack and some subsurface |Same as above stainless steel voids 2 to 4 mils deep Special 18.24 [ 12.32 [0.151 {0.0044 {0.021 5 to 7 } Subsurface voids, some intergranular | Same as above stainless steel* attack 2 to 4 mils deep; attack 10 mils deep in one place along carbide segre- gation 18.55 [12.44 |0.006 [0.0039 [0.015 |2 to 4 [Mainly intergranular attack 3 to 5 mils |{Phase change deeper than in depth; one 10-mil-deep area along | attack along grain grain boundary boundaries 18.03 8.52 10.011 {0.096 {0.018 |2 to 5 {Mostly intergranular attack; some sub- |Some phase change noted surface voids 3 to 4 mils deep; one stringer attacked to 6 mils 18 12 0.001 (0.2 2 to 4 | Mostly subsurface voids, some inter- |Same as above granular attack to a depth of 5 to 7 mils 18.51 8.20 10.007 |0.0031 {0.031 {2 to 4 {Mostly intergranular, some subsurface | Same as above voids 3 to 7 mils deep 18.55 B.43 10.193 {0.9055 {0.017 |4 to 6 | Both subsurface voids and grain boundary |Same as above attack 2 to 4 mils in depth, attack tended to be deeper along carbide segre- gation *Prepared at MIT. ‘0T HAWALJAS INIOGNT aoiyiad £S61 ANP QUARTERLY PROGRESS REPORT TABLE 6.4. RESULTS OF STATIC TESTS OF VARIOUS STAINLESS STEELS IN LITHYUM AT 1000°C FOR 408 HOURS TYPE OF STAINLESS STEEL WEIGHT CHANGE (g/in.Z) METALLOGRAPHIC NOTES 309 -0.0022 316 -0 .0037 347 +0.0076 (crystals clinging to specimen) No evidence of mass transfer upon opening tube; metallographic examination revealed a white phase in grain boundaries throughout the specimen; 0,.5-mil crystals were attached to the surface in some areas; 0.25 mil of intergranular attack in scattered areas No evidence of mass transfer upon opening tube, but surface had an etched appearance; metallographic examination revealed very little attack; there were voids in some of the grain boundaries to a depth of 0.5 mil; there was a fine precipitate all around the specimen in a band approxi- mately 1 mil from the surface; in the vapor zone of the tube, there was 4 to 5 mils of intergranular penetration 1in a few areas LLarge amount of mass transfer on both tube and specimen, mainly at bath-level line; specimen was heavilyattacked intergranularly to a depth of 10 to 11 mils at one end; attack on rest of specimen did not exceed (.5 mil; tube 4390 No data attacked 1 to 2 mils; fine precipitate 1 mil under surface A few small crystals were attached to surface tube; surface of specimen had small grains and was attacked intergranularly to a depth of 4 to 5 mils; tubing attacked 2 to 3 mils hot- and cold-leg temperatures of 810 and 525°C. No plugging occurred during the test, but a small amount of mass transferred material was found in the cold leg of the loop. seen from Fig. 6.7, suffered slight, 1irregular surface attack, The results for this loop were comparable with those previously obtained for type 446 stainless steel, Neither nickel nor nichrome V (80% nickel~20% chromium) resistance to mass tramnsfer, As may be the test specimen showed much The loops for testing these materials were operated with hot- and cold-leg temper- atures of 810 and 525°C. Plugging occurred in the nickel loop after only 2 hr of operation, while the nichrome 82 loop operated for 12 hours. Large quantities of mass transferred material were found in both loops. The nickel specimens suffered heavy intergranular penetration., The corrosion encountered in the nichrome specimens was similar to that observed in Inconel, although the attack was more severe, As shown in Fig. 6.8, lead penetrated through-~ out the nichrome samples., Because of the good results obtained with i1ron-chromium alloys, such as types 410 and 446 stainless steel, 1t was expected that chromium would show a marked resistance to mass transfer in liquid lead., This was not the case, however; the chromium loop plugged in about 100 hours. Hot- and 6. 7. Fig. 545 hr at 810°C. 250X. cold-leg temperatures for this loop were also 810 and 525°C, The 30% nickel-70% iron alloy loop plugged in about 275 hours., Again, the hot- and cold-leg temperatures were about 810 and 525°C., Metal- lographic studies of this loop and of the chromium loop have not yet been completed, and therefore no data are available on the corrosion that occurred in these loops. As a result of the recent studies, 1t 1s apparent that much better results are obtained from the 400 series stainless steels than from the pure metals which comprise them. The reverse tends to be the case for the 300 series stainless steels previously tested. Further work will be done in an attempt to obtain an explanation of this behavior, PERIOD ENDING SEPTEMBER 10, 1953 f UNCLASSIFIED b Y9773 B -8 o — X Q & —0 — O o 5 | Z mtn — O o rx:;JQ, B Hot Leg of Type 410 Stainless Steel Loop After Circulating Lead for FUNDAMENTAL CORROSION RESEARCH J. V. Cathcart G. P. Smith L. Dyer Metallurgy Division F. A, Knox L. G. Overholser F. Kertesz J. D. Redman H. S. Powers Materials Chemistry Division Oxidizing Power of Hydroxide Cor- rosion Products. The preparation of higher valent nickel compounds, that 1s, the corrosion products which are formed in hydroxide-nickel systems, was attempted in order to study the oxidizing power of the compounds. Manganese, as the permanganate, was found as an impurity in lithium nickelate preparations, and NaNiQ, and LiNiQO, preparations were found to 83 ANP QUARTERLY PROGRESS REPORT LEAD-NICHROME INTERFACE i LEAD-NICHROME INTERFACE Fig. contain some cobalt as an impurity; therefore the preparations analyzed quantitatively for elements, Standard methods oxidizing power were unsatisfactory, and therefore a new technique and an were these for determining apparatus for analyzing LiNiO, prepa- rations were developed. The density of NaN10O redetermined to be 4.72 g/cm%. Several preparations of NaNi0O, were made for x-ray investi- gations. The reason for these x-ray studies was that some regularly oc- curring striations on the crystal faces were thought to be evidence for a decomposition reaction. By x-ray analysis 1t has ascertain that such striations origi- was been possible to 84 6.8. Hot Leg of Nichrome After Cirvrculating Lead for 12 hr at 810°C. ¢ ’ Lot .. UNCLASSIFIED © Y-9775 > 0.01 » Q o _ RIS | 500 A TSIRGN £ SO -fifibéfigfifi?fi ‘ ;_? NICHROME YfSava s’ " TEST SPECIMEN 5wt - ~‘%%pfin*¢:' \’; .. d 0.03 X | = 0.04 100X. nated from the polycrystalline nature of the particles. Equilibrium Pressure of Hydrogen Over Sodium Hydroxide-Metal Systems. Empirical tests in a number of labo- ratories have shown that some of the noble metals such as silver and gold, as well as copper and nickel, are sodium hy- although none of these metals possible containers for droxade, can be considered as adequate structural metal., Under a thermal gradient, mass transfer of these metals from hot to cold regicns of the apparatus occurs; this mass transfer 1s considerably decreased when a however, pressure of hydrogen is applied. It appears likely that reactions of the type Ni + 2NaOH <=—=Na,0'NiO + H, , in which the reverse reaction 1is represented by NiQ + [—[2 == Ni + H20 and N320 + HQO = NaOH may be responsible for these phenomena. Accordingly, an attempt has been made to determine the equilibrium pressure of hydrogen over several metal-sodium hydroxide systems as a function of temperature. The apparatus consisted of a quartz connected to a envelope which was a liquid-nitrogen through stopcocks, to mercury manometer, cold trap, and, a vacuum pump and a gas-sampling bulb. A crucible of the metal to be tested containing 10 te 15 g of sodium hy- droxide was enclosed in the quartz envel ope., The apparatus was loaded with sodium hydroxide in a helium dry- box and was heated and evacuated to a pressure well below 1 mm. The temper- ature was maintained at a constant level until equilibrium pressures were obtained. In general, the pressure attained a value near equilibrium in 30 minutes. Once attained, equilibrium values remained constant for at least 7 hours. The values obtained i1n this pre- liminary attempt are shown in Table PERIOD ENDING SEPTEMBER 16, 1953 stable to sodium hydroxide than is either copper or nickel; the value obtained at 850°C for nickel appears to be unaccountably high. The pressure of hydrogen observed appeared to be independent of the gas volume of the system., Repeated removal of the hydrogen formed by pumping at the reaction temperature did not seem to change the equilibrivom pressure observed, although the rate of attainment of equilibrium was adversely affected. Since several evacuations should decompose about 1% of the sodium hydroxide charged, this constancy of pressure may indicate either that the pressure measurements are not sufficiently accurate to detect differences resulting from changes of this magnitude or that the three- component system (that is, Na,0-Ni-H,0) 1s univariant. If the system is actually univariant, in addition to the liquid, the gas, and the solid nickel phases, another solid phase, probably Na,0*NiO or NiO, is required, Since this reaction appears to be fundamental to the sodium hydroxide container problem, additional studies with higher accuracy will be performed in attempts to determine equilibrium pressures of hydrogen and the effect of added H,0, NiQ, and other possible impurities or corrosion products, Chemical Equilibria in Fused Salts (cf., sec 10), The results of a great many corrosion experiments have 6.5. It is apparent that gold is more suggested that conversion of chromium TABLE 6.5. HYDROGEN PRESSURES OVER SODIUM HYDROXIDE~-METAL SYSTEMS ' AT ELEVATED TEMPERATURE METAL MEASURED HYDROGEN PRESSURE (mm Hg) TESTED At 700°C At 750°C At 800°C At 850°C At 900°C At 1000°C Nickel 1 23 Gold 1 3 6 Copper 3 10 85 ANP QUARTERLY PROGRESS REPORT in Inconel to soluble chromium fluoride (probably a complex fluoride) by reaction with some oxidizing fluoride in the mixture occurs during high- temperature exposure, A typical reaction would be 2UF4 + CrmeFs + CrfF, in which all the fluorides exist primarily as complex species 1in dilute solution in Nak-ZrF, melts. Since the equilibrium constant for such a re- action would be temperature dependent, the slight “mass transfer” of chromium counld be explained on this Direct evaluation basis. of equilibrium constants of such a reaction can be obtained from analysis of the filtrate for soluble chromium 1f the following conditions are assumed to hold: (1) the physical solubility of chromium metal, as such, 1s very small 1in fluoride melts at high temperatures, (2) phase separation at high tempera- ture is possible, (3) materials are of known the starting quantity and adequate purity, and (4) the container is inert and sufficiently clean that no other reactions contributors, Considerable time has been devoted to the development of adequate equi- are 1mportant librium and filtration equipment. In a previous report,{1%) a description was given of an assembly fabricated from nickel which has been modified to include a retractable filter stick, which promises to be simpler to fabri- cate and easier to use, The filter stick uses a sintered nickel filter and 1s connected through a metal bellows to the melt container so that the filter 1s above the melt during the equilibration period and 1s im- mersed only during collection of the liquid sample. (lg)J. D. Redman and L. Quar. Prog. Rep. Pec. 10, 1952, ORNL-1439, p. 120. 86 G. Overholser, ANP In preliminary experiments, the materials were charged into the hydrogen-fired nickel equipment, equilibrated at 800°C for 5 hr with agitation (the assembly was shaken), and allowed to equilibrate at 600°C before filtration, The data obtained show that Na,CrF, 1is quite soluble in the fluoride mixture at 600°C. 1In most cases, that the ma- terial was completely dissolved at the concentrations used. This behavior is to be contrasted with the relative insolubility of K,NaCr¥, in LiF-NaF-KF mixtures., Petrographic and x-ray examinations of the filtrate indicate the existence of solid solutions of Na,CrF, in the NaZrF, matrix over a wide range of compositions, Although analyses for zirconium and uranium 1in these systems show slight variation, it appears there 1s no evidence that insoluble compounds of these materials result from N33CPFG addition. In the filtrates from the Na,FeF, addition, the lack of discrete phases of iron compounds and the similar lowering of refractive index indicate that this conmpound also forms solad solutions., However, the solubility of Na,FeF, in this mixture is considerably lower than that of Na,Fel_ ., and the analytical results are less certain. Preliminary experiments have been performed in which known quantities of metallic chromium or iron have been added to the NabF-ZrF -UF, mixture. Again, equilibration at 800°C followed by equilibration at 600°Cand filtration was the general practice. The results indicate that reaction occurs to an extent which makes possible analysas of the products 1n these systems. Additional studies with more pure materials and refined analytical techniques will be conducted in the near future, PERIOD ENDING SEPTEMBER 10, 1953 7. METALLURGY AND CERAMICS W. D, Manly J. M. Warde Metallurgy Division High-conductivity radiator fins can be satisfactorily brazed to tubes of either a longitudinal or a transverse array by using a low-melting-point Nicrobraz, Since it has been found that placing the brazing alloy on the brazed joint by the use of wire or washers 1s advantageous, a technigue for the preparation of low-melting- point Nicrobraz wire through the use of an acryloic cement was developed. A new brazing alloy series (nickel- phesphorus and nickel-chromium- phosphorus) has been found and is now being studied. Phosphorus additions to nickel and nickel-chromium alloys are extremely effective in lowering the melting point, and phosphorus 1is suitable for reactor usage because 1t hasa low cross section for the absorp- tion of thermal neutrons. Another advantage of the nickel-phosphorus- chromium type of brazing alloy is the possibility of preplacing the alloy by convenient plating techniques, since tubing can be plated with a nickel strike followed by nickel- phosphorus coatingby the ‘‘electroless” plating technique followed by chromium plating on top of the nickel-phosphorus alloy. It was found that addition of chromium to the basic nickel-phosphorus alloys is advantageous because it increases the resistance of the brazing alloy to air oxidation and sodium corrosion and gives more strength to the brazed joints. ' In actual radiator fabrication, it has been found that Nicrobraz i1s not so good a brazing alloy to use as G-E brazingalloy No. 62, a nickel-chromium~ silicon alley. During fabrication, it is best to bring the heat exchanger to 900°C so that the fin and tube temperatures will equalize, quickly heat 1t to the brazing temperature, hold that temperature for the brazing time, and then furnace cool 1t. Nicrobraz 1s not satisfactory for this cycle because the boron diffuses out of the alloy while it 1s being held at 900°C, and the resulting alloy has a much higher melting temperature than that of Nicrobraz., The nickel-chromium- silicon brazing alloy in undergoing the same thermal treatment produces a good braze. The cracking tendencies of the nickel-chromium-silicon alloy when applied to a test radiator have been determined by thermally cycling the entire assembly from 1000°C to room temperature. Only one small leak occurred after a severe water quench from 1000°C, The drawing of tubular fuel elements has been initiated. Twelve tubes con- taining cores of type 302 stainless steel, iron, and nickel with 20 and 30% UO, are being reduced by plug drawing from a tube 0.750 in. in diameter with a 0.042-1n. wall to a tube 0.250 in. in diameter with a 0.015-1n, wall. To give copper the oxidation re- sistance needed for its use as a high- conductivity fin, plates of chromium and nickel deposited by the thermal decomposition of the carbonyls were tried, but the experiments were not successful. Copper clad with Inconel and types 310 and 446 stainless steel has been studied. The type 310 and the type 446 stainless steel cladding seem to offer oxide protection and to be free of the serious intradiffusion experienced with the Inconel-clad copper. The creep and stress-rupture meas- urements for both coarse- and fine- grained Inconel in fluoride fuels at 815°C have been completed for the stress range of 2500 to 7500 psi. Specimens have now been i1n test for 1000 hr at 1500 and 2000 psi. A graph 87 ANP QUARTERLY PROGRESS REPORT 1s presented which summarizes the current test data. In the study of the combustion of sodium and sodium alloys,it has been shown that additions of mercury of about 70 mole % have a pronounced retarding effect on the combustion of sodium. In an attempt to develop suitable high-temperature-fluoride seals for pumps, packing components have been fabricated from both hot-pressed com- pacts and vitreous materials. Mixtures of copper and stainless steel with MoS, have been fabricated into cylinders to produce rings for packing-gland seals on a 2 1/2-in.