v SYSTEMS LABRARIES ARt mmu—.flrfalcnakc y % A . -:-av.%f-:flg 'fin" __gfi\ Y J{?T "g_‘;-;g, "“& e v : TR ORNL-1729 This document censists of 155 pages. Copy ?J; 255 copies. Series A, Contract No. W-7405-eng-26 AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT For Period Ending June 10, 1954 W. H. Jordan, Director A. J. Miller, Assistant Director A. W. Savolainen, Editor DATE ISSUED T g0 1954 OAK RIDGE NATIONAL LABORATORY Operated by CARBIDE AND CARBON CHEMICALS COMPANY A Division of Union Carbide and Carbon Corporaotion Post Qffice Box P QOck Ridge, Tannessee MARTIN Mantzrry ¢ WA T 3 4456 0360703 o ——— _ 0P NO ;AL . Hollaender -,.;;:; . Larson "S. Livingston N, Lyon FERRO-MMZOOER-PPMEPECNLIAZIONOAOTPNEMIOMONNRO ORNL-1729 Progress INTERNAL DISTRIBUTION . B. McDoghld . Adamson 43. W g h 44. J. L. Meeg 45. A. J. Midter 46. K. Z.‘_E;gg‘s%rgan 47. E. J,Mur phy : JP Murray (Y-12) 2 J. Nessle . Patriarca : F. Poppendiek M. Reyling W. Savage W D S . Bredig . Bruce . Callihan . Cardwell . Center . Charpie . Clewett . Clifford . Cotirell . Cochran . Cowen romer . Savolainen . Shipley isman » Smith . Smith (consultant) . Snell . Storrs . Susano . Swartout DUOEPMIPMETCIPINLOEDOX L. Culler . Taylor B. Emlet (K-25) . frice K. Ergen . Yan Artsdalen P. Fraas . Vonderl.age R. Grimes . Worde E. Hoffman . Weinberg . White . Whitman . Wigner (consultant) . Williams BMEO-PETIMEmM-O0P - OOM» I DI T . Wilson Winters . D. Manly _ 95-97. Central Researé L ibrary . A. Mann Vii 98. 99. 100. 5 101, %102, 103. 104-115. 116, 17, 118-122. 123. 124. 125. 126-128. 129. 130. 131-136. 137. 138. 139. 140. 141. 142. 143. 144, 145-149. 150. 151. 152. 153-155. 156-163. 164, 165. 166. 167-170. 171-172. 173-174. 175. 176. 177. 178. 179-182. 183. 184-185. 186. 187. 188-189. 150. 191. EXTERNAL DISTRIBUTION Air Force Engineering Office, Ogk Ridge Air Force Plant Representative, Burbank Air Force Plant Representative, Seattle Air Force Plant Representative, Wood-Ridge American Machine and Foundry Company ANP Project Office, Fort Worth Argonne National Laboratory (1 copy to Kermit Anderson) Armed Forces Special Weapons Project {Sandia) Armed Forces Special Weapons Project, Washington (Gertrude Camp) Atomic Energy Commission, Washington (Lt. Col. M. J. Nielsen) Babcock and Wilcox Company Battelle Memorial Institute Bendix Aviation Corporation Brookhaven National Laboratory Buraau of Aernautics (Grant) Bureau of Ships Carbide and Carbon Chemicals Company (Y-12 Plont) Chicago Patent Group Chief of Naval Research Commonwealth Edison Company Convair, San Diego (C. H. Helms) Curtiss-Wright Corporation, Wright Aeronautical Division (K. 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Nelson) 202-211. o ~:»__:;_' w'smn (Fox Project) 212-213. B 214. : 215, Sylvanm E' Bt i foducts, Inc. 216. Tennessee V_ Authority (Dean) 217. USAF Hegdquarterl, 218. U.S. Nawal Rad:ologé%l Defense Laboratory 219.220. Umwé?é:ty of CaliforniciBadiation Laboratory, Berkeley 221-222. Up{‘varsuty of California R&fiwhon Lubomfory, Livermore 223, % ; : # Westinghouse Electric Corporatich, 230-2 0. Wright Air Development Center (WCSKS, Col J. R. Hood, Jr) 241-255. Technical lnformahon Service, Oak Rldge vi Reports previously issued in this series are as follows: ORNL-528 ORNL-629 ORNL-768 ORNL.-858 CRNL-919 ANP-60 ANP-65 ORNL-1154 ORNL-1170 ORNL-1227 ORNL-1294 ORNL-1375 ORNL-1439 ORNL-1515 ORNL-1556 ORNL-1609 ORNL.-1649 ORNL-1692 Period Ending November 30, 1949 Period Ending Februory 28, 1950 Period Ending May 31, 1950 Period Ending August 31, 1950 Period Ending December 10, 1950 Period Ending March 10, 1951 Period Ending June 10, 1951 Period Ending September 10, 1951 Period Ending December 10, 1951 Period Ending March 10, 1952 Period Ending June 10, 1952 Period Ending September 10, 1952 Period Ending December 10, 1952 Period Ending March 10, 1953 Period Ending June 10, 1953 Period Ending September 10, 1953 Period Ending December 10, 1953 Period Ending March 10, 1954 FOREWORD This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL records the technical progress of the research on the Circulating-Fuel Reactor and all other ANP research at the Laboratory under its Contract W-7405-eng-26. The report is divided info three major parts: . Reactor Theory and Design, Il. Materials Research, and Ill. Shielding Research. The ANP Project is comprised of about 300 technical and scientific personnel engaged in many phases of research directed toward the achievement of nuclear propulsion of aircraft, A considerable portion of this research is performed in support of the work of other organizations participating in the national ANP effort. However, the bulk of the ANP research at ORNL is directed toward the development of a circulating-fuel type of reactor. | The nucleus of the effort on circulating-fuel reactors is now centered upon the Aircraft Re- actor Experiment — a high-temperature prototype of a circulating-fuel reactor for the propulsion of aircraft. The equipment for this reactor experiment is now being assembled; the current status of the experiment is summarized in Section 1 of Part 1. The supporting research on materials and problems peculiar to the ARE — previously included in the subject sections — is now in- cluded in this ARE section, where convenient. The few exceptions are referenced to the specific section of the report where more detailed information may be found. The ANP research, in addition to that for the Aircraft Reactor Experiment, falls info three general categories: (1) studies of aircraft-size circulating-fuel reactors, (2) materials problems associated with advanced reactor designs, ‘and (3) studies of shields for nuclear aircraft. These phases of research are covered in Parts i, Il, and i, respectively, of this report. FOREWORD .. .. SUMMARY ..... CONTENTS ------------------------------------------------------ PART |, REACTOR THEORY AND DESIGN 1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT .. ... ... ... et e e The Experimental Pumps ..... Reactor System . . .. i v i it it i ittt it ettt aer i an s araae e ------------------------------------------------------- Heat exchangers . . .o it i it ittt it i i it it it s et i st e e e ’ Fluid circuits Fuel-enrichment ------------------------------------------------------- BYSTEM & h v s i s s i e e h e s sie s e s s s e e s s e v e e e s e n e n Loading facilities ... ... P Fuel-sampling facility . ............. S Fue[*UnlmdinngCillfy nnnnnnnnnnnnn :-n..onononlpuoonauu---au-.----.lc--: Reactor control ------------------------------------------------------ Fuel System Mock-up Tests .. v i ittt i i i it it st st i st o st aa e o s e Operation of the F vel System . ......... e et e e s s e e e Clecningnon.olonn.c-lconl.poululot.a..uno’ ------------------------- Pressure filling -------------------------------------------------------- Pressure and flow charaCteristics v v v v v v s o o s o s 0 s s 8 0 s o oo s n o nssseanncensenasens Functional tests Hot-gas test . Pump Fabrication ----------------------------------------------------- ------------------------------------------------------- AN T SIS v v ittt et e e e e e e e e e e e e s e ae e et Reactor System Component Loop . .. ... .. e e e e e e h e e e e Fuel Recovery and Reprocessing . ... .... i e s s e e e a et e e e e e 2. EXPERIMENTAL REACTOR ENGINEERING o v vt attiee e e ieneeeneanenns | In-Pile Loop Component Development .. ... ...t ienserns e e - Centrifugally sealed pump « . . . . L it i it e e e e e e e e e e Horizontal-shaft SUMP PUMP v s v o v s v s s R Gt e e et e e e e s e Vertical-shaft sump pump .. ..... M e e e e e e e e i e e e e e : Hydraulic pump drives ..« v v v i i i it it i it i i et et e e e e e e h e s e e Fluoride-to-wate rheat exchanger . .. v i ittt it et e i e e P h e _ Forced-Circulation Corrosion Loops ... ... e et e et e e Sodium-Beryllium-Inconel Mass-Transfer Test .. . ... . i i i e e e e ‘ Fluoride-to-Sodium Heat-Exchanger Test ... ............ e e e ‘ Gas-Furnace Heat-Source Development ... ... ..... .. e e e e e e | Sodium Sampler . v v v vt i vt i e e e e e e ey ek e e e e T Removal of Fluoride Mixtures from Equipment . . . . . i i i it i ittt it it i i e e i e Bearing-Materials Rotary-Shaft Seals Tests v v e oo onnens e e e s e e m e e e s e e s ----------------------------------------------------- 3., REFLECTOR-MODERATED REACTOR vttt it ittt et s e oot nnonassnennss R A Comparison of Lithium- and Zirconium-Base Fluoride Fuels .. ... ....... ... .o oL, Reactor Physics ... v i h i it v v s n v e w e s en e e e e s et aa e e Reactor Calculations « v v v v v v vt e e v s v e e a e e e n e e e e e s e h e e e e : PART 1. MATERIALS RESEARCH 4. CHEMISTRY OF MOLTEN MATERIALS .. i it i i i it i i et e es Quenching Experiments with Fluoride Systems . . . . . v ittt ittt i i i ittt sce i aeaa NaF-ZrF o e e s e NaF-UF, e e e Visual Observation of Melting Temperatures . . . v o i ittt it ittt i i it et e i e o nos Differential Thermal Analysis of the NaF-ZrF, System .. ..... ..., Filtration Analysis of Fluoride Systems . . .. v v i it it i i ittt i ittt s i ettt enn NaF-ZrF, .o e e e NaF-LiF-RbF-UF, . i i i e Thermal Analysis of Fluoride Systems . . o v i it ittt ittt ittt i e b i o e n e e e R NaF-UF, o i e s RbF-UF, o e e e e e KE-LiF-UF, o i e i e as et REF-LiIF-UF, i i s e NaF-ZrF -UF -UF, i i e NaF-ThF, . i i i i e LiF-BeF -ThF,-UF, oo i i i e L T R R Thermal Analysis of Chloride and Mixed Chloride-Fluoride Systems . .. ... ... oot UCL-UCT, o i i e s e e UG -UF o e s UG U, e i e KLU, i i s s i i i s s e RECLUCE, o i st e e O O O T X-Ray Diffraction Studies of the NaF-ZrF System ........... ... .o i, Chemical Reactions inMolten Salts . . .. o i i ittt ittt ittt it s ane s Chemical equilibriain molten fluorides . . . v v v i v it i it il i i s i et es s et n o Solubility of UF, in NaF-ZrF mixtures . ... ..o, Solubility of UF, in NaF-KF-LiF mixtures .. ..oty Solubility of UF, in NaF-RbF-LiF mixtures . ... oot Chlorinationof UF, ..o vv it i e i s e Preparation of UF3-QZrF4 .............................................. Preparation of various fluorides . . v vt v i i it it il ittt ittt st e s e e Fundamental Chemistry of Fused Salts . .. . 0 i ittt i e e it et v EMF measurements o v v v it e it sttt o st ns i s st e e e Physical chemisiry v i it ittt ittt ittt ettt ettt ettt Chemical Effects of Fission Products . . v v v i vt it i i it i it it e et i s et e e e n e ones Quantity of fission products v v v v vttt ot n e ottt v oot ettt e e e Separation of solid phases . . . . i i i i it i i it e i ettt e e e s s Effects on viscosity and heat capacity . . . . i it i it it i i i s e e e e Effects on COMmOSIoN v v i i i it o it sttt s et s n st s st nossstnenonsossssnsnss Purification and Properties of Alkali Hydroxides . .. ... ... 0.ttt Purification of hydroxides ........... e e e r et a e et e Reaction of sodium hydroxide with metals . . v i i ittt i e ittt e e e i e Production of Purified Molten Fluorides ... 0 it it i ittt ittt it et et s e Laboratory-scale production .« . i it it it i e e e et e Experimental production . . i i it i it it i i i s e et e e e e Large-scale production . .. Production of enriched fuel for in-pile loop .. « n o 9 5. CORROSION RESEARCH . ..... 6. ¢ N e B8 4 0w ® B & ® B8 9 3 4 w a a * a2 % % B B N A ® & 8 8 2 8 s o8 ¢ & & & Effect of reduced phases in fluoride melt ... .. Effect of fission products Static and Seesaw Tests of Yarious Materials in Fluoride Mixtures and Liquid Mefols s e e e e Dissimilar metals in NaF- ZrF UF Molybdenum-coated Inconel in NaF LrFA -UF, Stainless steel in lithium with lithium mtnde udded e e e e Chromalloyed steels in liquid metals . Cermets in NaF-ZrF ,-UF, « n o & & 3 m = p * & v o B o n " s a % a0 00 4.8 0 Fluoride Corrosion of Inconel in Static and Seesaw Tests .. ... Effect of chromium addition to fluoride meh . Effect of temperature Effect of surface area ... * = & » @« @ e s L ] o« » - * s » 4@ o A W ® &4 o B & & 5 2 > w 8 P 4 2 B & 5 B 4 2 8 & » 4 8 4 4 O 4 & & e 3 a ¥ R ¢ 0 o 3 ® & @ s & 8 2 & & B 8 > 2 & 4 B & B.% B3 W & s & ¥ & ® & 0 4 2 Graphite in NaF-ZrF -UF ; and in sodium .. ... NaF-ZrF NaF-ZrF 4 L ° B 8 M 9 4 w in special Inconel in Hastelloy B . in stainless steel in Incone! with stainless steel or 2 ¢ e e w0 4 ® 8 8 & & s a 8 v . 5 m % B 8 A u « 8 s .2 B 0 » 4 » & & * & = & a 2 ¢ a = Sodium in Inconel with beryllium inserts ... ... Lithium in stainless steel Fundamenta! Corrosion Research Mass transfer in liquid lead a 8 & » ¢ o LI O L 4 8 s 4 w & B & s O 4 & & 8 & B B & 0 & B B 3 B & 0w Products of hydroxide-metal reactions . ...... Dehydration of sodium hydroxide a a ¢ " o m e & e ¥ e Color changes in fused hydroxides .. ........ METALLURGY Stress-Rupture Tests of Inconel High-Conductivity Metals for Radiator Fins . ... ... Special Materials Research . .. Hastelloys B and C Nickel-molybdenum alloys & @ & o B 8 s 8 @ " e A s A2 0 e v oa » @ * % 2 B @ R M A 2 4w 0.k s = Stainless-steel-clad molybdenum and columbium Columbium Boron carbide Welding and Brazing Brazing alloy development Beryllium test assembly @ 6 & & D N s p a & & @ o n ® #F & B 0 2 0 W B B & 8 & 3 * e & & 2 U & @ 4 W 9 2 8 4 ® 6 5 8.8 B % a --------------- 7. HEAT TRANSFER AND PHYSICAL PROPERTIES Heat capacity Physical Properties Measurements @ o ® A B S B s 8 8 % 4 I A e e ow -------- 2 w & » e & @ & » Flueride Corrosion of Inconel in Thermal-Convection Loops . .. Effects of UF, and mixtures of UF, and UF, Effect of hydrogen fluoride . ....... Effect of temperature ... ... Effect of exposure time Thermal-Convection-Loop Tests of Various Materials -UF, -UF, NoF-ZrF ,-UF NoF-ZrF4_~UF4 in fluoride fuels @ u 4 4 % W e Y & e 3 8 & v & » 8 a 3 s o ------ LI L I I ) nickel inserts * 8 % 3+ 3 & b0 * 4 » o & w & B A O B e 8 8 a2 ¥ & e o 4 m 5 a2 5 x4 4w ----- ¢ e - ¢ 4 2 8 & 8 @ ® « 5 & 5 & 3 & a'e o » @ . w0 + 0 % . 0] ------ * & 8 6 & 3 @ © o w & @ 8 3 + e - e & 6] 4 = a 3 o 3 % a @ * & v o2 4 4 . v e 6] e A 2 e u - 2 s 8 » = . > & 3 4 s o 62 « v o & @ - » v =2 8 e a4 3 a 63 . - L ® 5 8 8 4 4 P 4 0 ® & & 3 o ¥ 63 L T T T T - s . . 63 4 ¥4 & 8 © 50 & 4 8 0 B A e o % 4 B ® B B g ; 64 4.3 8 4 4 & B ¢ o A & ® & e % 7 3 *+ e = w B 65 a % 4 & w 2 0 » & v ¢ 3 0 & % O & b o & 65 66 # P % 0 s 8 8 s @ s % s 8 D & a8 8 3 m & B 66 --------- - « oA 4 v 4 8 @ 66 » o a s » e o o 8 > . 5 66 LI L] s 4 s ? . s b & D B 66 a4 v @ * 2 e A e & m 9 8 o8 b @ v - . 68 s s & 2 a * ¢ & o8 2 e« ¥ 2 o * & o 70 . e "4 o % a3 B e . 2 a3 . - 72 . s 5 s & & 9 8 B . a o a . 72 o A+ 8 4 2 o ¢ @ » ® 4 5 o © & & > & o 74 2 ®w o ¥ a4 " s - - . . a @ ° 74 --------- * * 5 8 »# v oa .« a3 s 76 4 s s e B o oo . ° « e & - 77 4 o 2 % 8 . I ) » % p w8 e 77 + & 3 b A a . s » o e 2 8 & 3 » 77 ----- » a * 2 % p v & * s b . o ?7 a u v » # 0 4 % & 2 @ » L 77 FI a 7 3 & 8 & 3 ° & & v P e b e e = 79 b e " « - " oa e a2 3 s & 3 B » 79 ------- . a2 s L A 79 e e e e oo 19 ------- ° s » 2 % A e B B e @ 81 ooe 8 ® a2 @ A % 9 2 8 & w - a 84 ---------------- » s = ° 85 = 4 % & B & D S A4 % A W B o & a & e a s @ 88 LI T I a v w o # - . v . . s 89 . a s ° a o ®» b © o a o 4 8 o @ « » 9] CI ] @ * % 4 % & % ¥ F & o w » oa 8 B & 8o ow 9] f et e RN 21 ................ b e 93 ® 0 % o P & B A 3 A & B & & & & * s . a : 93 o . + & 3 s 2 8w s ° 08 & ¢ o b & o 93 “ s 4 & & o @ 4 ® & 3w o0 a & o 8 o . 93 o s e * & & 0 * * 2 3 s » s 8 W l 94 B r 4 s A . * * @ e @ e s 8 94 ......... . 97 e et e e s e e e cee . 99 e e e e C e e .. 99 --------------------- 2 ® 4 3 & o ¥ 4 . 99 F X Density and viscosity o o i it i i it i e s e e e e 99 Thermal conductivity o v ittt i ittt it e it ittt ettt e e e e 100 Electrical conductivity oo it i i i it it i et i s i e ittt e e e 100 Vapor PreSSUFES v v v v v v v v v v ot m st st b s s e e ae e e e et e e e 101 Fused-Salt Heat Transfer .. o0 it ittt i i i ittt ettt ettt ennenas 101 Reactor Hydrodynamics .o v o v vt i i i i i e e e e e e 102 Heat-Removal Study of BSF Reactor . .o i it it i ittt it i it ittt ettt e nnanenns 102 Heat Transfer in Circulating-Fue! Reflector-Moderated Reactor . .. o v v ot e i o vt i h e v e 103 Heat Transfer in NaOH-Moderated Circulating Fuel Reactor . ... ... .. vttt i v et vt 103 8. RADIATION DAMAGE . ... it ittt ittt ittt it ettt et n et anennoneneennses 105 Radietion Stability of Fluoride Fuels .. . oo i it it et ittt i it e e 105 LITR Fluoride-Fuel Loop v v v e i ittt i i it i it i ittt e tan st e 107 In-Pile Stress Corrosion and Creep « v v vt v it e s i it te ot ettt naneesonsonenneess 107 Remote Metallography v v i i i i il it it it i et ittt e it et enter et nnenns 107 9. ANALYTICAL STUDIES OF REACTOR MATERIALS .. .. ... it ittt it i i e e 109 Analytical Chemistry of Reactor Materials . ..o . it ittt ittt i ie v e ee e 109 Oxidation states of chromium and uranium in Nquanbase fuels . .. i i e 109 Determination of oxygen in NaZrF -bose fuels . ... ... . oo, 110 Oxidation of trivalent uranium by hexavalent uranium .. . . v i ittt i i ittt it e e s e 111 Stability of trivalent uranium in hydrochloric acid solutiens . . . ... oo it i it e i i i 111 Removal of film from lnconel tubing .. .. .. i i it i i it i it e et ianan 1 Determination of sulfur innatural gas . . o oottt i i it i i i e i et 112 Petrographic Investigations of Fluoride Fuels .. .. . i iiii i, 112 Summary of Sercvie Anolyses ... o it i i i e e i e st e e e 112 10, LID TANK FACILITY o i i i ittt it s ettt a et a s aanensenenens 117 Slant Penetrations of Neutrons Through Water .. .. . . ittt it ie et i et enn 117 Secondary Gamma-Ray Study . o oo it it i i i i e et e e e e e 118 11, BULK SHIELDING FACILITY .. ittt ittt it st e st tnnns st anaesonsnns 121 Gamma-Ray Air-Scattering Caleulations .. . o it ittt it i i it s e e e 121 Thermal-Neutron-Flux Perturbation by Gold Foils inWater . . ... ... it inen. 128 Reactor Power Calibration Techniques . ... vttt ittt ittt nnenrannens 128 Leakage-Flux Changes Due to Control-Rod Settings . . . v v v v vt i ittt et innennnmnnsen, 135 12. TOWER SHIELDING FACILITY .. i ittt i ettt e ittt e e ananeen 136 Experimental Program . .. 0o it i i i it i it i ittt e et et 136 Operation of the Reactor ... ... vivn . P e i e e e s et ee ettt 136 PART IV, APPENDIX 13. LIST OF REPORTS ISSUED DURING THE QUARTER .+ .. it ii i ittt et e e, 141 xii ANP PROJECT QUARTERLY PROGRESS REPORT SUMMARY PART I. REACTOR THEORY AND DESIGN The main pump for the fuel system of the Aircraft Reactor Experiment was installed, and it was therefore possible to start the first operational phase of the experiment — the water test of the fuel system (Sec. 1). The objectives of the water test are, in addition to c¢leaning of the system, determining effectiveness of the pressure fill of the system, checking tightness of the valves, determining the flow characteristics of the system, ascertaining the helium consumption rate, checking ability to transfer liquid (from one fill tank to another) while holding a fixed level in the pump tank, determining operability of the fuel enrichment system, and checking process instrumentation. The main sodium pump has also been installed and the stand-by fuel and sodium pumps will be installed soon. The fuel-to-helium and sodium-to- helium heat exchangers which had faulty welds have been refabricated and reinstalled. Facilities for loading ‘the sodium, fuel carrier, and fuel con- centrate, for sampling the fuel, and for unloading the fuel after the experiment have been or are being constructed. Operation of a large-scale test loop for circulating fluoride mixtures to test cor- rosion and structural stability of ARE components started June 5, 1954; the duration aim of the test is 2000 hr. The loop contains an ARE-type pump, a fuel-to-helium heat exchanger, and two reactor- core hairpin tubes. The experimental reacter engineering program (Sec. 2) has included the development of com- ponents for in-pile loops, the design of forced- circulation ‘corrosion-testing loops, and the con- struction of a unit for testing the mass-transfer characteristics of a sodium-beryllium-Inconel system. In addition a sodium-sampling device was developed, and a method for removing fluoride mixtures from equipment to be sclvaged was de- vised, Further tests of bearing and shaft-seal materials were also made. Development and hot testing of the vertical-shaft, down-flow sump pump for operation of the initial in-pile loop in the LITR were completed, and a model of an air-driven horizontal-shaft sump pump was fabricated. In additional development work on the centrifugally sealed pump, it was found to be possible to de- crease the size with no sacrifice in pump per- formance. - A compact heat exchanger for the removal of fission heat from in-pile loops is being fabricated. An lnconel corrosion-testing loop is being designed for circulafing fluoride mixtures that will provide high-velocity turbulent flow and large temperature differentials. The tests of the i-Mw, regenerative fluoride-to-sodium heat ex- changer were terminated and the test unit is being dismantled for examination. A 100-kw gas-fired furnoce is being fabricated to determine its suita- bility as a heat source for future heat-exchanger tests. If tests indicate that a heat source of this type will be satisfactory, a 1-Mw furnace will be developed. Components of the proposed &60-Mw Circulating- Fuel Reoctor Experiment (CFRE) are now being designed and constructed and are to be tested to determine operational characteristics (Sec. 3). De- tailed designs will be prepared from the data thus obtained. A stress analysis of the reactor is being prepared, and a series of charts has been con- structed for use in determining temperature distri- bution and thermal stress. A comparative analysis of lithium- and zirconium-base fuels was made that indicates ‘the superiority of the lithium-base fuels for reflector-moderated reactors, xenon-poisoning effect in the Reflector-Moderated Reactor have emphasized the need for removing the xenon during operation. Estimates of the An experiment is being planned to determine whether adequate purging of the xenon can be obtained. Calcu- [ations of a set of 48 reactors are being made that evaluate the effect of reactor dimensions or con- centration of U?33 in the fuel, on total U235 investment, on peak-to-average power density in the core, and on the fraction of thermal fissions in the core. : PART Il. MATERIALS RESEARCH The intensive studies of the fluoride systems of inferest as reacter fuels were continued with par- ticular emphasis on systems in which the vranium- bearing component is the less corrosive UF_ or a mixture of UF, and UF, rather than UF, alone (Sec. 4). It appears that attainable concentrations of UFB in Zer—based fuels are insufficient to fuel aircraft reactors; however, the use of UF, in ad- dition finF4 in such mixtures appears to be promising. The UF, is apparently sufficiently " @ | o ANP QUARTERLY PROGRESS REPORT soluble in NaF-KF-LiF mixtures (which would re- quire Li’7 for utilization) to make such fuels at- tractive; the wuse of NaF-RbF-LiF mixtures as solvents for UF, (without UF } has not yet been shown to be possible. The elucidation of the complex NaF-ZrF , system has virtually been com- pleted, and exploratory phase-equilibrium studies of the chloride-fluoride systems are under way. An apparatus for the visual observation of liquidus temperatures has been tested and found to be satisfactory. The chemical effects of fission products on aircraft reactor fuels are being ex- plored. The static, seesaw, and thermal-convection loop facilities were used extensively to further test the corrosion resistance of various materials in fluoride mixtures and in liquid metals (Sec. 5). It has been demonstrated in both seesaw and thermal-convection-loop tests that Inconel is not attacked by ZrF -based fuels containing UF, instead of the customary UF . Since UF, is not sufficiently soluble in NaF-ZrF | and several other fluoride mixtures of interest, tests are under way for determining the amount of UF, required to eliminate corrosive attack on lnconel by a fuel which contains UF, and UF . The effects of temperature and exposure time have been investigated further and additional con- firmation of the relationship of these variables to mass transfer in fluoride-Incone! systems was obtained. The amount of mass transfer increases with an increase in temperature or an increase in temperature difference between the hot and cold legs of a thermal-convection loop. The previously reported rapid initial attack and slower secondary attack after 250 hr of exposure of Inconel to NaF- ZrF ,-UF , were again demonstrated in a loop oper- ated for 5000 hr. The depth of attack increases about 4 mils for each 1000 hr of exposure. Reductions in depth of attack by UF4-bearing fluoride mixtures, in comparison with the attack on standard Inconel, were found in loops con- structed of Hastelloy B and of a special Inconel with a portion of the chromium replaced by mo- fybdenum. In both seesaw and thermal-convection- loop tests it was shown that, when Inconel and type 316 stainless steel are exposed in the same system to o fluoride mixture, the steel is inferior to Inconel in its resistivity to attack. Two type 316 stainless steel thermal-convection loops filled with high-purity lithium operated for 1000 hr with only small amounts of mass transfer and no indi- cation of plugging, in contrast to those operated previously, which plugged in 200 to 300 hr. The increased life was probably due to the higher purity of the lithium and, especially, to the de- crease in the lithium nitride content. In the study of corrosion and mass-transfer characteristics of materials in contact with liquid lead, it was found that quartz thermal-convection loops which con- tained types 410 and 446 stainless steel speci- mens had much longer life prior to plugging than similar loops which contained pure iron and chro- mium or one of the 300-series stainless steels. Studies are continuing on the identification of compounds produced by hydroxide-metal reactions. The metallurgical research effort has been de- voted to studies of the mechanical properties of Inconel in contact with fluoride mixtures, to in- vestigations of materials suitable for high-thermal- conductivity fins, to searches for container ma- terials other than Inconel for fluoride mixtures, and to the development of fabricational techniques (Sec. 6). Tests have shown that Inconel exposed to fluoride mixtures has o much longer rupture life when stressed under uniaxial conditions than when stressed under the multiaxial conditions that will prevail in reactor applications. Therefore tube-burst or other multiaxial-stress tests are to be emphasized in further studies. In previous investigations of materials suitable for high-thermal-conductivity fins for radiators, it was found that copper fins clad with types 310, 446, or 430 stainless