-dia pump shaft. In addition, a Bel,-KF-Mglk,-AlF, com- position has been prepared for testing of a viscous seal. WELDING AND BRAZING RESEARCH P. Patriarca G. M. Slaughter Metallurgy Division Low-Felting-Point Nicrobraz., One of the metallurgical problems as- sociated with the use of high-con- ductivity fin materials for heat exchangers has been the selection of suitable brazing alloys. In order to satisfy the corrosion requirements of resistance to both sodium and air at 816°C, the search fora suitable brazing alloy was confined to alloys with a nickel-chromium base. Since the melting point of the copper coreof the radiator fin would be about 1083°C, the maximum suitable brazing alloy flow temperature was chosen as 1050°C. A nickel- chromium~iron-silicon-boron brazing alloy manufactured by the Wall Colmonoy Corporation and known as ‘““Low~Melting Nicrobraz” (LMNB) was found to have suitable flowability on Inconel-clad copper radiator fins when brazed in ~70°F dew-point hydrogen at 1050°C, A nominal composition of this alloy 1s given in Table 7.1, along with that of standard Nicrobraz (NB). As may be seen, the basic difference 1s in the 88 TABLE 7.1. COMPOSITION OF NICRORRAZ AND “LO%-MELTING NICROBRAZ” COMPOSITION (nominal =t %) CONSTITUENT LMNB (flow point, (flow point, 185@°C) 1156°C) Nickel 80 70 Chromium 5 15 Iron 5 5 Silicon 5 5 Boron 5 5 nickel and chromium contents of the two alloys. The techniques used for fabrication of IMNB wire are of general interest, since they are applicable to a large group of brazing alloys which must be initially prepared as a powder. The procedure used 1is, briefly, the fol- lowing: (1) 50 g of alloy powder, preferably ~100 to -200 mesh, 1is mixed with approximately 25 cm® of Bohm and Haas Acryloid B-7 to provide a thick homogeneous slurry; (2) excess binder 1s driven off by baking for 10 min at 200°C to a hard cake; (3) this com- posite 1s softened with a small quantity of acetone and kneaded to the con- sistency of putty; (4) the billet thus formed 1s extruded on a hydraulic press through a 0,060-1in.-ID Lavite die and wound ona 3/16-in.-dia mandrel rotated at approximately 60 rpm by a variable-speed motor; (5) the mandrel is then set aside for curing at room temperature for a minimum of 1 hr; (6) the brazing-alloy helix thus formed is removed from the mandrel and stored 1n l-in. over-all lengths until with a knife to supply individual rings. Brazing of Badiator Fins with LMNB. An attempt was made to braze clad- copper fins with LMNB, and in order to Iimit the problems in this preliminary study, the following variables were arbitrarily fixed: (1) The fin matertal would be Inconel-clad copper, cut that is, 10 mils of copper clad on both sides with 2 mils of Inconel. (The fabrication of this material by roll-cladding techniques 1s described below under “High Conductivity Metals for Radiators Fins,"”)(2) The 0.188-in. 0D, 0.020~in.-wall tubing would be fabricated of Inconel. (3) The service environment would include sodium at 1500°F within the tubes and air at 1500°F on the exterior of the tubes and on the fin surfaces. The two radiator configurations of interest were a longitudinal -fin design that required attachment of the fins parallel to the tube axis and a radial-fin design that consisted of the popular punched fins and a multitude of tube-to-fin joints. Fig. 7.1. Nicrobraz. on both sides with 2 mils of Inconel. PERIOD ENDING SEPTEMBER 10, 1953 The radial fin material was pre- pared by punching holes nominally 0,190 in. 1n inside diameter 1in the Inconel-clad copper fin; a lip was left to provide a I5-fin-per-inch spacing. Commercial Nicrobraz cement was used to firmly attach each pre- placed braze ring prior to brazing of the tube-to-fin assembly. A series of radial, Inconel-clad, copper fins brazed to an Inconel tube are shown 7.1, As can be seen, the copper core 1s adequately protected from attack by an oxidation-resistant brazing alloy fillet. Examination of the joints also revealed that no significant diffusion had occurred to af fect the properties of the core material. The dark lines delineating in Fig. Q.04 50X 0.02 0.03 t | 0.04 0.05 0.06 0.07 INCH 0.08 Section of Radial Fin-to-Tube Joint Brazed with Low-Melting-Point Tube material, Inconel; fin material, copper (10 mils thick) clad 50X, 89 ANP QUARTERLY PROGRESS REPORT the copper core were attributed partially to the presence of a heavy precipitate in the c¢lad material and partially to relief polishing. The precipitate is characteristic of the dilution and diffusion effects observed in nickel-base alloys and stainless steels brazed with the high~boron, high-silicon Nicrobraz alloys, To determine the feasibility of the longitudinal (delta) fin design, Inconel-clad copper sheet was preformed into 6-in. lengths of the triangular delta configuration., In order to braze this delta configuration to the tubes, a quantity of LMNB wire was formed in straight lengths and pre- placed along the full length of one side of each tube with Nicrobraz cement, A photomicrograph of the brazed assembly is shown in Fig. 7.2, The success of the brazing operations confirmed the ability of this cement to hold the brazing alloy wire in place until a temperature of 800 to i UNCLASSIFICD i v.9750 § TERMINAL ' B JOINT [ Fig. 7.2. Section of Longitudinal Fin-to-Tube Joint Brazed with Low-~ Melting-Point Nicrobraz and 8xidized for 500 hr at 800°C inm Air. Tube material, Inconel; fin material, copper (10 mils thick) cladon both sides with 2 mils of Inconel. 10X Reduced 36%, 90 900°C is reached, at which time sintering of the wire to the base material becomes pronounced. The most critical joint of the basic delta fin is that shown 1in the lower right corner of Fig. 7.2, where the sheared and therefore exposed edges of the fin mated with each other and the Inconel tube. Previous experi- ments had verified that LMNB did not flow on copper. Fortunately, pre- placing a LMNB wire on the inside and another on the outside of the delta fin effectively reduced the joint design to two I joints, and adequate protection of the fin edges was achieved. The effects of the 500-hr oxidation test at 1500°F on the copper core are evident from close examination of Fig. 7.2. The voids in the copper and the diffusion between core and cladding indicate that Inconel-clad copper is not entirely suitable for high-conductivity fins, at least not without diffusion barriers. This and other clad fins are discussed below under ‘‘High Conductivity Metals for Radiator Fins.” Nickel-Chromium-Phosplorus Brazing Alloys. The nickel-phosphorus and nickel-chromium-phosphorus systems are being studied to determine their suit- abilityas elevated-temperature brazing alloys., The evaluation of these systems 1s in preliminary stages, but the results indicate that phosphorus additions are extremely effective 1in lowering the melting point of nickel and nickel-chromium alloys. Phosphorus is especially suitable for reactor applications because of its low cross section for absorption of thermal neutrons. Alloys of the approximate compositions given in Table 7.2 have been prepared from a master alloy of 87% Ni-13% P. Stainless steel 1T joints brazed with these alloys exhibited excellent flow and wetting properties. anticipated, however, quite brittle. As was the joints were It is expected that TABLE 7.2. MELTING POINTS OF SEVERAL NICKEL~CHROMIUM-PHOSPHORUS BRAZING ALLOYS APPROXIMATE ALLOYS MELTING POINT o : (°c) 87% Ni1~13% P | 900 79% Ni-12% P-9% Cr 1000 72% Ni-11% P-17% Cr 1050 67% Ni-10% P-23% Cr 1100 diffusion treatments may improve the duccility of these alloys so that they will be comparable to the Nicrobraz and G-FE No. 62 alloys and will thereby retain the obvious advantage of a low melting point and a low cross section. A survey of the recent literature revealed a unique technique for pre- plating these alloys which involved an “electroless” method of nickel plating that was developed several years ago at the Bureau of Standards. The method requires, first, the deposition of a high-phosphorus-content nickel plate which then becomes the brazing alloy. Briefly, the deposition 1is effected by the reduction of a nickel salt to metallic nickel. A hypo- phosphite is converted to the phosphite with subsequent deposition of phos- phorus-containing nickel. An obvious advantage of this method is the possi- bility of depositing a uniform plate on a complex surface without ex- periencing the difficulties introduced by variations in ‘‘throwing power’ of the electrolytic methods. : To determine the feasibility of the use of this “‘electroless” plate as a brazing alloy, a series of type 316 stainless steel strips 1/2 by '3 by 1/16 were submitted to be “electroless” plated with a pominal thickness of 1 mil of nickel-phosphorus in a T-hr period by using procedures based on the method developed by the Bureau of Standards. Standard T joints in. PERIOD ENDING SFPTEMBER 10, 1953 were made for flowability tests by using a plated stainless steel strip against an unplated stainless steel strip. A photomicrograph of such a sectionis shown in Fig. 7.3. Fxcellent wetting was observed when the specimens were brazed in dry hydrogen at 900°C, which indicated that the reported approximate compositionof 90% Ni~10% P was correct. However, when the joint was subjected to an oxidation test of 500 hr at 1500°F in still air, some defections appeared, as shown in Fig. 7.4. The voids at the interface in the T are believed to be due to removal of constituents during pol- 1shing, whereas the undercut and the irregularities of the fillet surface are believed to be due to oxidation. Since i1t was felt that the oxidation resistanceof the “*electroless” nickel- phosphorus brazing alloy could be improved by alloy addition of chromium, a series of stainless steel test strips were prepared (by using con- ventional electrolytic plating methods) with approximately 0.2 mil of chromium on 1 mil of “electroless’” nickel- phosphorus alloy. These test strips were brazed at 1000°C, and they ex- hibited the same excellent flow as that observed with the ‘“‘electroless” nickel-phosphorus alloy. However, the oxidation resistance of the nickel- chromium-phosphorus alloy is better than that of the nickel phosphorus alloy. A photomicrograph of a section of the T joint after exposure for 500 hr to static air at 800°C is shown in Fig. 7.5. What appears to be voids in the brazing alloy matrix were found to be polishing pits, whereas the irregularities at the brazing alloy fillet surface were attributed to oxidation. Static tests were also run for 100 hr at 816°C to check the corrosion resistance of these T joints in sodium. Figure 7.6 shows a nickel- phosphorus braze after exposure; note that small voids are scattered through- out the fillet and across the joint. 91 ANP QUARTERLY PROGRESS REPORT UNCL ASSIFIED Y.9751 | 0.01 4 S ’,' c i i » o 2 0.02 .03 I Q & 0.04 ~ Fig. 7.3. “Electroless” Ni-P Brazed Type 316 Stainless Steel T Joint, As- Brazed. 100X, " UNCLASSIFIED | ¥.9758 ¢ o o 0.02 0.03 X Q z 0.04 Fig. 7.4. “Electroless” Ni-P Brazed Type 316 Stainless Steel T Joint, Tested in Air for 300 hr at 800°C. 100x. 92 PERIOD ENDING SEPTEMBER 10, 1953 ':) { i "",‘ N e Bev e L e ¥UEUNCLASSIFIED R Y.9759 ' b 0.02 0.03 X | & Z 0.04 Fig. 7.5. Chromium-Coated “Electroless” Ni-P Brazed Type 316 Stainless Steel Joint Tested in Air for 500 hr at 800°C. 100X, L U&CLASS#&Eéifi ; Y-9756 0‘0_1 » o o 0.02 0.03 T 2 z 0.04 Fig. 7.6. “Electroless” Ni-P Brazed Type 316 Stainless Steel T Joint Tested in Sodium for 100 hr at 816°C. 100X, ' 93 ANP QUARTERLY PROGRESS REPORT Figure 7.7, which shows a chromium- plated nickel-phosphorus brazed joint, reveals a few voids in the fillet to a depth of 3 to 4 mils., Tt may be concluded, then, that the nickel- phosphorus braze does not have satis- factory corrosion resistance to sodium but that the addition of chromium improves the resistance. The results of the investigation, to date, 1ndicate that the nickel- chromium-phosphorus system shows promise both when applied conventionally and when applied by ‘““electroless’ plate methods. Preliminary remelt tests indicate that remelt tempera- atures of 1250°C can be achieved by treatment at 1000°C for times as short at 10 min because of the relatively high diffusivity of phosphorus. The Fig. 7.7. Steel Joint Tested in Sodium for 100 hr at 81§°C. 04 physical tests conducted to date have been inconclusive, but there 1is evidence that ductility canbe improved by control of brazing temperature and time. Brazing of Radiator Assemblies with Nicrobraz. As has been indicated 1in previous reports, efforts to apply Nicrobraz as an elevated-temperature brazing alloy for fabrication of ANP sodium-to-air heat exchangers have been complicated by a number of factors. The results of numerous experiments conducted on laboratory-scale test specimens and on full-size heat ex- changer assemblies have shown that the rate of heating to the brazing tempera- ture is of great importance. Although early experiments indicated that diffi- culties in brazing did not arise from Y UNCLASSIFIED L o7 Y-9752 < - __;\‘\ . _ , N L 0.01 > o O Q.02 0.03 xI. QO z 0.04 Chromium-Coated ‘““Electroless’” Ni-P Brazed Type 316 Stainless 100X, di ffusion of constituents from the brazing alloy during heating, which would have resulted in a loss in flowabilityat the brazing temperature, later tests gave contrary indications. In tests with Nicrobraz in the form of flat washers prepared by stamping from sheet rolled or extruded from powder mixtures, it was found that during slow heating sufficient boron diffused from the brazing material to appreciably increase the melting point of the brazing alloy and thereby pre- vent the flow needed to produce a satisfactory joint. A remedy for the brazing diffi- culties seemed to be fast heating to the brazing temperature to minimize diffusion and the subsequent lack of flow. (The other possible alternative, that of increasing the brazing tempera- ture, was found to be i1neffective because dilution and undercutting alseo increased materiallywith temperature.) Bapid heating to the brazing tempera- ture was found to introduce associated problems. As would be expected, un- equal heating occurred, and the thin fin material reached the furnace temperature before the itubes did. 1In addition to an increase in distortion, the unequal heating resulted in “stealing” .of preplaced brazing alloy,. That is, the alloy would flow on the fin surface and would, consequently, be unavailable for adequate wetting of the tube wall when the temperature equalized, Since the remedy for this alternate effect would be a preheat and since preheating would permit diffusion, there appeared to be no remedy for the situation unless dif- fusion could be decreased. This was accomplished by the use of especially extruded Nicrobraz rings. FExperiments revealed that rings formed of 60-mil wire and preplaced were relatively unaffected by the rate of rise to temperature because of the relatively small contact area afforded for dif- fusion., The production of a large PERIOD ENDING SEPTEMBER 10, 1953 quantity of these rings was therefore initiated on a laboratory scale for future experiments and for possible use in a test heat exchanger. Brazing of Radiator Assemblies with G-E Alloy No. 62. Brazing evaluation tests were also conducted with the G-E brazing alloy No. 62 (69% Ni-20% Cr- 11% Si). FEach experiment revealed that this alloy was relatively free of the complicating factors that influence the behavior of Nicrobraz. The dilu- tion, diffusion, and heating-rate studies indicated that the boron-free G-E alloy could be preplaced in any convenient form and that preheat could be applied without impeding subsequent flow. Seven pounds of this alloy, as ~200 mesh powder, was obtained and is being used to study its suitability for brazing heat exchangers. A 1000-joint tube-to-fin radiator was fabricated with the G-E alloy preplaced, as a slurry, and subjected to the following brazing cycle: (1) heat to 900°C at an average of 50°C per minute; (2) hold at 900°C for 30 min to equalize fin and tube tempera- tures: (3) heat to 1150°C at an average rate of 30°C per minute; (4) hold at 1150°C for 60 min, furnace cool to black, and air cool to room