steel or with Inconel were quite satisfactory in the unstressed condition if a suvitable diffusion barrier was provided; however, in oxidation tests under stress it was found that high stresses greatly increased the oxidation of the fin material. Experimental work was started on the preparation of high-boron-content material for use as shielding between the moderator and the heat exchanger and between the heat exchanger and the pressure shell of the Reflector-Moderated Reactor, High-temperature oxidation tests of several brazing alloys have shown that the majority of the nickel-chromium-base alloys are suitable for service in an oxidizing atmosphere at 1500°F and that several of the alloys retain this resistance at 1700°F. A new semiautomatic heliarc-welding process for the production of tube-to-header joints is described and its use in the construction of a prototype sodium-to-air radiator is illustrated. An assembly has been fabricated for use in de- termining the effect of thermal stresses and thermal cycling on beryllium metal and for studying the compatibility of sodium and beryllium. The physical properties of several fluoride mix- tures and other materials of interest to aircraft reactor technology were determined, and the heat- transfer characteristics of reocctor fluids were studied (Sec. 7). The enthalpies and heat ca- pacities of the ARE fuel (NaF-ZrF -UF,, 53.5- 40.0-6.5 mole %) and of K CrF were determmed Density and viscosity meosurements were made for molten RbF-LiF (57-43 mole %), and thermal conductivities were measured for molten RbF-LiF and solid NaF-KF-LiF (11.5-42.0-46.5 mole %). Electrical conductivity measurements were ob- tained on molten NaOH over the temperature range of 625 to 1490°F. A Lucite model of the circu- lating-fuel Reflector-Moderated Reactor was fabri- cated and is to be used to study the hydrodynamic structure in that system. The results of a mathe- matical analysis of convection are presented for the case of forced flow between parallel plates of Hluids with volume-heat sources; this analysis is useful in estimating the temperature structure in the flow annuli of refiector-moderated reactor cores. A study has been initiated to investigate the heat-transfer and fluid-flow characteristics of a NaOH-moderated circulating-fuel reactor system. The radiation domage program included ad- ditional irradictions in the MTR of fluoride mix- tures in Inconel capsules, construction of in-pile circulating-fuel loops, and development of creep- testing equipment for use in the MTR (Sec. 8). The Inconel capsules now being irradiated in the MTR contain UF,- or UF -bearing fluoride mix- tures so that additional comparative data can be obtained on the effect of UF; in decreasing the corrosiveness of fluoride mixtures. Temperatures of the capsules are being carefully controlled so that the Inconel-fuel interface temperature will remain approximately 1500°F throughout each test. A re-examination of a group of Inconel capsules from the earlier irradiations showed that previously reported excessive grain growth and deep pene- tration of the Inconel had occurred in only a few capsules and in those capsules the Inconel-fuel interface hdd been heated to temperatures muc:h higher than the desired ISDOOF ' PERIOD ENDING JUNE 10,.1954 The analytical studies of reactor materials in- cluded the problems of separating UF from U'i‘:4 in NquFs—b‘ase fuels and fuel solvents, the formu- lation of a method for determining oxygen in these mixtures, stability tests of trivalent uranium in hydrochloric: acid solutions, and petrographic ex- aminations of Zermbased fuels (Sec. 9). In the studies of the separation of UF, from UF4, methods were investigated for the conversion of the fluorides to chloride salts and the simul- extraction of the chlorides Petrographic examination of taneous info non- aqueous solutions. NaF-ZrF -UF, fuels reduced with metallic zir- conium, w:th mefqllac vranium, or with hydrogen has shown that the predominant phase in the re- duction complexes obtained is alwagys a solid solution of U4 and U3 ¥ in Nog Zr F 8 471° PART Hl. SHIELDING RESEARCH The Lid Tank Facility has been used primarily for studying special attenuation problems which arise in aircraft shield design (Sec. 10). The crew-shield plastic sides attenuate nevtrons which arrive at slant incidence, and therefore experi- ments are being carried out to measure the effecy of slant incidence on the attenuation. At 60 deg to the normal, the short-circuiting effect becomes quite important. The gamma-ray attenuvations in lead and bismuth have been compared as a function of the neutron flux at the metal in a metal-water shield. A first study of the data shows little difference between the two materials, but more work is to be done to clarify this point. The work at the Bulk Shielding Fqcrhfy has included gomma-ray air-scattering calculations, the determination of the thermal-neutron-flux pertur- bation by gold foils in water, a study of reactor power calibration techniques, and a determination of the effect of control-rod settings on leakage flux (Sec. 11). The calculation of air scattering for the divided shield with a lead shadow shield was carried out some time ago for gamma rays by using the spectral data obtained on the reactor shield mock-up at the Bulk Shielding Facility, The data have now been extended to include at- tenuation by the crew shield so that a direct comparison is possible with earlier Shielding Board calculations. For scattered radiation there is good agreement, but the leakage around the shadow shield was badly underestimated in the ANP QUARTERLY PROGRESS REPORT early work., This will require redesign of the shield configuration but will not introduce an ex- cessive weight penalty. The correction for flux depression by detector foils is known for indium foils, but since much work is done with gold foils, the determination of flux depression in water has been extended to this element. In connection with intercalibration of the several shielding reactors (Bulk Shielding Facility and Tower Shielding Fa- cility reactors at ORNL and a reactor at Convair), it has been demonstrated that bare-foil activation gives a reliable measure of reactor power density. The effect of sofety and control-rod positions on fast-neutron leakage from the reactor has been detetmined experimentally for use in correlating data taken on two otherwise similar reactors. Fortunately, the effect appears to be small, as is the effect of small changes in fuel loading. The Tower Shielding Facility reactor was made critical for the first time at the site on March 12 (Sec. 12), Since that date much time has been spent in determining the operational character- istics of the equipment, and the reactor now oper- ates regularly in the air at powers of up to 4 kw. It will soon be operated at up to 100 kw, the design maximum power level, Part | REACTOR THEORY AND DESIGN 1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT E. S. Bettis J. L. Meem ANP Division The first operational phase of the Aircraft Re- actor Experiment is now under way. With the installation of the main fuel pump, it was possible to start the water test of the fuel system. The water was charged to a fill tank and then forced by gas pressure into the system. Subsequent circulation of the water by the fuel pump carried the gas from pockets in the system, which will not pressure fill, to the pump tank where the water degassed. The system probably became gas free, that is, full of liquid, but since there is no posi- tive indication of fullness, vacuum filling of the system is being considered. Data on the pressure and flow characteristics of the system are being obtained. The main sodium pump has also been installed and the stand-by fuel and sodium pumps will be installed soon. The fuel-to-helium and sodium-to- helium heat exchangers which had faulty welds have been refabricated and reinstalled. Minor modifications were made during refabrication that are expected to improve the performance of these heat exchangers. The thermal-barrier doors, which gre lowered to permit preheating of the heat ex- changers to above the melting point of the fuel or the sodium, were modified to eliminate thermal buckling. Operation of the helium supply was checked, and the arrangements were made for assuring an adequate supply of helium for oper- ation. The original fuel-injection system, which was dependent on the operation of valves, was modified to permit injection of the fuel concentrate hy a more reliable gas-pressure equalization tech- nique. Facilities for loading the sodium, fuel carrier, and fuel concentrate, for sampling the fuel, and for unloading the fuel after the experi- ment have been or are being constructed. Plans hove been made for a leak check of the fuel system at operafting temperature. Helium spiked with krypton will be introduced into the fuel system and the helium in the annulus around the fuel piping will be checked for krypton. A large-scale test loop for circulating fluoride mixtures is being operated to test corrosion and structural stability of ARE components. The loop contains an ARE-type pump, a fuel-to-helium heat exchanger rebuilt from one originally purchased for the ARE, and two hairpin tubes purchased as spares for the ARE reactor core. The system is operating isothermally at about 1375°F with a flow rate of 20 gpm. The duration aim of the test, which started June 5, 1954, is 2000 hr. THE EXPERIMENTAL REACTOR SYSTEM W. B, Cottrell J. Y. Estabrook J. K. Leslie ANP Division G. J. Nessle J. E. Eorgan Materials Chemistry Division G. A, Cristy E. Wischhusen Engineering and Mechanical Division Pumps. The main fuel pump and the main sodium pump have been installed in their respective systems. The stand-by sodium pump, which is on hand, and the stand-by fuel pump will be installed in the immediate future. All these pumps have been satisfactorily hot checked with sodium at 1300°F for 100 br.. ' Operation of the main fuel pump with water re- vealed that the maximum oil flow in the lubricating system was too low (~ 2 gpm) at high pump speeds (~ 1800 rpm). Consequently, the auxiliary lubri- cating pump, now a single-phase Chempump, will be paralleled with a three-phase Chempump, and the single-phase pump will be heid in reserve. Tests with the three-phase Chempump show thot the oil flow is ample (>3 gpm) at all speeds of the primary pump. : Heat Exchangers Disassembly, refabrication, and reinstallation of the fuel-to-helium heat exchangers and the sodium- to-helium heat exchangers which had faulty welds have been completed. The welding was com- pleted by the two qualified ARE welders dhead of schedule. There were two significant design changes in the fuel heat exchangers as refabri- cated. FEach fuel heat exchanger now has three parallel passes, rather than two, in series with ANP QUARTERLY PROGRESS REPORY three similar paralle! passes. Also, a small by- pass line was eliminated to simplify fabrication of the fuel heat exchangers. Without the bypass fines, however, the heat exchangers are not com- pletely drainable. Two insulated and heated barrier doors are pro- vided for each heat exchanger to permit preheating of the heat exchanger to above the melting point of the fuel or the sodium before these liquids are loaded into their respective systems. These doors, which are lowered within the closed duct to block the flow of helivm through the heat ex- changers, were tested at operating temperature (1300°F) before the heat exchangers were removed for refabrication. The doors stuck at temperatures above 900°F, and therefore they have been modi- fied to eliminate thermal buckling. These doors have not yet been retested at operating tempera- ture because of the water tests now in progress. Fluid Circuits The major fluid circuits (fuel and sodium) have been virtually completed for some time. The re- welding of some pipe lines necessitated by the removal and reinstallation of the heat exchangers has been completed. Except for the bypass around the reactor in the sodium system, all fuel and sodium piping is completely welded, instrumented (thermocouples located every 3 ft), heated, and insulated. The heafer electrical connections are essentially complete, but a substantial number of thermocouple connections remain to be made. An operational check of the helium supply system disclosed o few leaks which have been repaired. The helium supply problem has been resolved. Three railroad tank cars have been allocated for ARE use so that a helium tank car should be on hand ot all times. Arrangements have been ef- fected to have two helium trailer trucks (capacity 20,200 scf) on hand at all times to transport the helium from the tank cars to the ARE building. The bank of 12 helium cylinders (capacity 2640 scf) held in reserve at the experiment will be supplemented by a bank of 15 larger cylinders (capacity 16,000 scf), which are now being fabri- cated. Fuel-Enrichment System The original fuel-enrichment system provided for the injection of the fuel concentrate into the system below the liquid-gas interface in the pump tank. Since such an arrongement is inherently dependent upon the operation of the system valves, freeze valves were inserted in the ¥%-in. transfer lines to back up the bellows valves in the system, which are of questionable reliability at the temperatures involved (1300 to 1400°F). These freeze valves were merely sections of tubing from which the heat could be removed so that the liquid fuel inside the tubing would freeze. However, in tests of similor freeze sections the 3/8-in. tubing frequently ruptured upon thawing, and this injection technique could not be con- sidered sufficiently reliable. It was therefore decided to effect transfer of the fuel concentrate into the fuel system by a gas-pressure equalization technique. In order to employ this technique, only two modifications to the existing system were necessary. First, the transfer line had to be installed so that the fuel concentrate would be injected in the pump above the liquid-gas interface, and second, gas equalizer lines, both with gas valves, had to be provided from the storage tank to the transfer tank and from the transfer tank to the pump. The concen- trate will now be injected through the pump flange into the pump tank. Since the pump flange is maintained at 600°F, it was necessary to develop a special resistance-heated annular fitting which could attain a temperature of 1400°F in the center fuel passage. The transfer tank, originally fabricated of type 316 stainless steel, has been refabricated of In- cone! to conform to the rest of the system., The fuel-enrichment system will be tested with water as soon as the transfer tank is installed and connected. Loading Facilities Facilities for loading the sodium, fuel carrier, and fuel concentrate into their respective con- tainers have been, or are being, fabricated. Each of these facilities will effect transfer of the liquid from a portable container into the tank provided for it at the ARE. The transfer will be effected by gos pressure in each case. The sodium, which will assay less than 0.02 wt % oxygen, will be received in four 55-gal drums. The fuel carrier, NaZer, will be received in 13 cylinders, each containing approximately 250 Ib of the fluoride. The fuel concentrate, Na,UF,, will be received in 15 cylinders, each weighing approximately 75 1b. All these materials are on hand. In all cases the fluid to be trans- ferred will be maintained under helium pressure. The transfer temperatures are 600, 1200, and 1300°F for the sodium, carrier, and concentrate, respectively. ' : .FueI-Scmpling Facility It is anficipated that from 5 to 10 samples of fuel will be drawn from the ARE between the time the fuel system is filled with fuel carrier and the time the reactor is operated ot an appreciable power level. For the chemical analyses to be performed on these samples, only 10 to 20 g of material are required; however, to ensure that the sample is actually representative of the flowing fuel, at least 300 to 400 g of material must be withdrawn. Equipment for accomplishing this sampling operation has been designed, fabricated, and successfully tested. ' The sampling apparatus consists, as shown in Fig. 1.1, of a nickel pot containing a machined graphite liner and a flanged top through which PERIOD ENDING JUNE 10, 1954 pass the fluoride entrance tube, three electrical contacts, the gas manifold connection, and a lever for moving the grophite-lined sample cup under the flucride entrance tube. This apparatus will be connected to the system by a heated 3{g-in. Inconel line containing a bellows valve and will be placed in the reactor pit at a level 4 to '8 ft above the highest fuel level in the ARE system. With the sampling cup out of the fuel delivery stream, a vacuum will be drawn on the nickel pot; when sufficient fuel has been drawn into the graphite liner to contact the lowest probe, the sample cup will be swung under the delivery When the sample cup has overflowed sufficient fuel to contact the second probe, an atmosphere of helium will be admitted to stop the flow of fuel. As soon as the apparatus is suf- ficiently cool to handle, it will be disconnected and replaced by an identical apparatus. It is anticipated thot sampling can be accomplished without closing the valve in the sample line and without resorting to the use of freeze valves in the delivery lines. ' stream. Fig. 1.1. Fuel-Sompiing Apparatus. ANP QUARTERLY PROGRESS REPORT Four complete tests of the apparatus have been made with four different values for pressure above the fuel and inside the sampling apparatus. These tests, in which the sampling apparatus was not heated, were satisfactory; it is anticipated that, unforeseen complications in the ARE, the fuel! circuit should proceed barring sampling of smoothly. Fuel-Unloading Facility The radioactive fuel mixture will be discharged from the hot-fuel dump tank through a l-in. pipe to aluminum cans for removal to fuel recovery and reprocessing equipment. A mock-up of the unloading equipment was constructed and tested with the fuel carrier NaZrF . The feasibility of unloading the ARE with equipment of this type has been satisfactorily demonstrated., Reactor Control The control actuator assembly has been mounted over the reactor, and the shim and confrol rods have been installed obove the reactor. Accurate measurements have been made from the bottom of the reactor to assure proper location of the rods. The fission-chamber circuits have been checked, and the chambers are now being installed. In order to prevent a chamber from catching where the sleeve changes from 3 in. to 2]/2 in. in di- ameter, a tapered transition section will be pro- vided. Except for the final check on the ion chambers, the nuclear instrumentation has been completely checked. The actual installation of the ion chambers and shielding plugs will be accomplished just before the fuel carrier is loaded into the system. FUEL SYSTEM MOCK-UP TESTS H. W. Savage A. G. Grindell W. G. Cobb W. R. Huntley ANP Division The complete filling of the ARE fuel system would be most easily accomplished with the system under vacuum, but the problems involved in obtaining and holding vacuum are formidable, Therefore, a mock-up was constructed and oper- ated to test filling and degassing characteristics under a helium pressure greater than atmospheric. The mock-up was also used to test the drainability of the circuit. The test unit was an essentially full-scale mock-up of the ARE fuel system without the heat 10 exchangers. Included were a reactor core, a pump, a fill tank, piping, and valves with which to dupli- cate ARE opressure drops. The fill tank, core, and pump were set ot levels corresponding to the respective levels in the ARE building. The working fluid used was tetrabromoethane becouse its room temperature density and viscosity (sp gr = 2.95 un =93 cp at 77°F) are similar to those of the fluoride fuel at operating temperature. A helium blanket at 20 to 30 psig was maintained in all tests, The primary objective of the tests was to quanti- tatively determine the flow rate required to com- pletely fill the reactor core and sweep all gas bubbles from it. A secondary objective was a determination of the degassing and self-priming characteristics of the pump. In all the tests, flow rates of 60 to 64 gpm were required for filling all six parallel circuits in the core and sweeping out all gas bubbles. Since the tubes in the mock-up were 1.00 in. in inside diameter and the ARE core tubes are 1.115 in. in inside diameter, it is estimated that flows of 85 gpm may be required for filling the ARE. In the mock-up it could be ascertained by visual observation that all gas had, been removed from the core, but visual observation will not be pos- sible in the ARE. The pump degassed well in all tests. In no case was there sufficient surface turbulence to endanger the probes or gas lines by splashing (the liquid surface was visible through glass portholes in the pump casing). In each filling operation, the pump lost its prime several times but regained it each time without adjustments of the controls, In the drainability tests, gravity draining unas- sisted by gas pressure resulted in approximately 25% removal of liquid from the reactor core. Helium-pressure-assisted drainoge (15 psi differ- ential between pump and dump-tank pressures) re- sulted in @ maximum of approximately 60% removal. OPERATION OF THE FUEL SYSTEM R. G. Affel W. B, Cottrell G. D. Whitmon ANP Division The installation of the main fuel pump es- sentially completed the fuel system, and the water test of the system is now under way. While this represents attainment of the first operational phase described in ‘the ““ARE Operating Procedures,””’ the immediate objectives of the water test are (1) cleaning of the system, (2) determining ef- fectiveness of the pressure fill of the system, (3) checking tightness of the valves, (4) de- termining the flow characteristics of the system, (5) ascertaining the helium consumption rate, (6) checking ability to transfer liquid (from one fill tank to another) while holding a fixed level in the pump tank, (7) determining operability of the fuel-enrichment system, and (8) checking process instrumentation. Cleaning : In order to remove the scale of calcium meta- silicate now deposited throughout the system as a result of previous tests with water and Conklene (o detergent — sodium metasilicate), the system will be washed with o 1% solution of the tetra- sodium salt of ethylenediaminetetroacetic acid in water. According to the analytical results (cf. Sec. 9, "Analytical Studies of Reactor Materials’’), this solution will dissolve the calcium deposits and leave no residue when it, in turn, is flushed out of the system. |t was recommended that the solution be circulated for 4 hr in the system at 170°F. Furthermore, because of the significant holdup volume of the system (about 1 cu ft of each 6 cu f1), the rinsing step should be repeated six or seven ftimes to ensure complete removal of the washing agent. To date, the initial cleaning and rinses hove been effected, and the water shows the expected decreasing concentration of impurities. = The water for the second rinse is being used in the operational tests described above., ' two Pressure Filling The water was first charged to the fill tank and then forced into the system by gas pressure. Al- though the system (reactor and heat exchanger) contained gas pockets which would not pressure fill, subsequent circulation of the water by the fuel pump carried the gos pockets around to the pump tank where the water degassed. Although it is probable that the system thus became gas free (full of liquid), there does not appear to be e —ma— e IW. B. Cottrell, ARE QOperating Procedures, Parts [, 1, and 1H, ORNL C¥F-54-2-68 (Feb. 11, 1954). PERIOD ENDING JUNE 10, 1954 any positive indication available, The gas re- maining in the system is of little consequence in water operation, but it would be operation with fuel. Unless operational pa- rameters permit a more positive defermination of the extent’' of filling, vacuum fill the system. intolerable in it may be necessary to Vacuum filling of the system has thus far been avoided because oxygen could conceivably enter from the vent system and contaminate the hot fuel. ‘ Pressure and Flow Characteristics Many data have been taken on the pressure and flow characteristics of the fuel system. It is apparent that the flow through the bypass (in parallel with the heat exchangers) is less than that through the heat exchangers. This is un- doubtedly caused by the pressure drop across the bypass throttling valve being greater than an- ticipated and that across the hear exchanger being less. The pump speed and system pressure-drop data do not check with the system flow data. |t is most probable that the flowmeter (a Rotameter which fransmits a signal from a coil within which moves a tapered iron core attached to the Ro- tameter bob) is not calibrated correctly. This possibility is being checked. The Rotameter has shown a tendency to drift, and it has not been - possible therefore to moke occurate measurements of the system flow chamcterfisfics. Functional Tests In the operation of the system to date, the dump- line valves appear to be tight, although there is some indication that the valves to individual tanks leak slightly under pressure. Such slight leaks are tolerable, although not desirable, and the leak tightness of the valves may improve upon repeated operation. The helium consumption rate during the test with water in the fuel system has been about 3 cfm. It is therefore probable that the maximum helium consumption at any time during the ex- periment will be about 10 ctm. It was found to be possible to transfer water from one fill tank to another through the system and to maintain o fixed liquid level in the pump. This operation could be of significance should it be desirable to replace the initial batch of carrier and/or to replace the hot fuel at the con- clusion of the test,’ 11 ANF QUARTERLY PROGRESS REPORT Hot-Gas Test Although it would be desirable to leak-check both the fuel and the sodium systems at operating temperature, there is no completely satisfactory method for doing this except with fuel and sodium because the systems are not completely drainable. However, the inadequacy of u cold check in com- parison with a hot check is so apporent that a gas check of the fuel system at the operating tempera- ture will be effected. Basically this test will consist of heating the system to 1200°F, filling the fuel system with one gas under pressure, tilling the fuel-system annulus with another gas at lower pressure, and spectrographically ana- lyzing the gas in the annulus for the presence of the gas from the fuel system. Although in prin- ciple any two gases would suffice for this test, it would be impossible, for example, to detect a small leak of helium into nitrogen because of the background of acir in the annulus system. Ac- cordingly, helivm spiked with krypton will be used in the fuel system, and the helium in the annulus will be examined for krypton. PUMP FABRICATION AND TESTS H. W. Savage W. G. Cobb A. G, Grindell W. R, Huntley R. E. MocPhersen ANP Division P. Patriarca G. M. Slaughter Metallurgy Division All six ARE-pump rotary elements have been assembled and were found to meet hot-test re- quirements in tests made by using the two im- pellers which have been accepted. Four of the six rotary elements have passed all acceptance tests.2 The other two rotary assemblies were rejected because of unsatisfactory upper seals which permitted excessive gas leokage. Re- lapping of the seals resulted in no improvement because the lapped surfaces were scored in sub- sequent testing. It was thought that the lapping abrasives (1500-grit diamond dust, and others) were not being completely removed from the soft nose of the upper seal. As o check on this possi- bility, the adequacy of lapping with water-soluble 24, W, Savage et al., ANP Quor. Prog. Rep. Mar. 10, 1954, ORNIL-1692, p 10. 