temperature, Examination of this test radiator indicated that adequate flowand wetting had occurred. After repeated tempera- ture cycling, including a severe water quench from 1000°C, the radiator was found to have only one small leak, and that was subsequently repaired by re- brazing. HIGH CONDUCTIVITY METALS FOR RADIATOR FINS E. S. Bomar B, W, Johnson J. H. Coobs H. Inouye Metallurgy Division As indicated 1in the previous report, (1) a high-conductivity radiator (I)E. S, Bomar et el., ANP GQuar. June 10, 1953, ORNL-1536, p. 81, Prog. Rep. 95 ANP' QUARTERLY PROGRESS REPORT fin, with oxidation resistance at 1500°F, is expected to be realized by merely protecting a 10-mil copper fin by cladding 1t with one of several metals 1ncluding, among others, chromium, Inconel, and stainless steel. Good bonds have been obtained with Inconel and various stainless steels as claddings, but only the types 310 and 446 stainless steel claddings also give the desired oxidation protection withodut serious intradiffusion. : Vapor Plated CThiromium and Nickel on Copper. samples of copper plated by the thermal decomposition of the carbonyls of chromium and nickel were supplied by the Commonwealth Engi- necering Companyof Ohio. The plates de- posited were both codeposited and de- posited as separate layers of chromium on Several nickel on copper. Bright smooth deposits were limited to small areas, which were brittle and failed to protect copper from oxidizingat 1500°F, The plate thicknesses were between 0.2 and 1.6 mils. No further work on these materials 1s contemplated. Chromium-Plated Copper, Samples of ““ductile” chromium have been re- ceived from the Ductile Chrome Process Company. Cracking of the plate 1is not evident except by metallographic examination and oxidation tests, Alteration of the copper core does not seem to occur as a result of diffusion. Inconel-Clad Copper. Several square feet of clad material has been requested from commercial manufactuvrers for assembling experimental heat exchangers and for evaluation studies. Thus far, about 9 ft? of clad copper (10 mils of copper clad on both sides with 2 mils of Inconel) has been received. Cladding thicknesses varied between 0.75 and 3 mils, and numerous pinholes. Oxidation tests for 100 and 500 hr at 1500°C resulted in the eruption of numerous copper oxide nodules. Metallographic exami- nation showed the formation of voids there were 96 in the copper core near the interface and diffusion that affected a 2one of about 3 mils at the end of 100 hours. In 500 hr, the voids became fewer and larger, and they were distributed throughout the copper core. Diffusion seemed to extend across the entire cross section of the sample (14 mils). By suitable rolling and annealing, cladding variations can be minimized to about #0,5 mil and the pinholes can be reduced to an inconsequential number. Type 310 Stainless Steel-Clad Copper., The results obtained 1in previous tests of type 310 stainless steel-clad copper have been checked by running additional experiments. As was found i1n the first series of tests, diffusion appears to be limited to the formation of 1slands of a dark phase in the cladding to a depth of 0.5 mil. An increase in the testing time from 100 to 500 hr at 1500°F increased the number of these i1slands but not the depth of penetration. Surface failures were limited to a few copper oxide nodules. Type 446 Stainless Steel-Clad Copper. Type 446 stainless steel was roll clad onto copper by reducing 50% at 1000°C in several passes; no bonding was obtained. Good bonds were obtained by melting the copper into a type 446 stainless steel capsule under hydrogen. The clad composite was cold rolled with intermittent anneals to 8 mils, of which 4 mils was copper and 2 mils on each side was cladding. Oxidation tests for 100- and 500-hr periods resulted in the formation of a few nodules of copper oxide on the surface and severe warpage of the test pieces. Examination of the 1interface showed no apparent diffusion. In 500 hr, a dark precipitate, which is not continuons, forms at the interface to a depth of about 0.1 mil. Copper Clad with Copper-Aluminum Alloy. Ingots of 6% Al-94% Cu and 8% Al1-92% Cu have been made and rolled into a sheet. Attempts to roll clad the alloys onto copper resulted in bonding on only one side. This may have been due to the formation of A1,0;, during the welding. MECHANICAL PROPERTIES OF INCONEL R. B. Qliver K. W. Reber D. A. Douglas J. M. Woods C. W, Weaver Metallurgy Division Creep and Stress-Rupture Tests, A compilation of creep and stress- rupture data for both coarse- and fine- grained Inconel tested in NaF-ZrF, -UF, (46-50-4 mole %) at 815°C has been completed for the range of stresses from 2500 to 7500 psi, Additional specimens have been in test for 1000 hr at stresses of 1500.and 2000 psi. Data for fine-grained Inconel tested PERIOD ENDING SEPTEMBER 10, 1953 in argon at 815°C are complete for the stress range of 3500 to 7500 psi; other specimens have been in test at 1250 psi for about 1000 hours. Figure 7.8 summarizes the data on the Tnconel heat that is currently being tested, This new heat of Inconel sheet exhibits roughly twice the rupture life observed for the original heat tested in this program. The improved properties are thought to result from the larger amounts of the minor alloying elements present in this heat as compared with an earlier one. A series of tests with fine-grained Inconel specimens have been started at 704°C (1300°F) andat 899°C (1650°F). Tests are being run both in argon and in fuel. Tube -Burst Tests, A series of Inconel tube-burst testsat 815°C, with Bwe, 21113 CREEP RATE (% per hr) 5000 0.001 0.04 oX| 10 10,00 [ it - : - - 0000 | e LD LTI T T T ’ L 1 T ,~TIME TO RUPTURE FOR - 8,000 [rmememeofoo -ann-mm Q‘A- o/ FINE -GRAIN MATERIAL = | el e i Pl IN ARGON 7,000 |- s e | ,,,,, J rrrrr 42 ‘ ;<’ ;Hi ....... | | ] 5000 ~mw~—mmranwm~w¢4¢uu~m~w~m s BRI R R ST TOTAL STRAIN AT RUPTURE INDICATED 5,000 fsrmrmeprm e fo— et ( »»»»»»»» +~m~4—i -------------------------------------------------------------- ————————— 4,000 [ S e T S WA T e 2 L ¢ | @ Dol \\L i W 3,000 feomege o bttt AR L L L L e N P : | ¢ , B0 TIME TD RUPTURE FOR 8% 6% . FINE-GRAIN MATERIAL | s\ A i IN THE FLUORIDE -]\, N 2.000 - it A s L bbb b e ) e ' ’ LT Lt 3 | 1 N | Qfi\/ ;GQ{O : i 1 N\ ? | sy ‘ i ! TIME TO RUPTURE FOR // ‘ é@ P COARSE-GRAIN MATERIAL ‘ Q% . : | N THE FLUDRIDE - ‘ & ‘ | | ‘ oo | | | i I L ] Ll 1,000 Lo _— 1 A ] 5 . LLL 4 10 100 1,000 10,000 TIME YO RUPTURE {hr) Fig, 7.8. Creep and Stress-Rupture Data for 65-mil Inconel Sheet Tested in Argon and NaF-ZrF4-UF4 at 1500°F. 97 ANP QUARTERLY PROGRESS REPORT argon on both sides of the tube, has been completed in the range of tangen- tial stresses from 2000 to 4500 psi; the results are not consistent enough to allow interpretation. Another series of tests with fuel in the tube and argon on the outside is in progress; the rupture life appears to be roughly half that observed when the surfaces are exposed to argon only. From the tests to date, 1t was observed that not only is the rupture l1fe longest in air, but, the greatest elongations are observed when oxygen 1s present either in the gaseous atmosphere or as an oxide film. The shortest rupture life and the least ductility are observed for tests 1n hydrogen, while intermediate values are found for tests in other environments. It was also observed that the total elongation at rupture for specimens tested in the fuel 1is also, more regular and reproducible than that for tests in the other environ- ments. FABRICATION OF PUMP SEALS As discussed in Section 2, “Experi- mental Reactor Engineering,” a number of packed seal pumps have been operated with fluoride mixtures at 1300°F, Although operation with these seals has been encouraging, the leakage rates are higher than desired. Accordingly, a number of unique packed-seal materials ha ve fabricated, including vitreous seals and hot-pressed com- pacts with self-lubricating properties. Hot-Pressed Pump Seals (E. S, Bomar, J. H. Coobs, H. Inouye, Metal- Jurgy Division). Four additional cylinders of the 92% Cu-8% MoS, com- position were fabricated for testing as pump seals. These cylinders were 3 3/8 in. OD by 2 1/8 in. ID by 2 in. long, and a sufficient number of rings could be machined from them to obtain a packing gland to seal a 2 1/2 in.- dia pump shaft. The density of the cylinders averaged 96.0% of theoreti- cal. been 98 In addition, 1t was decided that the stainless steel-MoS, composition should be tested because the Cu-MoS, composition may not be sufficiently resistant to chemical attack for this application. As mentioned before, the components of the stainless steel-MoS, compacts react to some extent during the hot-pressing cycle; there 1is subsequent conversion of much of the austenite to ferrite, and a sulfide phase of unknown composition remains. Two cylinders, 1 11/16 in. OD by 1 3/16 in. ID by 2 long, were prepared by using -325 mesh type 304 stainless steel with 9% MoS,. These were fabricated by hot pressing in a graphite die at 1225°C. The density of the cylinders averaged 93.5% of theoretical. It 1s 1nteresting to note that, under identical conditions of temperature and pressure, straight stainless steel powder is consolidated to only 87% of theoretical density, Vitreous Seals (L. M. Doney, J. A. Griffin, J. R. Johnson, Metallurgy Division)., A limited experiment with a NaBeF, mixture in a pump sealing gland(?) indicated that such viscous fluorides might be developed for seal matertals. In the initial tests, a number of metal washers served to isolate rings of the fluoride. The ring voids were filled with the powdered fluoride mix. As a consequence of the encouraging results from the use of this fluoride, a program was initiated to develop more suitable high-tempera- ture viscous seal materials. The need 1s for a glassy substance with ap- propriate viscosities over the temper- ature range involved and also stability against devitrification. To provide a freely flowing viscous seal, 1t 1is believed that a viscosity gradient of 10?2 poises in the hot end to 10!° poises or higher in the cold end should be established. Softening of in, (2} June 10, B. McDonald et al., ANP Quar. Prog. Rep. 1953, OBNL-1536, p. 19. the whole seal should be carried out first so that wetting will take place on the shaft, housing, and metal spacers. Devitrification, or crystal- lization in the pglass, will lead to a fluid ceoentaining relatively hard crystals which will act as an abrasive on the metal parts. Beryllium fluoride isa glass former that produces a structural network of randomly oriented BeF, " jon tetrahedra that is analogous to the silica-glass network. Alkali and alkaline earth ions serve the same functions in the fluoride network as they do in the oxide systems. Very few data are available on the physical properties of the fluoride glasses; however, on the basis of similarity to oxide glasses, it is possible to predict their general behavior. Thus, for a glass with little tendency to de- vitrify and with reasonable stabilaty against atmospheric attack, Mg** or Ca'? and A1"*" ions should be inc luded. A glass which softens to a viscosity of the order of 10% poises at 250°C is desirable. The addition of alkali ions such as Na' or K¥ will produce the desired viscosity., The glass com- position proposed is the following: COMPOSITION (wt %) MATERIAL 50 BeF, 25 KF 16 MgF, 9 AIF, This glass was found to melt satis- factorily, and it did not devitrify in the handling operations. The glass was melted in platinum at about 900°C and cast into a graphite die to form the desired ring shape. It was not possible to cool the rings in air without cracking. By slow cooling in a blanket of glass wool, most cracking was avoided, and it can be PERIOD ENDING SEPTEMBER 10, 1953 eliminated entirely by furnace cooling and annealing. TUBULAR FUEL ELEMENTS E. 5. Bomar J. H. Coobs H. Inouye Metallurgy Division Preparation of tubular fuel elements by drawing has been initiated. Twelve tubes containing cores of type 302 stainless steel, 1ron, and nickel with 20 and 30% U0, are being reduced a total of 87% from 0.750 in. in diameter with a 0,042-1in. wall to 0.250 in. 1in diameter with a 0.015-1in. wall. BRe- ductien by plug drawing is being tried in preference to drawing on a mandrel because the latter process applies a shear stress to the core. The shear stress may have contributed to the failure during earlier experiments(s) of the core material in many of the tubes drawn at the Superior Tube Company. The plans call for the re- duction of six tubes on each of two schedules in steps of 15 and 20% re- duction per pass, respectively, The results obtained to date with the 15% schedule have been quite en- couraging. Six tubes have been processed through three steps of the schedule without difficuley. All six tubes have excellent inside and ocutside finishes, and they show no signs of “rippling’ or folding in the core region. However, the tubes being processed by the 20% schedule have given discouraging results. Three tubes have been processed for two steps, and all three failed in temnsion during drawing, one on the first pass and two on the second. In addition, slight rippling was evident at the leading end of the core, and, in some cases, drawing was accompanied by chattering in the die. Evidently, this schedule is fairly severe for (S)E. . Bomar and J. H. Coobs, Met. Quar. Prug. Rep. Jan. 31, 1%52, ORNL-1267, p. %3. 99 ANP QUARTERLY PROGRESS REPORT reduction of laminated tubes by plug drawang. However, it is planned to continue drawing the remalning tubes until complete failure or until the drawing 1s finished. INFLAMMABILITY OF S0DIUN ALLOYS G. P. Smith M. E. Steidlitz Metallurgy Division The new apparatus for examining the combustion of sodium and some of its alloys in various atmospheres has been completed., The equipment consists of a steel box connected to a vacuum pump, a filter and a source of dry air., The sodium capsule 1s heated to a temperature of 700 to 800°C in a furnace located on top of the box, The tip of the capsule is then broken off Lo cause a jet of molten metal to spray into the atmosphere in the box. Observations of the flammability are made visually. A total of six runs has been made with pure sodium at 800°C in the new apparatus. Of these, two were in dry air and four were in room air, all at a pressure of 1 atmosphere. The only observable difference i1n these tests was the formation of a heavier scum on system, 100 the surface of the molten sodium on the floor of the box after burning in room air, This scum was effective in confining the cowbustion on the floor to a section at the edge of the puddle. The dry air samples burned vigorously all over the puddle. A series of tests on sodium-mercury mixtures has been run at 700°C. Cap- sules containing 50, 60, 62, 64, 66, 68, 70, and 90 mole % mercury were burned in room air. The results were as follows: with 50 mole % mercury, the combustion of the jet is essentially as vigorous as with pure sodium. With 60 mole % mercury, the combustion 1is noticeably less vigorous. With 60 to 70 mole % mercury, the rate of com- bustion decreases rapidly until with 70 mole % and greater, no fire 1is observed in the jet, although a small amount of white smoke 1s formed. It is possible that the compound NaHg,, which nccurs with 66.7 mole % mercury, 1s 1mportant in this rather sudden change 1n combustibility with composition. NaHg, is the most stable of the sodium-mercury compounds. 