12 scouring compounds (for example, Bon Ami) is being evaluated. One of the accepted elements, with a cast im- peller {to be exchanged for a fabricated impeller before start-up of ARE), has been installed in the fuel system, and one was installed in the sodium system. One accepted element has been assigned to K-25 for use in o heat-exchanger test. One unaccepted element is installed in the cold shake- down test unit for seal tests., The other unac- cepted zlement is being used in the hot shakedown test unit for hot-testing fabricated impellers. Two fabricated impellers have passed all acceptance tests,® and one has been furnished to K-25 for the heat-exchanger test, Shorting of the probes used for liquid-leve! indi- cation in the sodium pump has shown the clear- ance between the riser wall and probe to be inadequate. The original design called for a 3/8-in. schedule-40 pipe riser which allowed bridg- ing of sodium condensate between the inner pipe wall and the 3/32-in.—dia probe., The size of the riser was increased to 1/2 in, and no shorting of the probes occurred during subsequent festing. All the sodium pumps have been modified to utilize ]/:’L,-in. risers. The operation of an ARE-type d-c pump drive motor in a dry helium atmosphere continued un- eventfully during the quarter. Approximately 3500 hr of operating time have been accumulated at a test temperature of 130 to 140°F. This test will be terminated at 4000 hr, The detail and assembly drawings of the pumps, made in the fall of 1953, are being revised to incorporate the modifications in the pump cooling and lubricating systems and other modifications. The fabrication of five [nconel impellers for ARE pumps by machining and welding has been completed. The rough-machined parts are shown in Fig. 1.2. These parts, except the vanes, were subjected to a stress-relief annecaling heat treat- ment so that precise dimensional tolerances could be maintained during subsequent machining. The stress-relief thermal cycle consisted of heating at 125°C/hr to a temperature of 1000°C, main- taining this temperature for l/2 hr, and cooling at 125°C/hr to room temperature. The drive-shaft hub was then tack welded to a carbon steel strong-back to prevent warping during welding, and the vanes were heliare welded 3ibid., p 11, PERIOD ENDING JUNE 10, 1954 into place, as shown in Fig. 1.3. The stress- sition, and the joints between the hub and the relief heat treatment was then repeated. After vanes were heliarc welded as far as they were the vanes had been machine finished, the fluid- accessible, as shown in Fig. 1.4, The stress- entrance hub was heliarc plug-welded into po- relief heat treatment was again repeated. ' The %, D ;w. i o il . AN AN D Yoy Fig. 1.3. Impeller After Heliarc Welding of ‘ Yanes to Drive-Shaft Hub. Impeller is tack welded Fig. 1.4. Impeller with Fluid-Entrance Hub to carbon steel strong-back., Plug-Welded into Position, 13 ANP QUARTERLY PROGRESS REPORT welded impeller was then removed from the strong- back and machine finished. The completed as- sembly is shown in Fig. 1.5, UNCLASSIFIED Y107 9a Fig. 1.5. Completed Impeller. REACTOR SYSTEM COMPONENT LOOP H. W. Savage A. G. Grindell W. G. Cobb W. R. Huntley ANP Division The services of the K-25 Technical Division have been secured for constructing a large-scale system for circulating fluoride mixtures and testing reactor-system components. The major components of the system are the pump (mentioned above), a fuel-to-helium heat exchanger rebuilt from one originally purchased for the ARE, and two hairpin tubes purchased as spares for the ARE reactor core (Fig. 1.6). The system is operating iso- thermally at 1375 + 25°F, with o flow rate of 20 gpm. The duration aim of the test, which started June 5, 1954, is 2000 hr. When the oper- otion is terminated, the heat exchanger and hairpin tubes will be examined and evaluated in terms of corrosion, structural stability, etc. FUEL RECOYERY AND REPROCESSING D. E. Ferguson G. |, Cathers Chemical Technology Division The chemical process currently being con- sidered* for recovery and decontamination of 235 from the ARE fuel mixture NaF-ZrF ,-UF , consists of dissolution of the fuel in an aqueous aluminum nitrate—nitric acid solution, solvent extraction with 5% tributyl phosphate in a kerosene diluent, 4p, E. Ferguson, G. I. Cathers, and O. K. Tallent, A!‘]J.g] Quar. Prog. Rep. June 10, 1953, ORNL-1556, [ . 14 and isolation on an ion-exchange resin column. In one solvent-extraction c¢ycle a gross beta- decontamination factor of greater than 10* and vranium recovery of greater than 99.9% will be obtained. Additional work has now been completed on dilute-TBP extraction of uranium solutions having high nitrate-salting strength, In batch counter- current extraction experiments with six TBP con- centrations in the range 1 to 30 vol %, with feed nitric acid concentrations of 0.5 and 3 M, ru- thenium and total-rare-earth beta-activity decon- tamination factors improved with decreasing TBP concentration in the solvent (Fig. 1.7). The anomalous zirconium ond niobium beta decon- taminations obtained affected somewhat the gross- beta decontamination factor determinations in the low TBP concentration range. However, the best gross beta decontamination was obtoined at a TBP concentration of 7.5% or less regardless of acidity. There was some evidence of o maximum in gross beta decontamination at 5 to 7.5 vol % TBP with a feed acidity of 0.5 M. Maximums were also observed in the zirconium and niobium decon- tamination factors at a low acid concentration. Three extraction and scrub stages were used, with stoichiometrically neutral aluminum nitrate solution as the scrubbing agent. The salting strength was controlled by varying the aluminum nitrate content to obtain a uranium distribution coefficient of about 4 at the feed plote and 0.5 at the last scrub stage. Figure 1.8 is a plot of the distribution coefficients of the various ac- tivities at the feed plate. These data could be correlated, in general, with the conclusions drawn from Fig. 1.7. Testing of the data by log-log plots (Figs. 1.9 and 1.10) showed that the ruthenium and total- rare-earth distribution coefficients have close to second-order dependence on TBP concentration at constaont uranium distribution, The equilibrium constant for this case is represented by the equation D'C"(Ru Bor TRE f) (TBP conc)”® where n, the dependence, is equal to 2. A least- squares analysis showed the average dependence for all data in the case of ruthenium activity to be 2.10 and in the cose of total rare earths to be 1.95. PERIOD ENDING JUNE 10, 1954 UNCLASSIFIED ORNL-LR-DWG 1688 PUMP Lo PUMP TANK O PIPE = sosmafin- I = ! ~ HEAT EXCHANGER D) iy VENTURI W HAIRPIN TUBES H ! Il FiLL AND DRAIN l1 TANK B Fié. 1.6. Fuel#o-Helium Heat-Exchanger Test Loap. 15 ORNL-LR-DWG 1683 ANP QUARTERLY PROGRESS REPORT ORNL.- L'-DWG 1620 ) % | 0] 3 CONCENTRATION OF TBP IN AMSCQ 125 cs , - [ Lozr B 1 10 0.5 M NITRIC ACID —— 3.0 ¥ NITRIC ACID CONCENTRATION OF TBP IN AMSCO 125-90W (vol % Fig. 1.7. Decontamination vs TBP Concentration in Solvent Extraction. B = =+ GROSS S -— 801074 NOILYNIWVYLINOD3Q %) -3QW {vol Fig. 1.8. Activity Distribution Coefficients at Feed Plate vs TBP Concentration in Solvent Extraction, 16 PERIOD ENDING JUNE 10, 1954 2 . ORNL-LE-DWG 1692 10" oot I 1T - 2 g ool AR I - = o F e L i@ L (o 05 M NITRIC ACID -3 e, 30 M NITRIC ACID % = Sz . z 2 88 S P & 110 g 0 ¥ a0 e T R AT T o & H [ £E= 5 CootnInh . LT Tl A Bo 5 e T e e &g : . LT 8 g < ; - /] - T 05 M NITRIC ACID | = oo - — o = O % : o | o gl % Q% : 5 % W - = “ o R < g" 5 XL o = < < (s1] O - b= G xI Hot-pressed BeC 29 27 X Sliminated BeO-Ni X X wC X 21 30 25 23 24 31 28 A|203 X X Eliminated AIZOB-MQO X X BAC X X Cr3C2 X 22 X 33 18 19 TiC-Ni X 20 X 26 TiC-Co X X Graphitar X X _High-density graphite X X *The numbers are test serial numbers. not yet tested or results not yet available. ROTARY-SHAFT SEALS ¥W. C. Tunnel! P. G. Smith ANP Division Tests are under way to determine the feasibility of sealing fluoride mixtures in an in-pile pump with a packed seal on a horizontal shaft. One packing that was tested consisted of o mixture of 95 wt % Asbury Graphite and 5 wt % MoS,, re- tained by bronze wool at each end of a stuffing box. The stuffing box was 3!/2 in, long with a ]/4-in. annulus., For this test, the apparatus was operated for a total of 382 hr, and the test was then terminated because a heater burned out as a result of leakage of the fluoride mixture. Leak- 26 The notation X indicates unsatisfactory combinations. Blank spaees mean age from this seal (test No. 33)® was at the rate of 0.7 cc/day throughout the test. Postrun ex- amination revealed only slight shaft wear at re- gions where the packing retainers were located and in the glond region. The pot showed no sign of packing leakage from the hot end of the seal, In another test of the same packing material (test No. 35), an inert blanket was installed on the cold end of the seal to protect the section of the shaft under the gland from wear by oxidation products, A pillow-block beuring was installed on 8w. C. Tunnell et al., ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 26. the 1/z-in. shaft to improve alignment, The type 316 stainless steel shaft was coated with ¥ -in.- thick Colmonoy ground to a 20-gin. finish, The test apparatus was operated for 603 hr with NaF- ZrF -UF, at a temperature of 1250 to 1300°F and with a pressure of 2 psi on the seal. The leckage rate from the seal was about 0.5 cc/day and ap- peared to be mostly packing material. The power requirement of the seal varied from 100 to 200 watts, Termination was caused by shaft seizure. Postrun examination revealed shaft wear of about the same magnitude as in test No. 33. A section of the shaft surface at the hot end appeared to have peeled off or to have corroded slightly. Analysis of a sample from the pot showed that - PERIOD ENDING JUNE 10, 1954 an unsatisfactorily large amount of the packing material had ledked inte the pot. There was evidence of corrosion on the hot end of the gland, which indicates a need for an inert-gas bianket between the gland and the stuffing box, os well as between the gland and the shaft, Another test is planned for the near future with an improved inert-gas blanket and improved gland alignment. Seal test No. 31 was the same as seal test No. 28,% except that Asbury Graphite was substituted for the National Carbon Graphite in the 10 wt % BaF ,~90 wt % graphite packing material, The results obtained were very similar to those for test No. 28 in that excessive leakage occurred. The duration of the test was 165 hr, 27 ANP QUARTERLY PROGRESS REPORTY 3. REFLECTOR-MODERATED REACTOR A. P. Fraas ANP Division Components of the proposed 60-Mw Circulating- Fuel Reactor Experiment (CFRE) are now being designed ond constructed ond are to be tested to determine operational characteristics. Plans are being made for tests of full-scale fuel pumps and of elements of full-scale heat exchangers. Also, thermal-stress cycling tests for both Inconel and beryllium are to be performed. A fairly complete detailed stress analysis of the reactor is well under way. A series of charts was prepared to facilitate the determination of the temperature dis- tribution and thermal stresses in media having uniformly distributed volume-heat sources.! Also, a comparison of lithium- and zirconium-base fluo- ride fuels for aircraft reactors was made that indicates the definite advantages to be gained through the use of the lithium-base fuel. Cstimates of the xenon-poisoning effect in the Reflector-Moderated Reactor have confirmed the great incentive for removing the xenon during operation, Therefore an experiment is being planned to determine whether the xenon can be adequately purged. Studies of reactor dynamics indicate that the design of the Reflector-Moderated Reactor precludes the possibility of the coupling of mechanical oscillations with nuclear oscil- Ve, A, Field, Temperature Gradient and Thermal Stresses in Heat Generating Bodies, ORNL CF-54-5-196 (May 21, 1954). lations