1t has a melting point of 360°C and a heat of formation of about 18.3 kcal/mole, PERIOD ENDING SEPTEMBER 10, 1953 8. HEAT TRANSFER AND PHYSICAL PROPERTIES H. F. Poppendiek Reactor Experimental Engineering Division The heat capacity of the LiCl-KCl eutectic has been measured; in the solid state 1t was found to be 0,23 cal/g"°C, and in the liquid state, over the temperature range 351 to 840°C, it was found to be 0.32 cal/g"°C. The heat of fusion for this material was approximately 64 cal/g. The heat capacity of NaF-KF-LiF (11,5-42-46.5 mole %) was determined to be 0.41 cal/g*°C in the solid state and 0.45 cal/g-°C in the liquid state over the temperature range 475 to 875°C. The heat of fusion for this material was about 93 cal/g. Preliminary viscosity measurements have been obtained for two compositions in the NaF-ZrF, -UF, system, that is, 53-43-4 mole % and 53.5-40-6.5 mole %. These data were found to be very similar to those previously obtained for the 50-46-4 mole % mixture; for the viscosity of the 6.5 mole % UF, composition varied from about 16 c¢p at 580°C to 5.7 cp at 950°C. A preliminary study of NaF- KF-UF, (46.5-26-27.5 mole %) indicated that the viscosity varied from about 30 cpat 600°C to about 12 cp at 800°C. example, The density of NaF-ZrF4-UF4 (53.5- 40,0-6.5 mole %) has been determined over the temperature range 600 to 800°C. The densities of the pump seal materials PeF,, NaBeF,, and NairF, have been obtained over the tempera- ture range ~30 to 130°C., The density of NaF-ZrF,-UF, (50-46-4 mole %) at room temperature was also determined. The thermal conductance of an in- sulated safety-rod sleeve and the thermal conductivity of the diatomaceous earth insulation were measured; the conductance was found to be 2.6 Btu/hr* ft*°F, and the thermal con- ductivity of the insulation with a density of 29 1b/ft3 was 0.057 Btu/hr* ft* °F. : Measurements of the vapor pressures of three compositions in the NaF-Zr¥F, binary system have been completed. While the vapor pressure rises quite steeply with the ZrF, concentration, the previously repeorted values were considerably high, A summary of the vapor pressures of several ZrF, -bearing fused salts is presented. The analysis of all the heat trans- fer data on NaF-KF-LiF eutectic in Inconel has been completed. The re- sults are in agreement with the pre- liminary information previously re- ported, namely, that heat transfer data fall 50% lower than would be expected for the specific system being studied. This condition resulted because of the presence of a thin insoluble film composed mostly of K;CrF, at the inner tube wall. A series of wall to mixed-mean- fluid temperature differences have been measured in a forced-flow volume- heat-source experiment over a range of Reynolds and Prandtl moduli and a range of volume heat sources of from 0.13 to 0.42 kw/cm?®. The experimental temperature differences obtained were found te fall within *30% of the theoretical values, ENTHALPY AND HEAT CAPACITY OF HALIDES W. D. Powers G. C. Blalock Reactor Experimental Engineering Division The enthalpy and heat capacity of the LiCl-KCl eutectic (59 mole % LiCl) have been investigated with the Bunsen 101 ANP QUARTERLY PROGRESS REPORT ice calorimeter with the following results: (1) Hf(solid) - Hooc(solid) C P -4+ 0.23(6)T , 0.23(6) * 0.03, i for 97 to 351°C, H.(liquid) - HGOC(solid) C P 30 + 0.32(5)T , 0.32(5) £ 0.02 , for 351 to 840°C, where H is the enthalpy in cal/g, CP is the heat capacity in cal/g*°C, T is the temperature in °C. The LiCl-KCI (I)W. D. Powers and G, C, Blalock, Enthalpy and Heat Capacity of Lithiur Chloride, Chloride Eutectic, ORNL CF-53-8-~30 (Aug. It Potassiunm 5, 1953). enthalpy data are shown in Fig. 8.1. The heat of fusion for this material is about 64 cal/g. The enthalpy and the heat capacity of NaF-KF-11F (11.5-42.0-46,5 mole %) are(?) .HT(solid) - HUOC(solid) = 0.41T - 44 , C =0.41 £ 0.05 P ¥ for 300 to 455°C, H_(liquid) - Hooc(solid) =0.45T + 32, Cp = 0.45 * 0,03 , for 475 to 875°C, (2)W.[L Powers and G. C., Blalock, Heat Capacity of Fuel Composition No. 12, ORNL CF-53-7-200 (July 31, 1953). UNCLASSIFIED DwWG.21114 { 300 _ — Ry Q. & 0?/’ € @ CAPSULE 12 oii;p;:“ A CAPSULE 43 :g.&.a, | Q O CAPSULE 44 B E L0 e b 8 L o o S X 3':.\ ________ i 100 0 0 100 200 300 400 500 600 700 800 200 TEMPERATURE (°C) Fig. 8.1. 102 Temperatare-Enthalpy Relationship of the LiCI-XC1 Entectic. The heat of fusion of this fluoride mixture was found to be about 93 cal/g, VISCOSITY OF FLUORIDES S. I. Cohen Tn N. Reactor Experimental Engineering Division Jones Preliminary viscosity measurements have been obtained(3®) on either the Brookfield or the efflux viscometer ) (3)g, 1 . Cohen and T. N. Jones, Preliminary Measurements of the Density and Viscosity of Fluoride Mixture No. 40, ORNL CF-53-7-125 (July 23, 1953). (4)J M. Cisar et al., ANP Quar. Prog. Rep. June 10, 1952, ORNL-12%24, p. 146. 20 PERIOD ENDING SEPTEMBER 10, 1953 or on both, for two recently evolved fluoride mixtures, NaF-ZrF, -UF, (53-43-4 mole %) and NaF-ZrF, -UF, (53.5-40-6.5 mole %). These data are plotted in Fig. 8.2, together with the viscosity of NaF-ZrF, -UF, (50-46-4 mole %) for comparison., The viscosities of the three mixtures are quite similar, as would be expected because of their similar chemical compositions. In particular, the viscosity of the 53.5-40-4.6 mole % mixture, the mostl recent ARE fuel composition, ranged from about 16 cp at 580°C to 5.7 cp at 950°C, DWG. 21115 900 950 1000 1050 1100 TEMPERATURE (°K) t0 9 8 a L7 > FLUORIDE | COMPOSITION (mole %) £ 6 || MIXTURE | URg Zrfg NaF & NO.44 @ | 65 40 535 ) . = °l|no40om | 4 43 53 4 NO.30 & 4 46 50 3 2 750 800 850 Fig. 8.2. Viscosity of Several Compositions in the NaF-ZrF4—UF4 Systems, 103 ANP QUARTERLY PROGRESS REPORT Some preliminary measurements of NaF-KF-UF, (46.5-26.0-27.5 mole %) have also been made., The viscosity ranged from about 30 cp at 600°C to about 12 cp at 800°C. DENSITY OF EFLUORIDES S. I. Cohen T. N. Jones Reactor Experimental Engineering Division The density of NaF-ZrF, -UF, (53.5- 40-6.5 mole %) has been determined by the displacement method over the temperature range 600 to 800°C, and 1is best represented by the equation e = 4.06 - 0.00097T , where ©0 is the density in g/cm® and T is the temperature in °C, Determinations of the solid densities of BeF,, NaBeF,, and NaZr¥F_, all of which have been used studies, in pump seal have been made.(%) Figure 8.3 presents the density-temperature data of NaBeF,. The density of NaF-ZrF, -UF, (50-46-4 mole %) at room temperature (30°C) was found to be 4.09 g/cm?. (S)S. 1. Cohen and T. N. Jones, Measurements of the Solid Densities of Fluoride Mixzture No, 30, Bef,and NaBeF,, ORNL CF-53-7-126 (July 23, 1953). UNCLASSIFIED DWG. 20549A """"—'—'T—_——|—174'/"""' Tp T e ! ! i | ' | : ! | oo . - S i e — 2430 . L o1 L VN plogfem®)=246-0.000267 | N ) | B=t1x10"%C"" i 2.420 |— T DENSITY OF NoBeF, (g/em?) o410 L L ; b ~ -20 0 20 40 60 80 100 120 {40 TEMPERATURE (°C) Fig. 8.3. Temperature. Density of NaBeF3 vs, 104 ELECTRICAL CONDUCTIVITY OF LIQUIDS N. D, Greene Reactor Experimental Engineering Divisaion An experimental study of the elec- trical conductivity of molten salts has been initiated. A beryllium oxide conductivity cell has been fabricated and used 1n some preliminary high- temperature electrical conductivity experiments with molten potassium chloride. Also, the conductivities of concentrated sulfuric acid solutions (60 and 100% by weight) have determined as a function of tempera- been ture. These solutions may be used 1in volume-heat-source heat transfer experiments, THERMAL CONBDUCTIVITY The thermal conductivity of demsely packed diatomaceous earth insulation has been determined by measuring the thermal containing the material, These measure- ments indicate a thermal conductivity of 0.057 Btu/hr-ft<°F when the dia- tomaceous earth 1s packed to a density of 29 1b/ft3, In addition, development of both the longitudinal and the flat- plate thermal conductivity devices for accurate, high-temperature measure- ments has continued. Diatomaceous Earth (M. W. Rosenthal, J. Lones, Reactor Experimental Engi- neering Division), An experiment was performed to determine the thermal conductivity of diatomaceous earth Tesistance across an annulus annulus walls of a insulation contained in the between the safety-rod sleeve. The diatomaceous eartn had been tamped into the sleeve annulus to increase its density to approximately 29 1b/ft?, The sleeve had an 1nside concentric difimfit@f Of 2-38 in-, an outside diameter of 3.00 1n., and a length of approximately 5 feet., The thermal resistance of the sleeve was determined by heating the inner surface with condensing steam and by removing heat at the outer surface with flowing Determination of the heat flow rate and the temperature difference across the wall permitted calculation of the resistance and, from that, the thermal conductivity of the insulation, The tube was divided inte three chambers by rubber stoppers, and dry steam at atmospheric pressure was supplied to each chamber, Measure- ments were based on heat flow from the center section; the other two chambers served as guard heaters to eliminate end effects., Various leadsinto the system and between chambers provided for steam flow, condensate removal, venting of noncondensing gases, and temperature measurements, Jhe tube was centered in a length of standard 3 1/2-in. pipe, and water was passed through the anpulus between the pipe and the tube at a velocity of about 1 fps. The resistances of the steam-side, the water-side, and the metal walls of the tube were negligible compared with water, the resistance of the insulation. Hence, the temperature difference between the condensing steam and the water was essentially equal to the temperature drop across the insulation. Although the steam temperature was measured to ensure that it was super- heated (and hence carrying no water vapor), the condensation temperature corresponding to atmospheric pressure was used as the steam-side temperature. The conductance of the sleeve per foot of length, ¢/LAt (where g is heat flow rate, L is length, and Af is temperature difference), was measured "at the mean temperature of the insu- lation, that is, 140°F. The values obtained are based on the middle 43 1in. of the length of the tube. From the measured conductance (2,6 Btu/hr*ft* °F), the thermal conductivity of the dia- tomaceous earth insulation was calcu- lated to be 0.057 Btu/hr:ft? (°F/fv). Some data on the thermal ductivity of diatomacecus earth powder of various densities were found in the cone=- PERIOD ENDING SEPTEMBER 10, 1953 literature(®) and are plotted, to- gether with the value obtained in this investigation, in Fig. 8.4. Note that the measured value of thermal con- ductivity falls approximately on a curve which was extrapolated from the literature values. UNCLASSIFIED DWG. 21118 0.06 / A MEASURED VALUE A E ® LITERATURE VALUES 5t 2 = MEAN TEMPERATURE, 140°F A S 4 Q e 2 /D O o O WS T g |- a4 £ //// ? 2 0.04 [ fwww;@r‘m- ------ Ll & = j/ 1 & 0.03 10 15 20 25 30 DENSITY (16/1t3) Fig. 8.4, Thermal Conductivity of Biatomaceous Earth, Development of Thermal Conductivity Measuring Devices (W. D. Powers, R. M. Burnett, S, J. Claiborne, Reactor Experimental Engineering Division). The flat-plate conductivity measuring devices have been used for checking values previously obtained by use of the variable gap or Deem apparatus. Some difficulty has been experienced in completely filling the cells of the flat-plate conductivity apparatus. An x~-ray technique has been devised for detecting void spaces 1in the con- ductivity cells. Additional checks have been made on the longitudinal thermal conductivity apparatus, The reported values for the thermal conductivity of type 316 stainless steel have been checked to within 5% between 150 and 250°C, It found necessary to redesign the guard heaters so that considerably higher temperatures could be obtained, was (6)G. B. Wilkes, Heat Insulation, p. 166, Wiley, New York, 1950. 105 ANP QUARTERLY PROGRESS REPORT YAPOR PRESSURES OF FLUORIDES H. E. Moore R. E. Traber Materials Chemistry Division Mcasurement of vapor pressure by the method of Rodebush and Dixon¢7? was completed for three compositions in the NaF-ZrF, binary system during this guarter, In addition, wmeasure- ments were completed on NaF-ZrF, -UF, (65-15-20 mole %). The preliminary values previously reported(®) appear to be considerably too high. The vapor pressure values for these compositions are collected in Table 8.1, together with those for ZrkF,, UF,, and the other ZrF,-bhearing mixtures examined so far. Inspection of the calculated values for vapor pressures at 900°C indicates a low (7) 26, 851 (1925). (a)R.IL Moore and R. E. Traber, ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p. 89. ¥. H. BRodebush and A, L. Dixon, Phys. Rev. activity of ZrF, 1in the melt in all the mixtures, In the binary system NaF-Zr¥F,, the vapor pressure rises quite steeply with increasing ZrF, concentration. FORCED-CONVECTION HEAT TRANSFER WITH NaF-KF-LiF EUTECTIC H. W. Hoffman J. Lones Reactor Experimental Engineering Division The analysis of the data from the heat transfer experiment in which the NaF-KF-LiF eutectic (11.5-42.0-46.5 mole %) was flowing in an Inconel tube has been completed, The results are presented in terms of the Colburn J=function in Fig. 8.5. These data corroborate the previously reported result(%+1%) that heat transfer with (9)H. W. Hoffman and J. Lones, ANP Quar. Prog. Rep. June 10, 1953, ORBNL-1556, p. 90. (IO)H.Wh fioffman, Preliminary Results on Flinak Heat Trensfer, ORNL CF-53-8-106 (Aug. 18, 1953). TABLE 8.1. VAPOR PRESSURE DATA FoOR Zr¥, -BEARING FUSED SALTS SALT COMPOSITION (meole %) VAPOR PRESSURE CONSTANTS* VAPOR PRESSURE T : ” AT 900°C NaF KF Zr¥, UF4 A B (mm Hg) 100 9,171 7.792 0.9 100 10,936 12.113 617 66.7 33.3 5,421 5.057 2.8 57 43 7,289 7.340 14 50 50 7,213 7.635 32 42.2 57.8 8,250 9.000 93 65 15 20 6,944 5.86 0.9 50 25 25 6,906 6.844 9 53 43 4 7,105 7.37 21 50 46 4 7,551 7.888 28 46 50 4 7,779 3.281 46 5 51 42 2 6,879 6.743 8 A *For equation: log Py yos " 7o) v B 106 the NaF~-KF-LiF eutectic in an Inconel system was 50% lower than that in a nickel system., It was found that this difference in heat transfer was caused by a thin, insoluble film that formed on the inside surface of the Inconel tubes., This film has been identified as K,CrF, plus some in- soluble phases of Li,CrF, that form when KF- and LiF-bearing flvorides come in contact with the Inconel., A film having a thermal resistance of 0.002 (Btu/hr*fe-°F)~!, for example, a film 1.2-mils thick with a thermal conductivity of 0.5 Btu/hr-ft-°F, could account for the observed disparity. ST 0010 ————— T g -+ PREDICTED CURVE 0005 - v (NG INTERFAGE RESISTANCE ) 2/:fi i 5t- /= EXPERIMENTAL CURVE 0002 ' 0001 e 1000 2000 5007 10,000 20000 Re 50000 100,060 Fig. 8.5. Heat Transfer with the NaF-KF-LiF Euntectic (11.5-42.0-46.5 mole %) inm an Inconel Tube,. The effect on the film formation of electrical current flow through the tube wall was investigated., A piece of Inconel tubing was suspended in a guiescent pot of molten eutectic for 24 hours. On removal of the tube, a film similar to the films previously noted was observed., A piece of nickel tubing tested under the same conditions showed no film. system for ob- taining fused-salt heat transfer coef- ficients The test section and mixing-pot assembly has been modified to enable easier replacement of the test section. Further attempts will be made to The experimental is now being reassembled, obtain heat transfer coefficients for the NaF-KF-1iF euntectic in turbulent PERIOD ENDING SEPTEMBER 10, 1953 flow within a nickel tube., Additional smal l-scale experiments will be under- taken with this eutectic with several other fluorides flowing through Inconel tubing to check the time and tempera- ture characteristics of any films formed. 1In addition, the thermal conductivity of the films and of pure K;CrFy, will be determined and com- pared, CIRCULATING-FUEL HEAT TRANSFER H. F. Poppendiek G. M. Winn Reactor Experimental Engineering Division A series of heat transfer measure- ments have been obtained from the forced-flow volume-heat-source experi- ment previously described.(!?) This system 1is shown in Fig. 8.6. The heat generated electrically within the flowing electrolyte in the test section is transferred to an antifreeze coolant in the heat exchangeér., Mixed-mean fluid temperatures, tube wall tempera- tures, fluid flow rates, and external heat losses were measured. The purpose of this experiment was to measure the radial temperature differences for a range of Reynolds and Prandtl moduli and to compare them with the calcu- lations made by using the theory which was previously developed, (%) Parameters were studied over the following ranges: 5,800 < Re < 14,000 , 4.6 < Pr < 8.7, 0.13 < W < 0.42 kw/cc , k = 0.30 Beu/hr* ft2 (°F/fc) | d = 9/32 in. A typical set of experimental tube- wall and mixed-mean-fluid temperature OBy F, poppendiek, G. Winn, and N. D. Greene, ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p. 92, (lz)H. F. Poppendiek and L. D. Palmer, Forced Convection Heat Transfer in Pipes with Volume Hent Sources Within the Fluids, ORNL-1395 (Dec. 2, 1952). 107 ANP QUARTERLY PROGRESS REPORT TEST SECTION ELECTROLYTE SUMP COOLANT TANKS <=7 Fig. 8.6. measurements, together with the con- ditions of the experiment, 1s shown in Fig. 8.7. A plot of the theoretical, dimensionless differences between the wall and mixed-mean-fluid temperatures as a function of Beynolds and Prandtl moduli for the case of an insulated pipe wall is shown in Fig. 8.8. Also plotted in Fig. 8.8 are the temperature differences obtained from the experi- ment described. The experimental temperature data fell within *30% of the predicted values, Currently, laminar flow experiments are being conducted, Comparisans between theory and experiment for this type of flow will also be made, 108 UNCLASSIFIED PHOTO 20588a POWER LEADS CONTROL & INSTRUMENT PANEL MIXING CHAMBERS FLOWMETER HEAT EXCHANGER e ELECTROLYTE PUMP Experimental) Forced-Fliow Yolume-Heat-Source System. UNCLASSIFIED DWG. 24118 120 S Re=5300 ‘ Pr=87 | | "o Wo=0H kwiemd T T o - k =030 BTU/hé-ff-"’F \r‘“*INSlDE TUBE WALL L 5 | ‘ \ TEMPERATURE rLE 100 . ad = /32 n e e - . \\\ . e o (=) preg = 7 °F i— Lt < o 1 ‘ A - ‘{‘G.TOF /.J) ul i /,/ = % Q0 B —- i ’L e T e \Tw ”m);—_:./ E— 1 | — |’ = | A ‘ et o 80 b———— & ;/./‘%Fg‘:i,,xi,ifi,i,, SN S ,_,_—"'r AN ! o } 3 MIXED-MEAN-FLULID ; TEMPERATURE 70 N [ ‘ | 0 i 2 3 4 5 6 7 DISTANCE FROM TEST SECTION ENTRANCE (in) Fig. 8.7. Experimental Tube-Wall and Mixed-Mean-Fluid Temperatures. UNCLASSIFIED DWG. 21119 — Al i {fo Qn}f(flk O/k) Fig. 8.8. Theoretical and Experi- mental Dimensionless Differences Be- tween Wall and Mixed-Mean Temperatures with Pipe Wall Insulated, BIFLUID HEAT TRANSFER EXPERIMENTS D, F. Salmon, ANP Division The bifluid loop with a concentric tube heat exchanger operated for approximately 500 hours. Heat was transferred from NaF-ZrF, -UF, (50-46-4 mole %) in the center tube to NaK in the annulus. All structural material in contact with the fluoride was Inconel, except for several type 316 stainless steel parts in the sump. The center tube was 0,135 1in. ID with a 0.025-in. wall andL/D ratio of 587. The fluoride Reynolds numbers ranged from 2000 to 3500. The inlet fluoride temperature was maintained at 1500°F, the axial temperature difference was 300 to 350°F, and the radial tempera- ture difference was 100 to 125°F. PERIOD ENDING SFPTEMBER 10, 1953 The fluoride heat transfer coef- ficient was obtained from the over- all coefficient by means of a Wilson Plot, which was described in the previous gquarterly report(13) for experiments with a larger diameter heat exchanger tube. Figure 8.9 shows the data for the tubes of both heat exchangers compared with data obtained from the Hausen equation. Except for the run at lowest fluoride flows, good agreement 1is obtained, Visual 1inspection of the center tube of the heat exchanger showed a metallic deposit on the wall that was approximately 10 mils thick near the cold end, This deposit was similar in appearance to the iron layer found in the first heat exchanger, but it was not so thick. The effect of the observed deposit was evident in the gradual reduction of the over-all heat transfer coef- ficient during the 19 days that the loop operated. The total reduction in the heat transfer coefficient was 16% over this period. It is interesting to note that a 10-wmil layer of pure iron inside the center tube would have a thermal resistance only one tenth that of the Inconel wall aud would make only a 2% change in the over-all coefficient. This would indicate that the observed deposit had a higher thermal resistance than a 10-mil iron layer. A metallurgical examination 1is being made to determine the extent of corrosion and mass transfer, The layer observed in the heat exchanger tube was undoubtedly transferred by the fluoride from the type 316 stain- less steel pump to the nickel heat exchanger tube (cf., sec. 6, ““Cor- rosion Research’). An Inconel pump is being fabricated to obtain a mono- metallic system in which galvanic mass transfer will not be possible. (13)0. F. Salwoen, ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p. 92. 109 ANP QUARTERLY PROGRESS REFORT UNCLASSIFIED 100 DWG. 21120 | T - EE T - BRE T - | // 5 ‘ i ] 1 | | | / A - : ‘ ; - L/D=A0 o\ / 1{ } l | ""1.—'[\// A ‘ ‘. | | / _ O HEAT EXCHANGER NO. 2, THIS REPORT. _ i A A HEAT EXCHANGER NO.{, ORNL-1556, p 92. ‘ maE ‘ —- ‘ "‘ | ‘ "‘ — — T~ - i o ; T | . | \ Q > i i i | |/_§ ! . | ‘ . $ ] | b ] e N ] I 1 i . r/Q/ | | ......... 2 - } 8 ‘ ™o L/D=587 ——- -— — - - Vo _ ] g ; | ‘ | | o | r ‘ | i | | \ | ! ! P g b | ‘ | ) l o : 100 1000 10,000 40,000 Re Fig. 8.92. Compzrison of Fluoride Heat Transfer Data with Hausen’s Equation. 110 9. PERIOD ENDING SEPTEMBER 10, 1953 RADIATION DAMAGE J. B. Trice, Solid State Division An Ju Miller. Additional studies were made on fuel capsules irradiated in the LITR and 1n the MTR., Improved remotve- handling techniques were devised for fuel sampling and weighing, and several innovations were made in the methods of material analyses. No evidence of gross segregation of the uranium 1in irradiated fuel was evident in this recent work. Examinations of the Inconel capsule walls showed the previously reported tendency toward intergranular corrosion which does not occur in unirradiated static tests, but it cannot be assumed from the evidence obtained to date that this is due to radiation damage per se. Design and construction of in-pile circulating fuel loops are progressing. The in- pile creep experiments performed to date in the LITR show no serious effect of radiation on the creep rate, and construction of the equipment for creep testing in the MTR is nearing completion, Further details of these and other radiation damage studies are reported below, and complete data will be available in the Solid State Division semiannual progress report for the period ending September 10, 1953, ORNI-1606. IRRADYATION OF FUSED MATERIALS G. W, Keilholtz F. R. Klean J. G, Morgan M. T. Robinson H. E. BRobertson A. Richt C. C. Webster W, B, Willis J. C. Pigg M. J. Feldman Solid State Division In the past, chemical examination of fuel irradiated in Inconel capsules in the LITR and in the MTR had indi- cated that the concentrationof uranium was not uniform. There 1s evidence that under the conditions of these early tests some zirconium tetra- ANP Division flueride was distilled from the fuel into the free-space-portion of the capsule and that the sampling procedure may have been inadequate, To avoid such uncertainties, a new type of MIR test capsule has been used which has only a small, high-tempera- ture, vapor space above the molten fuel. The sampling facilities have been 1mproved so.- that the helium atmospheres are more nearly pure and the remote weighing 13 more accurate, Facilities were also provided for examination of specific core sections of the salt by drilling successively larger sections of the fuel column from the capsule. Analyses of Irradiated #uel, Since the new techniques were initiated, five capsules containing NaF-ZrF, -UF, with various concentrations of UF, have been analyzed for segregation data. Four of these capsules were irradi- ated in the MTR at 1500°F, and one was a special control test with a similar thermal history. The samples taken from these capsules were analyzed by two methods: (1) chemical analyses with both potentiometric and polaro- graphic techniques {analyses made by the Analytical Chemistry Division), and (2) mass spectrometry by the isotope dilution technique (analyzed by the Stable Isotope Division). The data from these tests are presented in Table 9.1. 1In experiments in which several cores were taken from the capsule, those labelled A were from the center of the salt column, those labeled B were the next sample out from the center, etc., The data show some scatter but no evidence of gross segregation of uraniumin the irradiated fuels. 111 ANP QUARTERLY PROGRESS REPORT TABLE 9.1. ANALYSES OF IRRADIATED FUEL THERMAL FLUX POWER TIMF QF ) URANIUM CONTENT (%) ( ]HEOHETIQAL OBIGINéL PLE | (e F Y ec | DISSIPATION | IRRADIATION [— e . URANTUM CONTENT | URANIUM % IN FUEL OR CONTROL By By AFTER BURNUP | CONTENT « 10 ) (wates/em?) TEST (hr) Mass Spectrameter | Chemical Analysis (%) (%) 75 A 0.0 0.00 510 8.46 8.68 B.68 B 0.0 0.0 510 B.26 8.85 8.68 B.68 C 0.0 0.0 510 B8.67 8.69 B.68 B.68 D 0.0 0.0 510 8.75 8.86 B.68B B.6B E 0.0 0.0 510 8.37 8.71 B8.68 8.68 F 0.0 0.0 510 8.90 B.68 B.68 8.68 G 0.0 0.0 510 8.97 B.42 8.68 B.68 H 0.0 0.0 510 .31 8§.70 B.68 8.68 219 A 2.4 2300 330 £.96 6.50 7.55 8.68 B 2.4 2300 330 7.50C 7.55 B.68 C 2.4 2300 330 7.49 8.20 7.55 8.68 410 2.4 4400 419 11.57 11.20 11.12 13.3 501 2.4 G600 120 24.28 24,7 26.3 502 2.4 | 9600 273 22.97 23.4 26.3 Examination of Irradiated Fuel Containers., The Inconel capsules used in the MITR were unavailable for metal- lographic examination while the fuel- sampling work was 1in progress, but several capsules from LITR irradiations at 1500°F were examined, The results are shown in Table 9.2. The method of reporting corrosion in terms of depth of corresive pene- TABLE 9. 2. tration does not lend 1tself to an accurate evaluation of the amount of cCorrosion In general, the in this particular case. corrosionh 1n un- irradiated tests has been of a globular or spotty 1intergranular type, while the corrosion noted on the irradiated samples shows as a continuous network of intergranular attack, cannot be determined, Also, 1t at this time, RESULTS OF EXAMINATIOGN ©F IRRADTATED FUEKEL CONTAINERS IRRADTATION TIME PENETRATION OF METAL (mils) (hr) Selt Region Interface Vapor Region LITR Tests (230 watts/cm®) 53 0 to 0.5 None 140 1 1 1 270 1 0.5 None 565 2 to 3 l to 2 1 685 3 to 4 3 to 4 1 to 2 810 1 1 Control Tests (0 watts/cm®) 100 None None 300 1l to 2 1 520 1 1 700 1 112 whether the signs of increased cor- rosion in the irradiated specimens are due to radiation damage per se or to some extraneous effect, such as temperature control of the convection currents in the in-pile fuel capsule. IN-PILE CIRCULATING LOOPS 0. Sisman M. T. Morgan W, E. Brundage A, S. Olson Solid State Division For the past several months, a considerable amount of design and construction work has been done on an in-pile loop for circulating fluoride fuel. It is planned to operate the loop, 1initially, in the LITR to demenstrate a properly functioning system and to obtain test data at low flux; then, the final experiments will be performed in the MTR, The design of the loop is such that less than 1 ft of 1/4-in.-ID Inconel tubing is in the highest flux portion of the beam hole. The pumping rate 1is on the order of 10 fps so that turbulent flow is achieved, but, as a result, AT is small., As it leaves the high- flux region, the fluid will pass through a large-capacity heat exchanger, which can lower the temperature several hundred degrees Fahrenheit, and through the pump and an electrical heating system, which can bring the fuel to almost 1500°F before it is returned to the high-flux region. Originally, a small in-pile packed-sealed cen- trifugal pump was to be used in these experiments, but i1t now appears that the only pump which can be available in the near future gas-sealed centrifugal pump which must be positioned outside the reactor shield., This arrangement requires a large volume of fuel with a corres- made is a pondingly high dilution factor, a large amount of external shielding, and the handling of a 15-ft radiocactive sectien of the loop. PERIOD ENDING SEPTEMBER 190, 1953 SODIUM-BERYLLIUM OXIDE STABILITY TEST F. M. Blacksher C. Ellis W. E. Brundage M. T. Morgan R. M, Carroll W, W. Parkinson 0. Sisman Solid State Division The beryllium oxide moderator in the ABRE will be exposed to confined or slowly flowing sodium. To determine the stability of the beryllium oxide with respect to the sodium coolant in the radiation field of the reactor, an experiment under static conditions has been carried out in the LITR, Peryllium oxide specimen blocks 1/4 by 3/16 by 1 in. cut from various regions of the moderator blocks for the ARE to give samples of representative densities., These specimens were heated to 825°C in a vacuum furnace until the weights were constant. FEach specimen was then placed in a stainless steel capsule, 2 1/4 by 9/16 in., and a retaining spring was welded in the capsule to hold the beryllium oxide beneath the surface of the sodium. About 2 cm® of sodium was charged into the capsule 1in a helium-filled dry box and the capsule was welded closed, were Eight filled capsules were sealed in a can fitted with heaters, thermo- couple leads, and inert-gas tubes, as shown in Fig. 9.1, The capsule assembly was attached to a plug which was inserted 1n the water-~-cooled reactor hole liner and placed in hole HB-2 of the LITR. The irradiation was carried out for 328 hr at 1500°F, 45 hr at 1300°F, and 110 hr at 750°F, Because of difficulties with the thermocouples, the temperature values are only approximate, The same heat- treating schedule was duplicated on eight unirradiated capsules prepared in the same manner as those which were irradiated. The capsules were opened (the irradiated ones in a hot cell), the 113 ANP QUARTERLY SO0DIUM ——- 4-TERMINAL KOVAR SEAL PROGRESS REPORT BeQ—"" SECTION SILICA WOOL AND TAPE INSULATION DWG. 21121 CAPSULE HEATERS BB CAPSULE CONTAINER T —— werocoun s e ——— INNER HEATER / SUPPQ Fig, 114 9.1. \WIRE SUPPORT_| CFRAME | ] L _ SODIUM— OUTER HEATER Capsules for Na- Bel Irradiation Stability Test. beryllium oxide specimens were re- moved, and the sodium charges were dissolved in ethanol orv mixtures, The empty capsules were leached in 15% NaOH for 48 te 72 hr to dissolve the beryllium oxide on the walls, The solutions, both leach and sodium for each capsule, were analyzed for beryllium by the Analytical Chemistry Division. ethanol-water The beryllium oxide specimens were protected from the atmosphere, during and after removal from the capsules, by immersion under oil. The sodium and o0il were removed from the speci- mens by heating them to 800 to 825°C for about 1 hr under vacuum-a cleaning treatment similar to that used prior to insertion of the specimens in the capsules. The specimens were weighed, and, with one exception, they were found to have slightly (0.2%) gained in weight, To remove any possible sodium or Na,O which might not have been removed by the vacuum heating, the specimens were soaked in water for seven to ten days, heated ir vacuum again, and reweighed. An average weight loss of 0.0005 g was found for both the irradiated and the unirradiated specimens, without differentiation, and all weights were within 20.002 ¢ (0.1%) of the original weight. The results of the chemical analyses confirm the absence of ‘corrosion. In all the solutions, both those of the sodium and of the capsule leachings, the beryllium content was below the limit of sensitivity of the analytical method, 0.0001 g total beryllium per sample, or less than 50 ppm in the sodium bath. Visual inspection of the beryllium oxide specimens showed no change between the irradiated and unirradiated material and neither seemed to have sutffered by the heat treatment. The surfaces of all the specimens remained smooth, and the corners were sharp. No evidence of cracks or pits could be PERIOD ENDING SEPTEMBER 10, 1953 found, The accumulated exposure was about 2 x 10!