to create antidamped oscillations that would destroy the reactor. Calculations of a set of 48 related reactors are weil under way. The effects of reactor dimensions on concentrotion of U233 in the fuel, on total U235 jnvestment, on outside peak-to-average power density in the core, ond on the per cent thermal fissions in the core ore presented. A COMPARISON OF LITHIUM- AND ZIRCONIUM- BASE FLUORIDE FUELS A key decision in the design of a circulating- fuel reactor is the choice of the fluoride fual to be employed. Of the fuels that cuirently seem most promising, the sodium-zirconium-uranium fluo- ride fuel has the poorest physical properties but would be the easiest and cheapest to prepars, while the sodium-lithium-potassium-uranium fluo- ride fuel has the best physical properties but would require Li’ — an expensive material. The properties of greatest interest for these two fuels for vranium concentrations of 25 Ib/cu it are given in Table 3.1. In a comparison of the two fuels it is evident that the lithium-base mixture has both a lower melting point and a much lower vapor pressure. The lower vapor pressure is particularly advan- tageous in that it should eliminate the many troubles experienced with sublimation of Zrl:4 TABLE 3.1, PHYSICAL PROPERTIES OF TWO FLUDRIDE FUELS CONTAINING PHYSICAL PROPERTIES 25 b OF U23% PER CUBIC FEET OF FLUORIDE MIXTURE AT 1472°F Melting point, °F Vapor pressure at 1500°F, mm Hg Prandtl number, CP,U/K | Viscosity,* cp Thermal conductivity,* Btu/hr-sq ft (OF /ft) Specitic heat,* Btu/lb Density,* Ib/cu ft 28 chF-ZrFA-UF4 Nc:F-LiF-KF-UF4 (50-44.5-5.5 mole %} (11-45-41-3 mole %) ....... e s 970 850 8 <1 3.71 1.24 6.5 2.5 1.25 1.95 0.295 0.40 L 203 132 and the formation of high-melting-point deposits on the roofs of expansion tanks, in pressure gage lines, etc. The higher viscosity of the zirconium- base fuel gives a lower Reynolds number in the reactor core and hence a more severe boundary- layer heating problem; the temperature differential between the peak temperature in the boundary layer and the mean fuel temperature is about 50% greater than that for the lithium-base fuel.?2 Thus the zirconium-base fuel, with .its higher viscosity and its much higher vapor pressure, would be more likely to give trouble with boiling in local hot spots ond hence, conceivably, violent fluctu- otions in power. ' Recent multigroup calculations show that for a representative core geometry the critical mass for a lithium-bose fuel will be about 16% greater thon that for a zirconium-base fuel because of thermal- neutron absorption in the potassium, The reactor under consideration had an 18-in.-dia core with a 12-in.-thick reflector and a 4-in,-thick fuel an- nulus, and the beryllium in the island and reflector was canned in 0.010-in.-thick Inconel. The core shells were Y -in.-thick Inconel. A further consideration in favor of the alkali- metal fluoride fuels is based on radiation damage; the alkali-metal fluorides are very stable com- pounds, and their recombination rate con be ex- pected to be larger than that for ZrF . One of the most important design criteria is the effect of the physical properties of the fuel on the heat exchanger. While many design compromises are possible, the most important of these can be deduced from Fig. 3.1, which ‘was prepared for a series of fuel-to-NaK heat exchangers. The pres- sure-stress limitations are usually controlling, ond if the same allowable stresses are to be used, the pressure drop must be held constant. The designs represenfed by Fig. 3.1 were prepared by determining the number of tubes required for each tube length to give a 50-psi pressure drop through the tubes in the NaK circuit. The tube spacing was then adjusted to give a 40-psi drop in the fuel passages. As shown in Fig. 3.1, the number of tubes in the heat exchanger must be increased to provide more flow passage area as the tube length is increased. - The curves for the temperature differential be- 2y, E. Poppendiek and L. D. Palmer, Forced Con- vection Heat Transfer in Pipes with Volume Heat Scurces Within the Fluids, ORNL-1395 (Dec. 2, 1952). PERIOD ENDING JUNE 10, 1954 tween the fuel and NaK are instructive. If the temperature difference for the zirconium-base fuel is to be the same as that for the lithium-base fuel, the tube length must be increased, for example, from 6.0 to 7.6 ft if the temperature difference is kept ot 91°F. This would require a 15% increase in the number of tubes. The Reynolds number would drop from 4850 to 2100, that is, from a value well within the turbulent range to one likely to give laminar flow and hence even poorer heat transfer than that estimated. [f the same tube spacing were used and a higher pressure drop accepted, the comparison would be less unfavor- able in most respects. However, the pumping horsepower would need to be increased by 50% with the zirconium-base fuel, and many stresses, such as those in the NaK outlet header sheets, would be more than doubled.. The effects of heat-exchanger volume on beth shield weight and the activation of the NaK in the secondary circuit are vitally important. -The weight of the reactor, heat exchanger, and reactor shield assembly, together with the radiation dose from the NaK in the secondary circuit at a point 50 ft from the engines, is given in Fig. 3.2 es a function of the temperature loss in the fuel-to-NaK heat exchanger. The data are for a 200-Mw reactor with a shield designed to give 10 r/hr ot 50 ft from the reactor at full power. It can be seen that for a given design temperaiure loss the lithium-base fuel is markedly superior to the zir- conium-base fuel in that it gives one-half the radiation dose and a 6000-lb saving in shield assembly weight. The limitations on high-temperafure heat-ex- changer design imposed by pressure stresses merit further explanation. The many closely spaced holes in the header sheets generolly make the header sheets much less strong than the thin tube walls. In the spherical-shell heat exchanger for the Reflector-Moderated Reactor, the NaK must enter the heat exchanger at about 100 psi to take care of the roughly 50-psi pressure loss in the heat exchanger, the 40-psi loss in the radidators, and the 10-psi loss in the connecting lines. Most of the estimated 50-psi pressure drop in the fuel system occurs in the heat exchanger. On this basis the pressure difference across the header sheet at the hot end is less than 10 psi, while that at the cold end is about 100 psi. Fortunately, the strength of Inconel is about five times greater 29 ANP QUARTERLY PROGRESS REPORT ORNL-LR-DWG 1603 4000 wn el m , : ‘ e S : L e S EE - — 3500 + : W O | | 2 350 |——-r{- - - ‘ - : 3000 g — 4 T 2 300 L QoW b o o < 2 250 = @ = © 2 200 14 £ g u w5 5 Z X 12 2 [n et Ld 10 [T = <1 T Ly -~ — o o T —~ {20 la a Ll O {10 = 1 i Ll & 100 o w « 90 = }._ <{ & a. 80 = wh ]—. ¥ 70 Qo < o K 60 ol w z 50 5000 @ m > 4000 32 w 0 e PETTP ‘ e a —_— . he—— 3000 2 Zr-BA = ‘ SE FUEL TUBE SPAGING = 0.020in E_. ————— S -Zr-BASE Fug( = ot — 2000 - 0.027i . i l ! 'n- 2 | ‘ i 1000 6 7 8 g 10 TUBE LENGTH {ft) Fig. 3.1. Effects of Tube Length on Design Variables for a 60-Mw Fuel-to-NaK Heat Exchanger for 50.psi Pressure Drop in the NaK Circuit. 30 REAGTOR-HEAT EXCHANGER-SHIELD ASSEMBLY WEIGHT {tb} Fig. 3.2. Weight of Reactor, Heat Exchanger, and Reactor the NaK in the Secondary Circuit at a Point 50 ft from the Loss in the Fuel-to-NaK Heat Exchanger. 88,000 85,000 80,000 75,000 70,000 20 ORNL-LR-DWG 1604 FUEL-TO-NaK TEMPERATURE DIFFERENCE (°F) 0.80 0.60 0.20 PERIOD ENDING JUNE 10, 1954 RADIATION DOSE FROM NaK AT 50ft FROM ENGINES (r/hr) -Shield Assembly and Radiation Dose from Engines as a Function of the Temperature 31 ANP QUARTERLY PROGRESS REPORT at the cold end than at the hot end. It has been shown? that the optimum tube diameter is close to % . in. and that the minimum thickness of liga- ment between holes in the header sheet should be at feast 0.10 in. for satisfactory welding. |f an essentially cylindrical header sheet, such as that shown in Fig. 3.3, is used to minimize the stresses, the ligagents between the tubes will be in tension. |f allowances are made for both the cross-sectional area lost in the holes and the increases in the stresses around the holes that arise from geometric effects, the maximum tensile stress in the ligaments will be about five times that for a simple cylindrical shell of the some wall thickness with ne perforations. The 1000-hr rup- ture strength for Inconel at 1200°F is 12,000 psi, and therefore the allowable stress should not exceed 6000 psi to give some factor of safety. |f the header-sheet thickness were arbitrarily chosen as 0.25 in., its rodius of curvature would be 3 in. (The allowable stress from simple loop tension considerations is 60 times the pressure differential for the unperforated cylinder, or 12 times that for the perforated cylinder. Thus the ratio of header- sheet thickness to radius of curvature becomes 1:12.) While other plausible designs can be evolved on a similar basis, that illustrated in Fig. 3.3 appears to be about the best devised to date. 3A. P. Froas and M. E. LaVerne, Heat Exchanger Design Charts, ORNL-1330 (Dec. 7, 1952). UNCLASSIFIED ORNL-LR-0OWG 1605 3e-in,-CD TUBES, 0.020~in.-THICK WAL / / —~ NaK INLET TUBE e — T F K*‘* HEADER SHEET Fig. 3.3. Section Through Header Region of o Tube Bundle for a 60-Mw Heat Exchanger. 32 REACTOR FPHYSICS W. K. Ergen ANP Division Estimates of the xenon poisoning? of the 60-Mw Reflector-Moderated Reactor (power density 1.35 kw/cc) indicate that unless the xenon is rather rapidly purged from the fue!l during operation a reactivity loss of the order of a few per cent will occur. Also, the maximum xenon concentration after shutdown would give a reactivity loss of about 10% or more. A control mechanism capable of compensating for such reactivity losses would be rather complicated, and the added uranium con- centration would be highly undesirable. Therefore, an experiment is now being planned to determine whether the xenon can be adequately purged. The elaborate coding of the multigroup calcu- lations, mentioned in the previous report,” has temporarily been set aside. Instead, o three- group three-region code is being prepared which does not exceed the present capabilities of the ORACLE. This code is adequate for a variety of problems, and the use of the ORACLE is far more convenient than the use of on out-of-town machine. The possibility of the coupling of mechanical oscillations with nuclear-power oscillations in the Reflector-Moderated Reoctor was studied, since such coupling is known® to be capable, in prin- ciple, of creating antidamped oscillations. In particular, the possibility of the reactor oscillating mechanically like a Helmholtz resonator was in- vestigated. The antidamped oscillations cannot occur if the frequency of the mechanical oscil- lations is very much higher than the frequency of the nuclear-power oscillations, as will be the case in the 60-Mw CFRE. REACTOR CALCULATIONS M. E. LaVerne ANP Division C. S. Burinette United States Air Force Core radii of 20, 30, 40, and 60 cm, fuel thick- nesses of 5, 10, 15, and 20 cm, and extrapolated 4J. L. Meem, The Xenon Froblem in the ART, ORNL CF-54-5-1 {May 3, 1954). SR. R. Coveyou and R. B. Bate, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 40. 6P. R. Kasten, Linearized Stobility Criterio for HRT Type Reactors, ORNL CF-54-4-183 (Apr. 21, 1954). reflector thicknesses of 20, 30, and 40 cm were selected for calculations of the set of related reactors referred to previously.” Poison (sodium and Inconel) distributions in the reflector and island were obtained from a previous study.? The ’M. E. LaVerne and C. S, Burtnette, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 39, : 8R. W. Bussard et al., Moderator Cooling System for the Reflector:Moderated Reactor, ORNL-1517 (Jun, 22, 1954). REFLECTOR THICKNESS = 20 CM A o> S o £ = <" 7 / UEIT CONCENTRATION (MOLE %) o /- £ %/ rne o UEE pONCENTRATION (MOL PERIOD ENDING JUNE 10, 1954 fluoride fuel NaF~ZrF4-UF4_ (53.5-40.0-6.5 mole %) was used for the entire set of calculations. From the set of 48 reactors determined by the specified independent parameters, it was found possible to select 30 reactors that would give results that could be used to cross-plot and interpolate the properties of the 18 omitted reactors to an ac- curacy sufficient for this survey. The results obtained are summarized in Figs. 3.4, 3.5, 3.6, and 3.7. — ORNL-LR~DWG 1£77 7 / REFLECTOR THICKNESS =40 CM Fig. 3.4. Effect of Reactor Dimensions on Concentration of U235 in Fuel, 33 ANP QUARTERLY PROGRESS REPORT In view of stringent limitations on permissible uranium concentrations in the fuel, it is clear from Fig. 3.4 that there exists a considerable advantage in increasing reflector thickness from 20 to 30 em but exceedingly little advantage in a further increase to 40 em. The remaining figures have, therefore, been restricted to a 30-cm re- flector thickness. EXTRAPOLATED REFLECTOR THICKNESS =30 CM TOTAL UESS INVESTMENT LB) EXTERNAL FUEL VOLUME = 0O FT? S INVESTMENT (LB) I > 7O074L L€ EXTERNAL FUEL VOLUME = 4 FT?> The effect on total uranium investment of the reactor dimensions and the external fuel volume, such as would exist in a heat exchanger external to the core, is shown in Fig. 3.5, The surface shown for zero external volume is, of course, simply that for critical mass., At a core radius of 30 c¢m the critical mass is virtually independent of fuel thickness. In the surfaces for both the ORNL-LR~-DWG 1176 EXTERNAL FUFL VOLUME = 8 FT?2 Fig. 3.5. Effect of Reactor Dimensions and External Fuel Yolume on Total U233 Investment. 