° nvt thermal and greater than 10'® fast flux, CREEP UNDER IRRADIATION W. W, Davis J. C. Wilson J. C. Zukas Solid State Division The work during this report period has been concentrated on the determi- nation of the effect of inert atmos- pheres on the creep rate of Inconel and the temperature dependence of the creep rate during irradiation. {ne test has been run at the stress and temperature of the ARE pressure vessel, The results reported from LITR and Graphite Reactor cantilever tests give no indication that serious effects of irradiation on the creep strength of Inconel may be expected, although tests in the higher flux of the MTR and further tests in inert atmospheres will be required to guarantee the strength of airecraft reactor structures, A cantilever creep test of Inconel was run at 1300°F at a stress of 6000 psi in the Graphite Reactor for 600 hr to approximate the conditions of the ABE pressure shell, The corresponding bench test has not vet been run. The creep strain was less than 0,08% at 600 hr and the secondary creep rate was about 0.05%/1000 hr. For com- parison, International Nickel Company data for hot-rolled plate give 6200 psi as the stress required to give a creep rate of 0.1%/1000 hr at 1300°F, The effect on creep strength of the decarburized layer normally existing on Inconel sheet i1s of considerable interest, since the cantilever test would be peculiarly sensitive to any changes in strength of surface lavyers relative to the bulk of the metals. A series of cantilever beams (sguare an cross section) was machined from the Inconel plate from which all ia-~-pile test specimens have been made. JIn one 115 ANP QUARTERLY PROGRESS REPORT pair of test bars, the original sur- faces of the plate were used as the tension and compression faces of the beam, and another pair was oriented with the plate surface at the sides of the beam; the in-pile test specimens have been made in the latter manner. In bench tests at 3000 psi and 1500°F, no difference ($10%) as a result of orientation was found in the creep behavior, Metallographic determi- nation of the degree of decarburization in this plate has not yet been made, but the tests showed that the creep behavior is not influenced by orien- tation of the specimen surfaces, The effects of temperature during the first few hundred hours of operation on the creep rate of Inconel at 3000 psi are shown in Fig. 9.2. band represents the range of values the LITR 1in the lower band summarizes the bench tests in air. Also, two points are shown for a helium bench test and an LITR test Despite a fair amount of scatter in the results, 1t 1s apparent that the temperature dependence of the creep rate is not The upper for tests conducted 1in helium; in air, seriously affected by the wide variety of test conditions. A reference slope is shown that was taken from Inter- national Nickel Company data for hot- rolled Inconel. A number of other creep rate determinations in air, both on the bench and in-pile, show that the temperature dependence of the creep rate 1S not sensitive to test con- ditions over the range of variables tested, Some pre- and postirradiation tests of constant-strength, cantilever-beam Inconel creep specimens have been run. The specimens were irradiated at room temperature in the LITR. The purpose of the work was to determine whether there was a real creep reducing effect of irradiation, to determine the kinetics of annealing the irradiated specimens, and to determine whether annealing during testing would result 116 SSD-A-813 DWG. 20926A 100 CREEP RATE (arbitrary units} LITR TEST IN HELIUM BENCH TEST IN HELIUM LITR TEST IN AIR L BENCH TEST IN AIR [ S LITR FAST FLUX >10'2 — 1650 1600 1550 1500 1450 TEMPERATURE (°F ON — SCALE) rubs Fig. 2.2. Temperature Dependence of the Creep Rate at 3000 psiof Inconel Irradiated in the LITR. in temporary increases 1n creep rate, such as those observed when cold- worked specimens are annealed, (1? Data obtained at 1500°F and 3000 ps1 were presented previously,(?? and it was noted that the irradiated speci- mens had a lower creep rate during the early part of the extension than they did later. Another test at 1800 ps1 and the same temperature gave the same results. Tests at temperatures down to 1300°F and at stresses designed to keep the creep rate approximately (I)J. N. Greenwood and H. K. Worner, J. Inst. Metals 64, 135 (1939). (2)y, ¢. Wilson, J. C. Zukas, and W. W. Davis, Prog. Rep. June 10, 1953, ORNL-1556, constant at each temperature are being The metal used for this work has a decarburized surface layer as a result of mill-annealing practice. Some material now available 1s of sufficient thickness to permit the decarburized layer to be machined off run., before testing. Figure 9.3 shows the creep rate vs, temperature relationship for type 347 stainless steel in bench, LITR, and Graphite Reactor tests in air. No major effects of irradiation are apparent. A tensile type of creep apparatus is being fabricated for use in the MTR, and it 1is expected that 1t will be inserted in the MTR in the next several weeks. PERIOD ENDING SEPTEMBER 10, 1953 S5D-B-814 DWG. 203974 1000 , ------ CORRESPONDING -} - BENCH TEST o | e ORNL GRAPHITE 3NN NG REACTOR - LTRFAST FLUX >wofe __ . 7 =7 GRAPHITE REACTOR FAST FlLUX =4 xi0' CREEP RATE {arbitrary units) CORRESPONDING BENCH TEST " _______________________ k,,‘L, A — oo I T 1600 1500 1400 , 1300 1200 TEMPERATURE (°F CN T SCALE) qaes Fig. 9.3. Temperature Dependence of the Creep Rate 0f Irradiated Type 347 Stainless Steel. 117 ANP QUARTERLY PROGRESS REPORT 10. ANALYTICAL STUDIES OF REACTOR MATERIALS C. D. Suysano, Analytical Chemistry Divisian J. M. Warde, Metallurgy Division Developmental work has been com- pleted on the volumetric determination salt mix- a more rapid and of zirconium in fluoride tures.(1) However, equally precise method is now being investigated that involves the appli- cation of differential spectro- photometry to the zirconium-alizarin red-S complex, Methods for determining the con- centrations of the reactants products of the reaction CrF, + UF, == Cr° + UF, were investigated, A solution of di- and sodium dihydrogen ethylenediamine- tetraacetic acid (EDTA) buffered at pH 4.0 with sodium acetate and acetic acid was used to selectively leach CrF; from a medium of NaZrF,. The rate of complexing of Cr(III) with EDTA was slow, after several but it was quantitative hours of stirring. Chromium metal and the trivalent oxide Cr203 were insoluble in EDTA; the divalent fluoride was soluble and formed the Cr{(II)-EDTA complex 1in a relatively short time. Uranium tetra- fluoride was shown to be very soluble in this medium; uranium trifluoride was only slightly soluble, A solution of EDTA buffered at pH 6.8 showed little solubility effect on UF,; no change in solubility with UF, was observed. It is likely that EDTA can be used to leach trivalent chromium and tetra- valent uranium selectively from NaZrF_. The major portion of the experi- mental work with reactions involving the use of bromine trifluoride for the determination of oxygen 1n metallic oxides was confined to alterations of the apparatus, The apparatus has been (l)J. C. White, ANP Quar, Prog. Rep. June 10, 1953, OBRNL-1556, p. 98. 118 simplified and further simplification is contemplated. Petrographic examinations of about 750 samples of fluoride mixtures were completed. Optical data are reported for CsUF, Cs,Zr¥F, LisUF7, KUFS, K,UF,, K;UF,, Na,Th¥,, Na,UF,, BbUFg, Rb UF,, and Rb,ZrF,. The activities of the Analytical Service Laboratory included the de- velopment of a method for the determi- nation of uranium in the ARE fuel con- centrate and a revision of the procedure for the determination of sodium in fluoride salt mixtures. During the quarter, 930 samples were analyzed that involved 9953 determinations. ANALYTICAL CHEMISTRY OF REACTOR MATERTALS J. C. White Analytical Chemistry Division Determwmination of Zirconium by Differential Spcctrophotometry (D. L. Manning, Analytical Chemistry Dai- vision). The reaction between conium and sodium alizarin sul fonate (alizarin red-S) is commonly used for the spectrophotometric determination Z17T - of low concentrations of zirconium., WOTK initiated to reaction, by means of The present adapt this differential spectrophotometry,(?’? to the determination of zirconium in ARE fuels which contain approximately 35% zirconium., The method currently used, the gravimetric determination of zirconium by precipitation as was zirT- conium tetramandelate and subsequent ignition to the oxide, although satisfactory in most respects, requires considerable time, The differential technique, which combines the speed of (2)C. F. Hiskey, dnal. Chen. 21, 1440 (1949). spectrophotometric procedures with the inherent precision of the gravimetric method, should greatly reduce the actual time of determination. In the differential method, the optical- density scale 1s set at zero against a blank of a reference standard, which is a highly light-abserbant solution. GCreater concentrations of the given component are then measured from this zero point, In order to obtain the amount of light required to adjust the zero scale against the reference standard, the spectrophotometer slit is set at wider apertures than those for normal use. For those systems which conform to Beer’s law at wide slit widths, the resultant optical density readings are directly pro- portional to the difference in con- centration between the standard and the unknown solutions., A standard curve was prepared by plotting the optical density of known zirconium concentrations over the range 1.0 to 1.5 mg per 25 ml against the increase in concentration over that of the reference solution, 1 mg per 25 ml. This plot is a straight- line relationship which indicates con- formity with Beer’'s law. The method was tested by taking a seclution of known concentration and determining its concentration of zirconium from the standard working curve, A practi- cal test was then conducted by adding known amounts of zirconium to a synthetic ARE fuel., The data obtained indicate that the precision of this method compares favorably with that of the mandelic acid gravimetric me thod. | Determination of Chromium and Chromium Trifluoride (D. L. Manning, Analytical Chemistry Division), The ANP Beactor Chemistry Group has under- taken the study of the kinetics of the reaction (1) CrF, + UFS::icr" + UF, in a medium of NaZrF of the order of 800°C. at temperatures The analytical PERIOD ENDING SEPTEMBER 16, 1953 chemistry problem is the determination of the concentrations of the reactants and products. The separation and determination of UF,; and UF, are dis- cussed In the following section, The sample must be dissolved or leached without altering the oxidation states of the components; thus a limitation i1s imposed on the usual methods of dissolving NaZrF,, namely, acid attack with nitric, perchloric, or sulfuric acids and fusion with alkali carbonates or nitrates., So- lutions that form stable complexes with the oxidation states involved were investigated for possible appli- cation. Pribil and Klubalova(3) re- ported that disodium dihydrogen ethylenediaminetetraacetic acid (EDTA) forms a very stable complex with Cr(III). Preliminary tests revealed that 50 mg of CrF, dissolved in 50 ml of an acetate-buffered solution (pH 4.0) of EDTA in about 2 to 3 hours. Chromium metal and the trivalent oxide Cr,0, were insoluble in this medium. Further tests were conducted 1n which CrF, was mixed with NaZrF_, and the mixture was leached with EDTA solution. The trivalent chromium was guanti- tatively leached, while the NaZrF, was not dissolved to any appreciable extent, Chromium difluoride dissolves much more rapidly in EDTA solution than does the trifluoride, The distinction of CrF, from CrF, on this basis would be difficult. From the data obtained to date, 1t can be concluded that EDTA in acetate-buffered solution can selectively leach CrF, from the NaZrF, matrix, and thus the distinction be- tween CrF; and chromium metal or oxide can be made, Petermination of UF, and UF, (W. J. Ross, Analytical Chemistry Division). In addition to the investigation of the determination of chromium 1in the reaction represented by Egq. 1, a study (3)3. Pribil and J. Klubaleva, Collection Czechoslov. Cher. Comsmun. 15, 42 (1950). 119 ANP QUARTERLY PROGRESS REPORT was undertaken to determine the con- centrations of the vranium compounds, The method developed by Manning, Miller, and Rowan‘*) for the determi- nation of UF, in the presence of UF, cannot be applied successfully for this determination because of the possible presence of materials, other than UF,, which liberate hydrogen upon acidification., The ease of oxidation of U(III) to U(IV), in addition to the very small concentration (500 ppm) of the reaction components involved, has led to the investigation of the selective dissolution of a component from the solid sample without chemically altering the oxidation state of the uranium. A large number of complexing agents have been investigated as possible selective solvents., These reagents include ammonium oxalate, ammonium citrate, ammonium tartrate, salicylic acid, sodium alizarin sulfonate, hydrogquinone, a,a! dipyridyl, quinali- and EDTA. In general, the results obtained can be summarized as follows: None of the reagents in ageuous solutions of various degrees of acidity provide separation or extraction. zarin, catechol, quantitative The tetra- fluoride 1is much more soluble than the trifluoride; however, the minimum solubility of UF, found was of the order of 3 to 5 mg per 100 ml of solvent, a solubility too large for quantitative application. Also, the solubility increases with increasing temperature, Some interesting data were obtained on the dissolution of UF, and UF, 1n EDTA solutions. As much as 50 mg of UF, will dissolve 1in 10 ml of a 5% (w/v) solution of EDTA buffered to pH 4.0 (with sodium acetate and acetic acid) in 30 min at room temperature; in contrast, less than 1 mg of UF, will dissolve under the same conditions, The solubility ratio in- (4)D. .. Manning, ¥%. K. Miller, and R. Rowan, Jr., Methods of Determination nof Uranium Tri- fluoride, ORNL-1279 (Apr. 25, 1952). 120 creases further as the acidity of the solvent is decreased. Although the data are not sufficiently complete to permit any definite conclusions to be made at this it would appear that the use of EDTA solutions for the separation of UF, and UF, 1is possible 1f the solubility of UF; can be reduced still further, or 1f an empirical relationship can be de- termined on the basis of the solu- bility of UF,. Determination of Oxygen in Metallic oxides(®) (J. E. Lee, Jr., Analytical Chemistry Division). The major portion time , of the current experimental work with reactions invelving the use of bromine trifluoride for the determination of oxygen in metallic oxides has been alterations of the The results of tests made confined to apparatus. to date on reactions with oxides indicate that considerable extension in the range of sample size may be accommodated if adequate instrumen- tation can be provided, Recent diffi- culties with the vacuum phases of the processing procedure show that the valves 1in the nickel section of the equipment are not dependable over reasonable periods of time, It has also been determined that the com- plexity of the entire apparatus can be greatly reduced. PETROGRAPHIC EXAMINATION COF FLUORIDES G, D. White T. N. McVay, Consultant Metallurgy Division Petrographic examinations of about 750 samples of fluoride mixtures were carried out. The optical data col- lected for various new fluoride com- pounds are given below. CsUF, Color: Z, sky blue; X, greenish blue Crystal form: monoclinic (S)J. C. White, ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p. 97. Interference figure; biaxial positive with 2V = 45 deg; X is at an angle to £ of 10 deg Refractive indices: a = ’y: Has polysynthetic twinning Cszszfi Color: colorless Interference figure: unliaxial negative 0 = 1.482 1.460 Befractive indices: i LiEUF7 Color: Z, dark green; X, light green Interference figure: biaxial positive with 2V = A5 deg Refractive indices: a = 1,468 vy = 1.476 KUF Color: green Crystal form: rhombohedral Interference figure: untaxial negative Refractive indices: ¢ = 1.510 E 1.504 it K,UF, Color: laght olive drab Crystal form: hexagonal Interference figure: wunidxial positive Refrartive indices: 0 = 1.484 E = 1.512 K,UF, Color: Z, light blue; X, colorless Crystal form: not cubic or tetragonal, as described by Zachariasen(®? Interference figure: biaxial negative with 2V = 70 deg Refractive index: 1.414; low birefringence (G)J. J. Katz and E. Rabinowitch, The Chemistry of Uranium, p. 379, NNES VIII-5, MeGraw-Hill, New Yoerk, 1951. PERIOB ENDING SEPTEMBER 10, 1953 NazThF6 Color: «colorless Interference figure: wuniaxial positive Befractive indices: (Q = 1.468 E = 1.492 NaaUF7 Color: greenish blue Crystal form: tetragonal Interference figure: uniaxial negative Refractive indices: & = 1.417 E = 1.411 BbUF5 Color: Z, blue; X, green Crystal form: probably monoclinic biaxial negative with 2V = 75 deg; Y 1is at at angle to € of 20 deg Interference figure: Refractive indices: a = 1.514 1.528 i Has polysynthetic twinning Rb,UF, Color; green isotropic Crystal form: cubac Refractive index: 1.438 BbzszG Color: c¢olorless Crystal form: hexagonal or tetragonal uniaxial negative 0 = 1.438 E = 1.432 Interference figure: Refractive indices: SUMMARY OF SERVICE CHEMICAL ANALYSES J- CI White Al Fc BO&mEr, Jre Cc B- WilliaMS Analytical Chemistry Division A procedure was established for the determination of uranium in the ARE fuel concentrate, NaF—ZrF4-UF4 (65-15- 20 mole percent), which was The procedure, formulated in cooperation with the Laboratory Division of Y-12 121 (L. A, Stephens, private communication to J, C. White), 1s based on the method reported by Voss and Greene,(7) A 5-g sample was dissolved in HCI1- ANO, ; the small residue (about 100 mg) was dissolved by fusing with potassium pyrosulfate, The solutions were then combined, converted to sulfate by fuming with sulfuric acid, and passed through a Jones reductor. Trivalent uranium was oxidized to the quadri- valent state by air. The completeness of oxidation was observed by inserting a platinum-calomel electrode couple 1in the solution and noting when a constant potential existed. A calculated amount of potassium dichromate (NBS standard) was added in slight excess of that required to oxidize the uranium quantitatively., This excess di- chromate was then reduced with standard ferrous ammonium sulfate solutien, The end point was detected potentio- metrically by using the platinum-~ calomel electrode couple. The standard deviation was 0.16%, and the relative standard error at the 95% confidence level was 0,1%. Changes in the procedure for the determination of sodium in the fluoride (Y)F. S. Voss and R, E. Greene, 4 Precise Potentiometric Method of Uranium Analysis, paper presented at Analytical Information Meeting, Qak Ridge National Laberatory, May 19-21, 1953. 