34 ORNL-LR-DwG 156 EXTRAPOLATED REFLECTOR THICKNESS = 30 tm ’ FEAE -TO-AVERAGE POWER DENSITY RATID AT - /‘?%} P = l L . g PO - i) e o E ., 1 ) - o Lo % - <5 "o \\\\ - - 11/// Q_t;";D ~ L & e Mo, 2| R <.‘Lg‘=9 ~ L/ ] C‘,,p,/ Sy Fig. 3.6. Effect of Reactor Dimensions on Out. side Peak-to-Average Power.Density Ratio in Core of Reflector-Moderated Reactor. ORNL*&! - !WG 1197 EXTRAPDLATED REFLECTOR THICKNESS = 30 oM 0 8g g & -~ =l =1 4 %3 By 3 & & - 0 Lk /%{6 Fig. 3.7. Effect of Reactor Dimensions on Per Cent Thermal Fissions in Core of Reflector- Moderated Reactor, Z- and 4-cu ft external volumes, simple visual examination indicates the probable existence of an optimum set of proportions. Whether any closed contours {as required for an absolute minimum) actually exist has not yet been determined. The reversal of scales for the 8-cu ft surface was necessitated by the existence of an extreme peak at what was the front corner. PERIOD ENDING JUNE 10, 1954 The effect of reactor dimensions on outside peak-to-average power-density ratio is shown in Fig. 3.6. The effects here are not large; an in- crease in core radius from 20 to 60 cm at a con- stant fuel thickness gives a decrease in the ratio of about 20%, while an increase in the fuel thick- ness from 5 to 20 cm at o constant core radius results in an increase in the ratio of about 33%. The per. cent thermal fissions as a function of reactor dimensions is shown in Fig. 3.7. The least-thermal reactor (28%) has the thickest fuel layer and the smallest core, while the most-thermal reactor (45%) has the thinnest fuel layer and the largest core. When the 60-Mw design reactor’ was modified by deleting the Inconel tubes and their surrounding stagnant sodium and fitting the resulting voids with beryllium, the critical mass decreased by 29%, that is, from an original value of 40.7 b to 29.0 Ib. This percentage decrease was duplicated by NDA? in a calculation on a 300-Mw reflector- moderated reactor for which the same method of poison reduction was employed. The Curtiss-Wright Corporation, under contract to USAF, is performing an extensive series of multigroup, multiregion calculations on the IBM- 701 for the ORNL-ANP Project. Specifications for these calculations are being prepared by the ORNL-ANP General Design and Reactor Physics Groups in cooperation with Curtiss-Wright. Critical mass limits computed by Curtiss-Wright for the two-region, Teflon-uranium foil core, be- ryllium-reflected critical experiment as affected by foil thickness and space-point interval are listed in Table 3.2. These results agree quite well with the UNIVAC calculation previously re- ported.’ qM. R. Adolph, R. C. Ross, and M. S. Silberstein, Circulating-Fuel Reactor Studies, Part [, NDA 10-122 (Apr. 4, 1954). TABLE 3.2. CRITICAL MASS LIMITS FOR RHOMBICUBOCTAHEDRON CRITICAL EXPERIMENTS FOIL THICKNESS Ar CRITICAL MASS (mils) (em) (Ib of U233 o+ 1.828 14 10 0.457 18 *Homogeneous mixture. 35 Part i MATERIALS RESEARCH 4. CHEMISTRY OF MOLTEN MATERIALS W. R. Grimes Materials Chemistry Division The experimental studies of the chemistry of molten materials have been. devoted almost ex- clusively to fluoride systems of interest as fuels. Considerable attention has been paid, during the quarter, to systems in which the vraniferous ma- terial consists wholly or in part of UF .. Studies of phase equilibrium by the method of thermal analysis have been continued. This tech- niqgue has been applied to exploratory studies in mixed chloride-fluoride systems, Quenching tech- niques have been significantly improved and have been used, with petrographic and x-ray examination of the melts so produced, to virtually finish the elucidation of the complex NaF-ZrF, system. Differential thermal analysis and filtration tech- niques have been used in several phase-equilibrium experiments, and an apparatus for visual obser- vation of liquidus temperatures has been tested and found to be satisfactory. ' The solubility of UF, in several molten-sait systems has been examined experimentally. |t appears that attainable concentrations of UF, in Z¢F ,-based fuels are insufficient to fuel aircraft reactors; however, use of UF, and UF, in such mixtures appears to be promising. The UF, com- pound is apparently sufficiently soluble in NaF- LiF-KF mixtures (which would require Li’7 for vtilization) to make such fuels atiractive; the use of NaF-RbF-LiF mixtures as solvents for UF, {without UF4) has not yet been shown to be pos- sible, The large-scale (250-lb-capacity} equipment for the production of fluoride mixtures has been re- activated and is now in three-shift five-day oper- ation for furnishing the ZrF -based materials re- quired for large-scale testing at ORNL and other installations, ' QUENCHING EXPERIMENTS WITH FLUORIDE SYSTEMS R. E. Thoma ‘R. E. Moore M. S, Grim C. J. Barton Materials Chemistry Division G. D. White H. Insley, Consultant Metallurgy Division The major part of the quenching work with com- positions in the NaF-ZrF, system has been com- pleted and a revised diagram for the system is presented. Work on the Na F-UF , system continued, but further work will be required before a revised diagram can be completed. It is anticipated that attention will shift from the binary system to the NaF-ZtF UF , system and possibly the NaF-ZrF - UF, system in the near future, NaF-Z:F, The methods and apparatus for performing quench- ing experiments were described previously. ' During this quarter, all the guenching experiments were made by heating the sample up to, but not above, the quenching temperature, , The incongruent melting point of the compound Na,Zr F ., was established as 538°C by quenching mixtures with 57 and 60 mole % ZrF ;. The liquidus temperatures of mixtures with 57 and 60 mole % ZrF, were found to be 546 + 5°C and 609 t 5°, respectively. The 57 mole % ZrF, composition (almost 100% NayZr,F,,) was converted to ZrF, and liquid at 538°C. From the shape of the liquidus curve and from x-ray diffraction studies of large samples of slowly cooled mixtures, it appears that the ZrF, primary-phase field extends to about 56 mole % erf:‘l in the NaF-Zer system. Quenching and glass-devitrification experiments indicate that the liquidus-solidus temperature of the 47 mole % ZrF, composition is about 522°C, and the liquidus-solidus temperature of the 50 mole % ZrFA composition is about 511°C. Recently, very careful thermal-analysis comparison of the 47 and 50 mole % ZrF4 mixtures showed that the 47 mole % mixture melts and freezes sharply at 520°C and the 50 mole % mixture freezes sharply at 510°C. These results mean that the single phase which is always obtained on cooling 47 mole % ZrF4 compositions is a congruently melting com- pound (NagZr F,,), while the 50 mole % ZrF, composition is near a euftectic of Na Zr,F . and Na,Zr F . A phase referred to as R-3 has appeared? in association with glass in compositions ranging 1C. J. Barton et al., ANP Quar. Prog. Rep. Dec. 10, 1953, ORNL-1649, p 54. 2R, E. Thoma et al., ANP Quar. Prog. Rep. Mar, 10, 1954, ORNL-1692, p 52-54. 39 ANP QUARTERLY PROGRESS REPORT from 45 to 57 mole % ZrF,. This phase forms only when large quench samples are used, that is, under poor quenching conditions. Experiments with very small samples at various temperatures have failed to produce R-3. It was concluded from these ex- periments that R-3 does not have a stable existence in the binary system but is an easily crystallized metastable compound (probably NaZrF;) that is produced by fairly rapid cooling below the equi- librium liquidus temperature, Petrographic and x-ray diffraction examination of slowly cooled preparations led to the postulation of the existence of o weakly birefringent compound Na,Zr,F,, and a series of solid solutions between it and Na,ZrgF .. Petrographic examinations of samples of 43 and 45 mole % ZrF , compositions heated to 483°C and held for 20 hr, before quench- ing, definitely pointed to an immiscibility gap in this series between about 42 and 44 mole % ZrF . Petrographic examinations of a large number of quenched samples with between 39 and 45 mole % ZrF, led to the conclusion that Na,Zr F | is not stable at the liquidus temperature. In 47 to 42 mole % ZrF, compositions, the solid solution NagZrgF ., is the primary phase. The liquidus temperatures . of compositions with 47, 45, 43, and 42 mole % ZrF, are about 522, 520, 517, and S08°C, respectively. In 39 and 40 mole % ZrF compositions, the primary phase is Na,ZrF . The liquidus temperature for the 39 mole % Zri, com- position is above 513°C, ond for the 40 mole % ZrF, composition it is near 510°C, For the 41 mole % composition the liquidus temperature is about 502°C, and this composition is near a eutectic between Na,ZrF, and the NayZr F,, solid so- lution. Below about 488°C, Na,Zr,F,, appeors as fine-grained fibers. Slowly cooled compositions with about 40 mole % ZrF, usually undergo a reaction just after complete solidification that produces a fluff of very fine-grained material, According to x-ray diffraction results, this material is a mixture of Na,Zr,F,, and NaZr F .. It is likely that the reocfron is the solldmsfofe conversion at about 488°C of the eutectic into these com- pounds, The formation of Na,Zr,F,, in the solid state explains why it has olways Leen found as fine grains, A number of different forms of Na,ZrF have been observed, and quenching experiments have been conducted to establish the transition tempera- tures., This work is not yet complete, but the approximate temperatures are known, The phases which have been found and the optical properties and the approximate temperatures at which these phases are stable are given in Table 4.1. The temperature of transition from phase 5 to phase 4 TABLE 4.1, CRYSTAL MODIFICATIONS OF Na2ZrF6 APPROXIMATE TEMPERATURE PHASE OPTICAL PROPERTIES °C) -1 | sotropic 545 to 620 2 Uniaxially positive, O = 1.406 540 1o 545 E = 1.408 3 Uniaxially positive, O = 1.376 505 to 540 E = 1.386 4 Biaxially positive, y = 1.412 Below 460 a = 1.408 5 Biaxially negative, y = 1.419 a = 1.412 6 Biaxiolly negative, y = 1.429 a = 1.420 40 is not known, but it is below 460°C. Phases 3 and 4 have been found in quenches of mixtures with 37 ond 40 mole % ZrF, at the ap- propriate temperatures. The refractive indices were the same as for the 33 mole % ZrF, compo- sition, and thus there appears to be ||f1e or no solid-solution range for these phases, The phase diagram presented in Fig. 4.1 repre- sents the data discussed here and in previous reports. 13 Thermal data obtained from cooling curves, together with visual observation of the Na,ZrF, compound, were the chief sources of IIqUIdUS temperatures in the 0 to 30 mole % ZrF region. The Na,.ZrF_, compound was observed to melt rather sharply between 840 and 850°C to a clear liquid that permitted observation of turbidity on cooling to a temperature slightly above 850°C, The liquidus temperatures in the 60 to 100 mole % ZrF , region were obtained from thermal data, NaF-UF, Eight samples covering the range in the NaF-UF , system from 18 to 67 mole % UF4, chosen to fill insofar as possible the gaps in the knowledge of the system,? were mixed and hydrofluorinated to minimize oxide contamination. These slowly cooled specimens were examined by petrographic and x-ray techniques. The results of this examination, some of which were published previously, are shown in Table 4.2, BR. E. Moore, C. J, Barton, and T. N. McVay, ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL-1609, p 59. ORNL - LR -DWG 204 1000 s W i par = o x W50 = I} *_ NoZiFe (MZTASTABLE) 400 [ : bl 'duEZrFS - Naz o fyy ‘Nag ZrgFay o 2 20 30 40 50 60 70 80 S0 4 100 Zriy (mate %) Fig. 4.1. Phase Diagram of the NaF.Z:F, System, PERIOD ENDING JUNE 10, 1954 TABLE 4.2, PHASES PRESENT IN SLOWLY COOLED NaF-UF , MIXTURES PHASES IDENTIFIED BY PETROGRAPHIC AND X-RAY DIFFRACTION AMALYSES COMPOSITION {mole % UFA) 18 B Na,UF o, NajUF,, NaF 22 ‘8 NuZUF NaF 25 '83'N°2UF6" N03UF7, NaoF (trace) 28 [‘3 Na,, UF (major}, Na,UF 30 f)) NaZUFfi (major), Na UF 31.5 [‘3 Na,UF 33.3 B NcQUF 35 B -NQZUFfi (major), NaUF ¢ 37 [) a,UF o (major), NaUF ¢ 38.5 fi NuzUF {major), NnUF5 40 3 --Nm2 g NaUFg 43 [33-N02UF6, NaUF 46 fi?’-quUFé, NuUF:5 (major) 48 NaUF ., UF, {trace) 50 NaUF (major}, fia-Nc:ZUFé, UF . U0, (trace) 53 NaUF (major), UF 67 NaUF UF 5! Small portions of these specimens have been used with conventional techniques in quenching experiments, From 18 to 30 quenches have been made on portions of all specimens except the Na,UF, composition, for which a total of about 80 attempts has been made. By quenching speci- mens from temperatures as high as ]022“(:; to as fow as 399°C, it has been established that four crystalline forms of the Na ,UF, composition exist, These crystalline modifications are a (cubic), B, (hexagonal), B, (hexagonal), and y (orthorhombic}. In no case could the Na,UF, be quenched com- pletely to glass; the a and y forms occurred in quenches from 637°C or higher. The y form appears to be most prevalent in the more rapid quenches. The B, form has been observed as the major phase in large, slowly cooled melts and in most quenches from temperatures below 637°C. The 3, form has been prepared in a practically pure state in the 41 ANP QUARTERLY PROGRESS REPORT temperature range of 612 to 589°C by using samples somewhat larger than those customarily used in quenching experiments. In general, the quenching data indicate that NalUF is the most uranium-rich compound in the NaF-UF , system. It is anticipated that a definitive diagram of this binary system will be prepared during the next quarter, VISUAL OBSERVATION OF MELTING TEMPERATURES R. J. Sheil C. J. Barton Materials Chemistry Division An apparatus has been constructed that permits visual observation of fused fluoride and chloride mixtures at high temperatures under an inert atmos- phere, It consists of a 2‘/2-in.