122 salt mixtures were effected to permit aliquots from the master solution (5-g sample in 500 ml) to be taken for this determination so that it would not be necessary to dissolve a separate sample, A considerable saving of time was thus realized, During the quarter, the work of the Service Laboratory, as before, con- sisted chiefly of the analysis of fluoride fuel mixtures and alkali metal fluorides. The Analytical Chemistry Laboratory received 878 samples and reported 930 samples that involved a total of 9953 determi- nations (Table 10.1)., The backlog of analyses was reduced to 35 samples, TABLE 106.1. SUMMARY OF SERVICE ANALYSES REPORTED NO. OF NC. OF SAMPLES | DETERMINATIONS Reactor Chemisztry 172 1111 Corrosion Studies 4835 5838 Experimental Fngineering; 233 2708 ARE Fluid Circuit 13 196 Radiation Damage 24 64 Physical Properties and Heat Transfer Studies 3 36 930 9953 INTRODUCTION AND SUMMARY E. P. Blizard, Physics Division The Bulk Shielding Facility (sec. 11) has been used primarily for critical experiments of reactor configurations for the Tower Shielding Facility. The Bulk Shielding Reactor is well suited for this work because it is similar to the Tower Shielding Reactor; also the loading facilities for the two reactoers are similar. The calculated gamma radiation reaching the crew compartment of the divided shield appears to be somewhat lower than that measured in the Bulk Shielding Facility, presumably because of limitations in the experi-~ mental setup. The Lid Tank Facility is being used to measure removal cross sections, as well as to examine specific shield designs (sec. 12). 1In addition to the measurement of the shielding efficiency of part of the G-Eair-cycle reactor, about 80 configurations of near-unit shields for the circulating-fuel reflector-moderated reactor have been examined. Since these latter shields are neither asymmetric nor dependent upaon much shielding at the crew position, the Lid Tank Facility data can be considered quite reliable. Improvementsin the reflector-moderated- reactor shield weights other than those accruing from replacement af water by a better hydrogenous material will probably be small. The removal cross sections of beryllium, fluorine, and lithium were found to be 1.13, 1.36, and 1,44, respectively, The removal cross section of uranium was measured in conjunction with a study of the effectiveness of uranium as a shield material, The Tower Shielding Facility is taking shape; however, construction is approximately one month behind schedule (sec. 13). Most concrete work has been completed and the towers are being erected. Fabrication and/or procure- ment of controls and instrumentation have been initiated and should be effected by the end of October. The Safeguards Committee reviewed the facility and requested only one change; approval 1s expected subsequent to a Committee inspection. Although several shielding studies, including the study of the spectrum of neutrons attenuated by water and by neutron streaming in iron, have advanced the shielding art, the most important contribution during this quarter accrued from the Summer Shielding Session (sec., 14). This four-week study session, which was attended by 18 representatives from interested groups, was for the purpose of standardizing shielding design methods. In addition to achieving this end, better methods for the interpretation of bulk shielding data, as well as the calculation of self- absorption in the core, were obtained, Perhaps the single most important conclusion of the session was that the large-scale tests planned for the Tower Shielding Facility are urgently needed for obtaining really reliable shield weight estimates, The shielding program at the Laboratory 1is being realigned so that as many as possible of the shielding uncertainties can be studied with the facilities now available, 125 ANP QUARTERLY PROGRESS REPORT 11. BULK SHIELDING FACILITY R. G. Cochran F. C. J. D. Flynn M. P. Haydon K. M. Henry Maienschein H, E. Hungerford E. B. Johnson T. A. Love G. M. McCammon Physics Division Testing of a series of critical loadings for the reactor for the new Tower Shielding Facility has been initiated in the Bulk Shielding Facility. One recently tested loading was for a water-reflected reactor, which was the first completely water- reflected reactor used at the BSF. Owing to 1its simplicaity, it was con- sidered desirable to use the same loading in a series of fundamental reactor experiments which included tests on the temperature coefficient, xenon poisoning, and temperature stability, ported in the Physics Division Semi- These tests will be re- annual Progress Report,¢!’ Calculations of the gamma radiation reaching the crew shield designed by the Shielding Board have been re- ported, (%) and the results have been compared with new experimental data obtained at the HSF, TOWER SHIELDING REACTOR CRITICAL EXPERIMENT The five- by six-element con- figuration shown in Fig., 11.1 was the first reactor arrangement tested for the Tower Shielding Reactor. The reactor was completely water-reflected, and the first loading required 3550 g of uranium for a critical mass, The final loading contained 3829 g of enriched fuel, an amount sufficient to (I)Phys. Semiannual Prog. Rep. Sept. 10, 1953 (to be published) (classified}. (2)Report of the ANP Shielding Board for the Aircraft Nuclear Propulsion Program, ANP-53 (Oct., 16, 1950). 126 override the xenon poison for extended operations at 100 kw, GAMMA-RAY AIR-SCATTERING CALCULATIONS Calculations of the gamma radiation reaching the outer side of the crew compartment in the standard divided- shield design(?’ have been reported recently.{3) A repeat of the air- scattering experiments, plus additional experiments{*) of a more basic nature, has led to a re-evaluation of the experimental data on the gamma dose received at the crew compartment, Both the calculated and experimental doses are compared in Table 11.1 with the dose predicted by the Shielding (3)F. Bly and F. C. Maienschein, A Calculation of the Gamma Radiation Reaching the ANP-53 Crew Shield, ORNL CF-53-5-117 (to be published). (4)H. E. Hungerferd, The Skyshine Fxperiments at Rulk Shielding Facility, ORNL-1611 (tobe published), TABLE 11.1. GAMMA DOSE AT THE OUTER SIDE OF THE CREW SHIELD DUE TO THE RADIAL SCATTERED BEAM METHOD SCATTERED BEA%G RATIO (r/hr/watt X 10°%) Calculated (ref. 3) 10 3 Experimental (ref. 4) 15 4.3 Shielding Board (ref. 2) 3.5 1 PERIOD ENDING SEPTEMBER 10, 1953 DWG 21122 5;7///'2 OO OO®E OODS OO 136. — f 52 ‘ {ufif5¢\' 38.18 6885! A NG — ~ o’ — — 54 \( N 56 } N 6 8 S 10 136.01 15 16 17 18 136.99 138.14 137.39 S "~ 27\ 28 35 37 38 39 137.15 @ O e - _— — - OOODEOOOE \( 5?\ ( 58\ ’59\\ 3BT ) \134.65, | I / ~ 3 SAFETY ROD POSITION C CONTROL ROD FPOSITICN TOTAL U = 3829.26 g —— Table Board. This table supersedes 13.2 in the previous report.(3) The discrepancy between the calcu- lation reported by the Shielding Board(?) and the present calculation, which i1s based on experimentally determined leakage spectra and the air-scattering experiment at the Bulk Shielding Facility, 1s not The Shielding Board made at a time when incomprehensible, calculation was (S)J. L. Meem et al., ANP (uar. June 16, 1953, OBNL-1556, p. 117. Prog. Rep. NN \,/j is Loading 22 for Tower Shielding Reactor Critical Experiment, very little experimental data existed, and 1t could not be expected to approach the accuracy of the calculations based on subsequent experiments. The dis- crepancy between the air-scattering experiment data and the present calcu- lation can probably be explained on the basis that the air-scattering experiment was not a true mockup, of the situation which was calculated. It is not hard to understand that the surface of the water in the pool is a poor approximation to a round hy- drogenous reactor shield such as was assumed 1in the calculations, 127 12. LID TANK FACILITY C. L. Storrs, GE-ANP G. T. Chapman J. D. Flynn J. M. Miller F. N. Watson Physics Division The series of shield mockups for the reflector-moderated reactor has been completed; studies of 82 con- figurations were made. These ex- periments have been used for exploring some of the many possible shield combinations and have also put the shield-weight calculations on a sounder basis. Calculated shield weights, based on these data, are given in the section on ‘“Aircraft Beactor Design Studies” (sec. 3). The Lid Tank Facility has also been used this quarter to 1nvestigate the shielding efficiency of the transition section of a direct-cycle reactor shield mockup which 1s to be tested on the Tower Shielding Facility. Some new basic shielding data have been obtainedwhich involved a measure- ment of the effectiveness of natural uranium as a neutron and gamma-Tay shield. REFLECTOR-MODERATED-REACTOR SHIELD TESTS The program of Lid Tank Facility tests for the reflector-moderated reactor shield(!’ has been completed with measurements made behind 82 con- figurations. Interpretation of the data and further design studies are 1in progress, and a complete report(?) of the measurements 1s being prepared. The configuration tests are described in Table 12.1. In addition, ploratory investigation of the optimum placement of lead 1n water was made; it was concluded that no important an ex- (1A, P. Fraas and J. B. Trice, ANP Quar. Prog. Rep. #Har. 10, 1933, ORANL-1515, p. 7 (Q)F. N. Watson, A. P. Fraas, M. E. LaVermne, and F. H. Abernathy, Lid Tank Shielding Tests of the Reflector-Moderated Reactor, ORNL-1616 (to be published). 128 decrease 1n weight could be achieved by this process. During this gquarter, two signifi- cant 1mprovements were made 1n the materials and equipment used in the Lid Tank tests. (1) The beryllium pellet tank and beryllium slabs(3) were replaced with a tank that con- tained 11.5 in. of solid beryllium in the form of small blocks. Reflector coolant tubes were simulated, to some extent, by the aluminum walls of this tank and by some additional stainless steel and aluminum. (2) The two dry tanks used to contain the various com- ponents in the first 54 configurations tested were replaced by one large tank partitioned to hold all the dry com- ponents, including the new beryllium tank, in one section and the borated water in a second section. This ar- rangement eliminated the spurious water layers between the tanks. After these 1mprovements were made, measurements behind many of the earli- er configurations were repeated. The gamma doses were found to be somewhat higher, probably because of capture gammas formed in the iron partition be- tween the dry and wet sections of the tank (Fig. 12.1) and because of an in- crease of hard capture gammas 1in the lead region as a result of the elimi- nation of the water layers, sertion of boron-impregnated Plexiglas (“plexibor”) between the lead layers reduced the gamma dose to the levels observed behind the original con- figurations and confirmed the capture gamma-ray hypothesis. The 1in- 3 ( )J. D. Flynn et al., ANP Quar. Prog. Rep. June 10, 1853, ORNL-1536, p. 119. TABLE 12.1 SUMMARY OF LID TANK MOCKUPS OF REFLECTOR-MODERATED REACTOR AND SHIELD CONFIGU- NEUTRON HEAT NEUTRON : (a) LIQUID NEUTRON OTHER INFORMATION RATION REFLECTOR CURTAIN EXCHANGER CURTAIN PRESSURE SHELL GAMMA SHIELD suteLpt®’ 1 Be pellet tank,{C) 9.2 ¢cm of solid B4C tank(d) NaF tanks(e) B4C tank I 3/4 in. of Fe None Plain H2O Air wafer(f) inserted between Be 1, 2 Be reflector elements la n " " " " " " No air wafer used 9 " " n " " 2 1/2 in. of Pb (D) " 3 ¥ " " " " 4 1/2 in. of Pb (D) " Last 2 in. of Pb in 15-in. dry tank, leaving large {~11in.) ’ air void after Pb 4 " " " f " 7 1/2 in. of Pb (D) " Large air void after Pb g " " " " " 2 1/2 in. of Pb (D), 5 in. of Pb (W) " Air void replaced with H,0 6 " " " " " " " Outer 3 in. of Pb moved oug, leaving 12-cm H,0 gap 7 " " " " n 2 1/2 in. of Pb (D), 2 in. of Pb (W) t Same as configuration 3 except air void replaced with HEO 8 " " " " " 9 1/2 in. of Pb {D), 5 in. of Pb (W) " 4-cm H,0 gap preceded each of . last four Pb slabs 9 " n " " " " 10 in. of borated H,0 Borated H,0 replaced plain H,0 (1% B) in 15-in. tank; 4-cm bo- rated H,0 gap preceded each of last four Pb siabs 190 " i " " " " # Last four Pb slabs moved toward source 11 " H " " " " 10 in. of borated H20 Last four Pb slabs moved toward (0.99 B) source 19 n n " n " " 10 in. of borated H,0 | 4-in. borated H,0 gap pre- (2.8% B) ceded last 3 in. of Pb 13 " " " " " " 1 Last 3 in. of Pb moved toward source 14 " " n " " 2 1/2 in. of Pb (D), 2 in. of Pb (W) 13 in. of borated H,0 (2.8% B) 15 n " Na¥ tanks " u None Plain HzO 1, 2, 3, 4 16 n " f f " 3 in. of Pb (W) 12 in. of borated H2O (2.8% B) 17 " " " L “ 6 in. of Pb (W) 9 in. of borated HQO (2.8% B) 18 n " n " " 7 1/2 in. of Pb (W) 7 1/2 in. of borated HZO {2.8% B} 19 Be pellet tank, 1/4 in. of boral, n " None " n 7 1/2 in. of borated 3 1/4 in. of H,0 , 1/4 in. of boral, H,0 (2.3% B) 9 .2 cm of solid Be Zia " " n " " 3 in. of Pb (W) 12 in. of borated HzO (2.3% B) 129 TABLE 12.1 (continued) CONFIGU- ; J NEUTRON HEAT NEUTRON LIQUID NEUTRON RATION REFLECTOR NEanon EXCHANGER N hon PRESSURE SHELL GAMMA SHIELD in Rl OTHER INFORMATION 208 Be pellet tank, 1/4 in. of boral, B4C tank NaF tanks None 1 3/4 in. of Fe 4 1/2 in. of Pb (W) 10 1/2 in. of borated 3 1/4 in. of HQO, 1/4 in. of boral, 1, 2, 3, 4 H,0 (2.3% B) 9.2 ¢cm of solid Be 20c f " 1 " " 6 in. of Pb (W) 9 in. of borated H20 (2.3% B) 20d " " 1 " " 7 1/2 in. of Pb (W) 7 1/2 in. of borated H,0 {2.3% B) 90 " " " " " 7 1/2 in. of Pb (W), 1/8 in. of 7 3/8 in. of bhorated boral (W) H,0 (2.3% B) 21 " " " " n 6 in. of Pb (W) 9 in. of borated H,0 011 {p = 0.88 g/cms) placed in (2.3% B), 15 in. of 15-in. tank behind borated o1l H20 tank 22 " * " " " None Plain H,0 93 " " " " it " " Fe pressure shell in plain H,0 just outside dry tank,leaving S5-cm air void in tank 94 " " " " 2 in. of Ni " " Ni pressure shell in plain H,0 just outside dry tank 95 " " # " 2 in. of Cu n L Cu pressure shell ih plain H,0 just outside dry tank 26 " " y i 2 in. of Cu, " " Cu and Ni pressure shell in 2 in. of Ni plain H20 just outside dry tank 27 Be pellet tank, 3 in. of H,0, 9.2 cm " " L 1 3/4 in. of Fe " " of solid Be 28 Be pellet tank, 1/4 in. of boral, " " " ¥ " " 9/10 in. of H,0, 1/4 in. of boral, 9.2 cm of so0lid Be 29 Be peilet tank, 1/4 in. of boral, " " " None " " 3 1/4 in. of H,0, 1/4 in. of boral, 9.2 ¢cm of so0lid Be 30 " " " " 92 in. of Inconel " " Inconel pressure shell in plain H,0 just outside dry tank 31 1 1/2 in. of Pb, Be pellet tank, 9.2 " " " 1 3/4 in. of Fe n f cm of solid Be 32 Be pellet tank, 9.2 cm of solid Be " " " " ‘ " " 23 " " " v " 1/4 in. of boral (¥), wC tank'8) (W) " 34 " " " B " WC tank (W), 1/4 in. of boral (W) " 35 " " " " i WC tank (W) ! 36 n " " " " WC vank (W), 1 1/2 in. of Pb (W) " 37 " " " " " WC tank (W), 3 ian. of Pb (W) " 130 TABLE 12.'1 (continued) CONFIGU- NEUTRON HEAT NEUTKON LIQUID NEUTRON THER INFORMATION :fiaégg REFLECTOR CURTAIN EXCHANGER N URTAIN PRESSURE SHELL GAMMA SHIELD SRIELD OTHER INFORMA 38 Be peliet tank, 9.2 cm of solid Be B,C tank NaF tanks None 1 3/4 in. of Fe 3 in. of Pb (W), WC tank (W) Plain H,0 1, 2, 3, 4 39 Be tank(™’ B tank(®) " L " 6 in. of Pb (D) " Began using new tank which held all dry components of shield 40 " " " " " " " 1/4 in. of boral in plainwater Just outside large drvy tank 41 " " # n " " " 1/4 in. of boral just inside and 1/4 in. of boral just outside large tank 492 " " " " " " " 1/4 in. of boral just inside large tank 43 n n " " " n 1 B'? tank spliced out 2 ft on each side with 1/4 in.-thick boral slabs 44 " " " n " 3 in. of Pb (D), 3/4 in. of HZO,(j) " Boral splicing removed 3 in. of Pb (D) ‘ 45 f 1 " " " 1 1/2 in. of Pb (D), 1/4 in. of " boral (D), 3 in. of Pb (D), 1/4 in. of boral (D), 1 1/2 in. of Pb (D), 1/4 in. of Tybor(®) 46 " B4C tank " ft n n " 47 n n " L # L 15 in. of borated H,O (1% B) 48 " " " " " N n Air wafer inserted between borated H,0 tank and pre- Ceding tank 49 " " " " 0 1 1/2 in. of Pb (D), 1/4 in. of 13 1/2 in. of borated boral (D), 3 in. of Pb (D), 1/4 HZO (1% B) in. of boral (M), 1 1/2 in. of Pb (D), 1/4 in. of Tybor (B), 1 1/2 in. of Pb (W) 50 1 " " " " 1 1/2 in. of Pb (D), 1/4 in. of 15 in. of borated HZO boral (D), 2 in. of Pb (D}, 1/4 {1% B) in. of boral (D) 51 " " " B,C tank " 6 in. of Pb (D) " 59 9 n " None 1 3 in. of Pb (D), 3/4 in. of " borated H,0,/? 3 in. of Pb (D) 53 1 1/4 in. of Al, Be tank " # " n 6 in. of Pb (D) i 54 Same as configuration 17 except borated H,0 was 1% B instead of 2.8% B 55 Be tank " " B,C tank " 3 in. of Pb (D}, 3/16 in. of plexi- 22 1/2 in. of borated { Section welded on tank con- boe (1) (D}, 1 1/2 in. of Pb (D}, H,0 (1% B) taining dry compounsnts to 3/16 in. of plexibor (D), 1 1/2 in. hold liquid neutron shield of Pb (W) and thus eliminate undesired water layers between tanks 131 TABLE 12.1 (continued) CONFIGU- NEUTRON 132 HEAT NEUTRON . LIQUID NEUTRON Sg;égg REFLECTOR CURTAIN EXCHANGER CURTAIN PRESSURE SHELL GAMMA SHIELD SHIELD OTHER INFORMATION 56 Be tank B4C tank NaF tanks 3/16 in. of 1 3/4 in. of Fe 3 in. of Pb (D), 3/16 in. offiplexi- 22 1/2 in. of borated 1, 2, 3, 4 plexibor bor (D}, 1 1/2 in. of Pb (D)}, 3/16 H,0 (1% B) in. of plexibor (D), 1 1/2 in. of Pb (W) 57 " " " " " 4 1/2 in. of Pb (D), 3/16 in. of 24 in. of borated H,0 plexibor (D), 1 1/2 in. of Pb (D), (1% B) 3/16 in. of plexibor (D} 58 " " " " " Four 1 1/2-in. Pb slabs (D), each " preceded by 3/16 in. of plexibor; 3/16 in. of plexibor behind fourth slab 59 " " n B,C tank " 6 in. of Pb (D), 3/16 in. of plexi- " bor (D) 60 " 1 i " " 6 in. of Pb (D), 3/16 in. of plexi- 22 1/2 in. of borated bor (D), 1 1/2 in. of Pb (W) H,0 (1% B) 51 " " NaF tanks " " " ! " 1, 2, 4 62 " " " " " § in. of Pb (D), 3/16 in. of plexi- 24 in. of borated H20 bor (D) (1% B} 63 " " " " " Four 1 1/2-in. Pb slabs (D), each n preceded by 3/16 in. of plexibor; 3/16 in. of plexibor behind fourth slab 64 " " NaF tanks " " Four 1 1/2-in. Pb slabs (D), each " 1, 4 preceded by 3/16 in. of plexibor; 3/16 in. of plexibor behind fourth slab 65 " " " " " Three 1 1/2-in. Pb slabs (D), each " preceded by 3/16 in. of plexibor; 3/16 in. of plexibor behind third slab 66 Be tank, 9.2 cm of solid Be " " " L Three 1 1/2-in. Pb slabs (D), 3/16 22 1/2 in. of borated in. of plexibor behind each slab; H,0 (1% B) 11/2 in. of Pb (W) 67 " " " " " Three 1 1/2-in. Pb siabs (D)}, 3/16 21 in. of borated H20 in. of plexibor behind each slab; (1% B) 3 1in. of Pb (W) 68 " " n " " Three 1 1/2-in. Pb slabs (D), 3/16 24 in. of borated H,0 in. of plexibor behind each slab (1% B) 69 x " NaF tanks " Bk 1 1/2 in., of Pb (D), 3/16 in. of 19 1/2 in. of borated i, 2, 3, 4 plexibor (D), 4 1/2 in. of Pb (W) HzO (1% B) 10 " n NaF tanks " " 3/16 in. of plexibor (D), 3 in. of 24 in. of borated H,0 1, 2, 4 U (b), 3/16 in. of plexibor (D) {1% B} TABLE 12.1 (continued) CONFIGU- . NEUTRON HEAT NEUTRON ; ' LIQUID NEUTRON - gfiaggg REFLECTOR CURTAIN EXCHANGER CHRTAIN PRESSURE SHELL GAMMA SHIELD SHIELD OTHER INFORMATION T1 Be tank, 9.2 cm of solid Be B,C tank NaF tanks B,C tank 1 3/4 in. of Fe 3/16 in. of plexibor (D), 3 in. of U |22 1/2 in. of borated 1, 2, 4 (D), 3/16 in. of plexibor (D), Hgoz(l% B) 1 1/2 in. of Pb (W) ? 12 " " " " " 3/16 in. of plexibor (D), 3 in. of U |24 in. of borated H,0 (D) pius Pb brick yoke, ("™’ 3/16 in. (1% B) of plexibor (D) 73 Be tank " NaF tanks " " 3 in. of Pb (D), 3/16 in. of plexi- " 1, 2, 3, 4 bor (D}, 1 1/2 in. of Pb (D), 3/16 in. of plexibor (D) 74 " " " " ' L 24 in. of borated H,0O (0.5% B) 75 " " " " " " 24 in. of plain H,0 76 W " NaF tanks f " 6 in. of Pb {D), 3/16 in. of plexi- 24 in. of borated H,0 | 3-cm air void in gamra shield 1, 4 bor (D} (1% B) 77 " " Li tanks(™) " " Four 1 1/2-in. Pb's]abs {D), each " 1, 2 preceded by 3/16 in. of plexibor; 3/16 in. of plexibor behind fourth slab (G)Lead thicknesses in the gamma shield were made up of 1- and 1 1/2-in. slabs, Where possible, these slabs were grouped with the dry components of the shield, but in many cases it was necessary to place some of the slabs in the liguid neutron shield container. (D) indicates dry slabs; (W) cates wet slabs, both for the plain and for the borated water, indi - (b)The borated water used in configurations 1 to 54 was contained in a separate tank. rations 55 to 77, the solution was contained in the wet section of a single large tank. In configu- 1.233 g/cms) encased in iron; dimensions were 0.3 cm of Fe, 28,5 cm of totaling 29.1 cm. (C)Beryllium pellets {p = Be, 0.3 cm of Fe, (d)Boron carbide {(p = 1.9 g/cms) encased in aluminum; dimensions were 0.16 cm of Al, 2.70 c¢m of B,C, 0.i6 cm of Al, totaling 3.07 cm. (C)SOdium fluoride {(p = 6.96 g/cma) encased in iron; dimensions for tanks 1 and 2 were 0.43 cm of Fe, 2.54 cm of NaF, 0.48 cm of Fe, totaling 3.50 cm; dimensions for tanks 3 and 4 were 0.485 cm of Fe, 4.44 cm of NaF, 0,48 cm of Fe, totaling 5.4¢ cm. {f)Air in wafer with 0.16-cm aluminum walls., (g)Tungsten carbide pebbles {(p = 8.05 g/cms} encased in iron; dimensions were 0.635 cm of Fe, 5.08 cm of WC, 0.635 ¢m of Fe, totaling 6.35 cm. h)gp1id beryliium (P = 1.48 g/cm>) blocks in aluminum tank. steel within tank simulated reflecter coolant tubes. Dimensions for the tank were .64 cm of Al, cm of Be, 0.16 e¢m of Al, 0.03 cm of stainless steel, 7.30 ¢m of Be, 0.16 cm of Al, 0.02 cnm less steel, 14.61 cm of Be, 0.15 ¢cm of blotter paper, 0.64 cm of Al, totaling 31.03 cm. Additional aluminum and stainless 7.30 of stain- (i)Boron isotepe 10; dimensions were 0.05 cm of stainless steel, 1,53 of BIO powder, 1.52 cm of Al, totaling 3.1 cm. (j)The 3/4 in. of water was contained in a thin-walled aluminum tank, (k) (D plexiglas impregnated with boron. (M)Uranium slab was 3 in. thick by 3 ft by 3 ft, Since the radiation cone from the source at the uranium slab position had a greater diameter than 2 ft, a 4-in,-thick yoke of lead bricks was built around the top and sides of the slab to reduce streaming. Tygon impregnated with boron, (n)Lithium (p = 0.534 g/cm3) encased in iron; dimensions for tank 1 were 0,64 cm of Fe, 2.54 cm of Li, 9.64 cm of Fe, totaling 3.82 cm; dimensions for tank 2 were ©.48 cm of Fe, 3.65 cm of Li, &.48 cm of Fe, totaling 4.61 cm. 133 /s in. OF PLEXIBOR B,C DWG. 21123 ) SN NSNS 1.84g/cm3 aF TANKS DO N NN N /300277, Be (p / . . O o 0 o — T g —_— 2 — ........ — O o o e L o - © BORATED H,0 —o- (1% B8) - —° e — o 0o oo Q G0 oo c o ° o 0 o Qg _Qo o o © 7 ~ H,0 GAP Fig. 12.1. Gamma-Ray Attenuation Data. The gamma-Tay attenuation measurements for configurations 16, 17, and 18 show the effect of 1ncreasing the thickness of the lead layer (Fig. 12.2). The measurements on configuration 33 show the etfect of substituting tungsten carbide for lead in the gamma shield. On the basis of equivalent density, 2 1in. of tungsten carbide pellets and 1/2 in. of steel (container walls) should have had the same gamma attenu- ation as 1.8 in. of lead. Actually, the attenuation was only as effective as 1.1 in. of lead. This indicates that the high rate of capture or in- elastic scattering gamma production 1n tungsten reduces its value as a shield- HORIZONTAL SCALE tcm =10c¢cm Typical Reflector-poderated Reactor Shield Configuration in the LLid Tank Showing Single Containing Tank, ing material when it must be used in a high neutron flux. Except for the two improvements described above and the addition of a 3/16-in. plexibor slab behind the lead slabs in the dry section of the tank, configuration 59 was the same as con- figuration 17. (Figure 12.11s a sketch of configuration 59; configuration 17 is similar to the configuration shown in Fig.14.1 o fOBNL-1556.¢*?) The longer relaxation length in water fol lowing confignration 59 (Fig. 12.2) indicated an increase 1n the penetrating capture gammas from lead and iron and a rela- decrease 1in the tive softer water 4 yyia., Fig. 14.1. 135 ANP QUARTERLY PROGRESS REPORT DWG. 21124 A CONFIGURATION 16 5 M CONFIGURATION 17 O CONFIGURATION 18 @ CONFIGURATION 33 A CONFIGURATION 59 0 CONFIGURATION 63 2 = -._:i = W5 Q ] < = = 42 ~ \J 1071 o ~3 10 100 110 120 130 140 150 160 170 Zz, DISTANCE FROM SQURCE (cm) Fig. 12.2. Gamma Dose Behind Typical Reflector-Moderated Reactor Shield Mockups (Configurations 16, 17, 18, 33, 59, 63). 136 capture gammas, Data taken behind con- figuration 63 showed the effect of in- serting four plexibor slabs into configuration 59, one slab preceding each 1.5 in. of lead slab in the gamma shield. One heat exchanger tank was removed to allow room for the plexibor slabs. The gamma dose was reduced, but the long relaxation length in the water indicated that soft gammas from the source or captures in the water did not contribute importantly to the total dose in either configuration 59 or 63. Configuration 64 (data not presented here) was the same as configuration 63 except that the number of heat ex- changer tanks was reduced from three to two. The result was an increase 1n gamma dose of about 10% over that ob- served behind configuration 63, The gamma measurements for con- figuration 66 (Fig., 12.3) show the effect of adding 9.2 cm of beryllium to the reflector thickness. It was impossible in assembling this con- figuration to place the fourth lead slab with the other three in the first dry tank compartment. Placing it in the borated water tank increased the effectiveness of the lead and tended to mask the effects of the increased reflector thickness; so configurations 65and 68 were tested without slab four in the shield, leaving only 4 1/2 in. of lead. Configuration 65 had 11.5 in. of beryllium in the reflector, and configuration 68 had 15.1 in. of beryllium. Measurements behind configuration 72 (Fig. 12.3) show the effect of using 3 in. of natural uranium clad on each side with 3/16 in. of plexibor as the gamma shield. By logarithmic interpo- lation, the natural uranium was found to be as effective for attenuation as 4.7 in. of lead; however, on a density basis, 3 in. of uranium would be ex- pected to be equivalent to 5.15 in. of lead. The data for comnfiguration 73, 74, and 75 show the effect on the gamma PERIOD ENDING SEPTEMBER 10, 1953 dose of borating the water in the neutron shield. The addition of 1.2% boron to the water eliminated a large portion of the medium-hard capture gammas in the water (Fig. 12.4) and thereby reduced the gamma dose by a factor of 2.7, This elimination of water capture gammas leaves a harder gamma spectrum {(presumably from metal captures), and consequently the attenu- ation curve exhibits a longer relax- ation length, Note also that the rear wall of the borated water tank acts as an increasing source of capture gammas with a decrease in boron content. In configuration 77 (data not pre- sented here), two lithium-filled iron tanks were substituted for the two sodium fluoride-filled iron tanks, which were used to simulate the heat exchanger in configuration 64. This substitution produced no measurable change in the gamma dose. Neutron Attenuation Data. Several neutron attenuation curves are shown in Fig. 12.5. Since changes in either the heat exchanger or the lead region had li1ttlie effect on the neutron at- tenuation, it appears from comparison of the measurements behind configu- rations 17 and 55 that the beryllium reflector accounts for the greater initial attenuation of fast neutrons. Substitution of uranium for lead in the gamma shield caused a noticeable decrease in the neutron flux (Fig, 12.5, configuration 72). A more systematic study of the shielding properties of uranium is discussed in a following section, ‘“‘Investigation of Shielding Properties of Uranium,” The effect of varying degrees of water boratiocn on the neutron flux is shown in Fig. 12.6. From the data taken outside the borated water tank, it is evident that the boron content affects only the thermal-neutron flux, since its effect is quitewell localized in the borated region. 137 ANP QUARTERLY PROGRESS REPORT S DWG. 21125 10— — e — — —~ s .| ® CONFIGURATION 17 | R P —————— A CONFIGURATION 65 — — e O CONFIGURATION 66 |~ toe - ® CONFIGURATIONSS | | L [0 CONFIGURATION 72 Ve I O B T ’» e e e —— b L - e .5 in. OF BERYLLIUM REFLECTOR, e E— 4.5in. OF LEAD, 9-c¢m VOID T s 1 m %’V/o 154 in. OF BERYLLIUM REFLECTOR, ool oo 4.5 in. OF LEAD,1.5-¢cm VOID— < E W 410! e O D e — e < ooy s = < 5 b < |4 2L o ] 5 . - - 2 ,,,,,,,, — e - 1073 | 100 10 120 130 140 150 160 170 Z, DISTANCE FROM SOURCE ({cm) Fig. 12.3. Gamma Dose Behind Typical Reflector-Moderated Reactor Shield Mockups (Configurations 17, 65, 66, 68, 72). 138 PERIOD ENDING SEPTEMBER 10, 1953 a——_ DwG. 21126 GAMMA DOSE {mr/hr) 1.2% BORON IN WAT O CONFIGURATION 73 _ i, . . ® CONFIGURATION 74 472 'n.OF LEAD A CONFIGURATION 75J IN GAMMA SHIELD 30 50 70 S0 110 130 150 170 z, DISTANCE FROM SOURCE (cm) Fig. 12.4. Gamma Dose Behind Typical Reflector-Moderated Reactor Shield ckups (Configurations 73, 74, 75). 139 ANP QUARTERLY PROGRESS REPORT OWG. 21127 o ) K ——6in. OF LEAD, ™ N~ 10-cm VOID — THERMAL-~NEUTRON FLUX(H%W T T T T Bin.OF LEAD, I | __2-cm VO'D‘;“T | T 1 B CONFIGURATION A7 RV‘_%—* N 2 —————=— [ CONFIGURATION 5 - B | O 05+ 3in. OF URANIUM, O CONFIGURATION 72 1 5-cm VOID ——1— [ ] 1~ - | ] B S T i L N 5 L - B _ T - ] ol ‘ R e o> 1 40 60 80 100 120 140 160 z, DISTANCE FROM SQURCE (cm) Fig. 12.5. Thermal-Neutron Flux Behind Typical Reflector- Moderated Reacto Shield Mockups (Cenfiguratioms 17, 55,72). 140 PERIOD ENDING SEPTEMBER 10, 1953 DWG. 24128 NO BORON 10 o 0.5% BORO PLAIN WATER REFERENCE CURVE , ™ (& THERMAL -NEUTRON FLUX (mfm) N — | - 2 O CONFIGURATION 55, 1.2 % BORON ¥ CONFIGURATION 74, 0.5% BORON tO O CONFIGURATION 75, NO BORON 10 40 60 80 100 120 140 160 z, DISTANCE FROM SOURCE {cm) Fig. 12.6. Therwal-Neutron Flux Behind Typical Reflector-Moderated Reactor Shield Mockups (Configurations 55, 74, 75). 141 ANP QUARTERLY PROGRESS REPORT SHIELD INVESTIGATION FOR GE-ANP A brief experiment was carried out to determine the shielding efficiency of the “transition section” in the air-cooled R-1 reactor designed by GE-ANP. This section is between the reactor and the annular air ducts and consists of alternate slots perpen- dicular to the reactor surface for cooling air and moderator water. A mockup of the inlet annular duct and the transition section was placed 1in the Lid Tank Facility, and radiation measurements were made along the source axis and outside the duct (Fig. 12.7). The measurements were then repeated with the transition section replaced by a rectangular box designed to pro- vide the same amounts of air, water, and aluminum. The results, givenin Figs, 12.8 and 12.9, show that the rectangular box provided better shielding, in all cases, by a factor between 1.5 and 2, MEASUREMENTS LOCATION OF SIDi LOCATION OF CENTERLINE MEASUREMENTS — = 1 AIR-FILLED DUCT TRANSITION SECTION-——— oh b At B E L N | Fig. 12.7. Mockun of Air Duct for GE~ANP R-1 Reactor. 142 It was therefore concluded that the mockup of the complete B-1 shield, now under construction for testing in the Tower Shielding Facility, could be made with the much simpler rectangular design, and appropriate corrections could be made to the resulting data. INVESTIGATION OF SHIELDING PROPERTIES OF URANIUM information on the use of natural uraniumas a shielding material has been obtained from measurements in the Lid Tank Facility. A 3-in. uranium slab was placed at various distances from the source. In some of the measure- ments, a thermal neutron shield of boron was placed either on the source Some side or on both sides of the uranium slab. The thermal-neutron fluxes measured 130 c¢cm from the source for the various configurations are recorded in Table 12.2. Note the effect of the boron shield (3/16-in. slabof plexibor) on the production of fission neutrons in the uranium. made when the uranium was not near the source are of limited interest because of the streaming of radiation around the slab., In the data presented in Table 12.3, this difficulty was avoided by having the detector close behind the uranium. (Gamma measurements however, Additional measurements on uraniui were made when it was substituted for lead in the gamma shield region of the mockups for the reflector-moderated reactor shield (cf., ‘Beflector- Moderated-Reactor Shield Tests,” above). The decrease 1in the neutron flux indicated that uranium has a higher fast-neutron removal cross section than either water or lead. As is the case with all heavy elements, uranium is not competitive with water as a neutron shield because of 1its much greater atomic weight. However, if the capture and fission gamma pro- duction of uranium could be held down so that it would be a satisfactory gamma shield, its neutron attenuation PERIOD ENDING SFPTEMBER 10, 1953 DWG. 24430 108 — 103 - 10° - 5 e 5 QO TRANSITION SECTION 0 RECTANGULAR VOID 2 - . 105 |— 10% 5 N 2 foremmen % w0t X 103 o ) - = I > £ ~— e e S < E £ = 5 — ~ -4 uw i L g & o o 5 - 2 : B O 3 o 2t @ Lt b = = o)