-OD metal pipe threaded at the ends and mounted vertically. Glass disks gasketed with Teflon are fitted to the tube ends, Large nuts are used to compress the gaskets which form virtually gosstight connections. Water- cooled copper coils are used to cool the furnace ends and to protect the Teflon gaskets. The inert gas, usually argon, is maintained at slightly above atmospheric pressure; a slow flow of gas enters near the bottom of the apparatus and leaves near the top. Samples are placed in 20-ml nickel crucibles supported on a tripod near the center of the furnace. A bare chromel-alumel thermocouple junction is immersed in the fluoride sample. The thermocouple was checked at the sodium chloride melting point after several exposures to molten fluorides; it was found that the slight corrosion had not affected the emf output. Visual observations have been confined thus far to a few compositions in the NaF-ZrF, and the NaF-UF, systems. However, the usefulness of the equipment for confirmation of liquidus and solidus temperatures has been demonstrated, and its use in phase studies will undoubtedly increase. DIFFERENTIAL THERMAL ANALYSIS OF THE NaoF-ZrF, SYSTEM R. A. Bolomey Materials Chemistry Division A critical study of the differential thermal-analy- sis data collected for the NaF-ZrF that they are consistent with petrographic and x- ray analysis data obtained from quenching and system shows 42 devitrification experiments, A comprehensive re- port on the technique for differential thermal analy- sis is being prepared, The report will describe the theory of differential thermal analysis in a quali- tative manner, the preparation of the samples, the experimental technique, the interpretation of the data, and the precision and accuracy of the method. Suggestions for further improvements of the method will be included, FILTRATION ANALYSIS OF FLUORIDE SYSTEMS R. J. Sheil C. J. Barton R. E. Thoma Materials Chemistry Division G. D. White Metallurgy Division The filtration apporatus was diverted from phase studies to solubility determinations during this quarter because of the urgent need for data on the solubility of UF; in alkali fluorides and in NaF- ZrF, mixtures. A few experiments were made, however, to demonstrate the solubility of UF, in the NaF-RbF-LiF eutectic and to define the pri- mary phase in some NaF-ZrF4 mixtures. NaF-ZrF“ One filtration was carried out in the NaF.ZrF, system to supplement information obtained by other techniques, A mixture containing 63 mole % ZrF was filtered at 552°C, and 79.5 wt % of the material filtered. The filtrate contained 55 mole % ZrF, and was predominantly Na,Zr,F . with a small amount of NayZr,F,,. The residue analyzed 78.5 mole % ZrF, and was predominantly ZrF , with some Nc:32r4F1 and NagZr F .. These results, which demonstrate that ZrF | is the primary phase in the 63 mole % er:4 composition, lend support to the revised diagram for the NaF-ZrF, system (Fig. 4.1). NaF-LiF-RbF-UF The solubility of UF, in the NaF-LiF-RbF eu- tectic at 500°C was determined by filtering a mixture of the following calculated composition (in mole %) 5.8 NaF-40.3 LiF-49.9 RbF-4.0 UF,. A cooling curve obtained prior to the filtration showed o questionable thermal effect ot 542°C and a definite break at 428°C, that is, slightly below the melting point of the ternary alkali fluoride eu- tectic (435°C). About 90% of the charge material possed through the filter at 500°C. All the uranium was apparently in the form of Rb JUF, complex. Petrographic examination showed fhat the filtrote contained about 10 to 20 wt % of this uranium compound and that the residue contained approxi- mately 50 wt %, Chemical analyses of the sepa- rated phases gave the following compositions (in mole %): In Filtrate Ih Residue LifF 48.5 . 30.3 NaF 6.9 9.4 RbF 41.7 50.2 UFA 2.8 (10.2 wt %) 10.1 The chemical analyses appear to confirm the petrographic finding that a RbF.-UF, complex is the primary phase that separates from the NaF- LiF-RbF-UF, mixture. This result was not un- expected becuuse phase studies of the alkali fluoride—UF, systems have shown that Rb,UF, has the h|ghesf melting point and is the most stable of the UF, complexes that can be formed by the components present in the NaF-LiF- RbF- UF, system. THERMAL ANALYSIS OF FLUORIDE SYSTEMS L. M. Bratcher A. B, Wilkerson Materials Chemistry Division G. D. White T. N. McVYay, Consultant Metallurgy Division - Preliminary data on the NaF-ZrF ,-UF, and NaF- ZrF sUF (-UF, systems disclose that conventional thermal-analysis techniques do not give a reliable indication of liquidus temperatures in these po- tentially useful systems. The apparatus developed for thermal analysis of chloride systems, which consists of a flanged stainless steel container which holds the samples in nickel crucibles, has been found useful for thermal onalysis of these UF ;-bearing systems. An investigation of phase behqvmr in the NaF-CrF, system has been initiated because of the importance of CrF. in corrosion of Inconel and stainless steels by molten fluorides. Work on the NaF-ThF system has been concluded, although the results obtained are not completely satisfactory; should this system become important other techniques will be requ;red for establishing the phase behavior, - PERIOD ENDING JUNE 10, 1954 LiF-UF, Efforts to obtain data on mixtures in the LiF-UF, system were made rather early in the phase-study program, but the results were unsatisfactory be- cause of the presence of UF, and UQ, in the fused mixtures. Relatively pure LiF-UF, mixtures can be produced by adding 100% excess of uranium metal to a previously fused LiF-UF, mixture, The few data obtained for the relatively pure mixtures indicate that there is a eutectic containing about 25 mole % of UF, 770°C., that melts at approximately F’efrogruphlc examination of the fused mixtures showed only LiF and UF,. It is though’r that the LiF-UF, mixtures probably compr:se a simple eutectic sysfern NaF-UF, Data previously reported?+® indicated the exist- ence of one eutectic and one compound in the NaF-UF, system, When a mixture containing 33 mole % UF, and 67 mole % NaF was prepared by fusing NaF with previously prepared UF,, thermal effects were noted at 750, 695, 650, and 595°C, These data agree quite well with the earlier data, but the petrographic examination of the mixture showed the presence of a green, presumably oxi- dized, phase in addition to the lavender NaF-UF, complex. Further study of this system is planned. RbF-UF, Very few compositions prepared in the RbF-UF, system were free of UD, and UF, complexes, The only compound which has been observed (either Rb,UF, or RbyUF ) is described as a brown- orange |so’rmp|c phcse with a refractive index of 1.44, The 33.3 mole % UF, composition showed a thermal effect at 860°C. The method of preparing alkali fluoride~UF ; mixtures by adding an excess of uranium metal to UF, mixtures is not suitable for the RbF-UF, system because the reduction of RbF by uromum gives volatile metallic rubldlum. KF-LiF-UF, A KF-LiF-UF, (48-48-4 mole %) mixture was prepared by reducing a previously fused KF-LiF- UF mixture with 100% excess uranium metal. Thermal effects were noted on the coeling curve 4\’. S. Colemun and W C. Whitley, ANP Quar. Prog. Rep. Sept. 10, 1952, QRNL-1375, p 79. 5W C. Whitley, V. S. Coleman, and C, J. Barton, ANP Qudr. Prog. Rep. Dec. 10, 1952, ‘ORNL- 1439, p 109. 43 ANP QUARTERLY PROGRESS REPORT at 558, 540, and 487°C. It is probable, although not yet established, that the thermal effect at 558°C represents the true liquidus temperature for this composition. The fused melt contained about 15% of an orange, low-birefringent phase of re- fractive index 1.43 and colorless isetropic ma- terials with the refractive indices of KF and LiF, These two alkali fluorides apparently do not form a complex comparable to the LiF-RbF complex. RbF-LiF-UF, A Rbl'"_-!..if:~Ui'"“3 (56-40-4 mole %) mixture was prepared by mixing uranium metal and previously prepared UF, with the fused alkali fluorides. No thermal effect was observed above 475°C, the melting point of the RbF.LiF eutectic. Since the filtration experiments reported above indicate that the liquidus temperature for a similar composition is above 550°C, it is concluded that the first separation of a UF j-containing phase from ternary mixtures in this region cannot be readily detected by conventional thermal-analysis techniques. Pet- rographic examination revealed that the melt con- tained a colorless birefringent phase, presumably a LiF-RbF complex, and a brownish isotropic phase that was probably a RbF-UF, complex. NaF.ZrF UF,-UF, A series of samples with various amounts of the UF, was prepared in sealed nickel capsules in mixture that contained initially NaF-ZrF -UF, (53.5-40-6.5 mole %) with sodium metal added to reduce the UF; to UF,. When the capsules had been heated to maximum temperatures of about P00°C they were inverted so that the contents would mix, Duplicate cooling curves were ob- toined with each composition by reheating and remixing the contents. The data obtained are presented in Table 4.3. In the data, the dilution of the melt by the NaF produced in the reaction is disregarded, The dota seem to indicate a minimum melting point when a little more than one-half the uranium is reduced. Petrographic examination of the fused melts showed a regular grodation of properties in the series. Two phases were observed in the un- reduced mixture, the Naq(Zr,U) F“ 8 solid solution ond a small amount of colorless Nog.ZrF o LrF . Four phases were observed in the reduced mixtures: UF,, o yellow isotropic material, a yellow-green solid solution, ond Nu2ZrF6. With increasing degrees of reduction, the solid solution became increasingly yellow, oand the isotropic yellow phase, fres UF,, ond yellow-green Na, ZrF ap- peared in increosing amounts, This crystalline form of Na,ZrF ., usually referred to as “normal” Na,ZrF ,, does not, according to present knowledge, dissolve tetravalent urgnium, NaF-ThF, Thermal analysis of the NaF-ThF, system has given data which are difficult to interpret. Petro- graphic examination of the fused melts has failed to clarify the situation; Na,ThF . was the only crystalline phase thot could be definitely identified. The further study that would be needed for clarifi- cation does not appear to be justified at this time, and therefore the available dota are presented in the form of a tentative diagram (Fig. 4.2)., Petro- graphic examination of some melts made recently indicated that the 50-50 composition is close to a pure compound and suggested that another crystal- TABLE 4.3. THERMAL EFFECTS OBRSERVED WiTH NaF-ZrF“-UF(UFa COMPOSITIONS COMPOSITION THERMAL EFFECTS (°C) (mole % UF,, calculated) First Cooling Curve Second Cooling Corve 0 565, 530 (halt), 490 557, 528, 492 1.5 555, 535, 510 576,525 3.0 510, 500 507, 500 (halt) 4.5 507, 495 509, 495 5.5 540, 519 (halt) 550, 520 525, 490 44 6.5 525, 487 b A0 DRN LR -DWG 1531 7 1000 b Cle s ‘R } e 00 | - |\ e wol e N e s LT _,il\ffifi{%fi . 20 20 A0 80 &0 7 82 21 Thr, Thi, [ mole %) TEMPERATURE (°C) Fig. 4.2. Tentotive Phase Diagram of the NaF- ThF, System, line phase observed in the 25 and 30 mole % ThF mixtures might be Na,ThFg. Zachariasen® re- ported the existence of this compound. These examinations also showed that more oxide was present in compositions formed by adding ThF, to the 50-50 mixture than in those formed by adding NaF. Therefore it seems probable that oxide and/ or water in the ThF is the major source of oxygen in the fused NaF-ThF , composition. Lif-BeF,-ThF -UF, Preliminary examination has been made of the LiF-BeF -ThF ,«UF , system which has been con- sidered as a p055|b|e fuel for a breeder-power reactor,’ Cooling curves obtained with two mix- tures of LiF- BeF2-T|1F UF, (48.6-48. 6-2.7-0.1 mole %) showed thermal effects at 515, 392, and 325°C., The thermal effect at 515°C was slight but reproducible and probably represents the first separation of a ThF -containing phase. This sepa- ration should be confirmed by other techniques, such as filtration or quenching, if this composition continues to be of interest. ‘It is possible that a composition with a higher LiF and a lower BeF, conterit might have a lower liquidus temperature, although presumably such a mixture would be less efficient as a moderator. Sw. H. Zachariosen, J. Am. Chem. Soc. 70, 2147 (1948). S, Peterson, Reacfor Chemistry, ORNL CF-54-4-198 {May 12, 1954). PERIOD ENDING JUNE 10, 1954 Na F-CrFa Mixtures containing 5 to 65% Crf_ in the NaF- CrF, system have been exomined, The thermal dum do not serve to define the system uniquely. Petrographic examination indicates the existence of a compound which seems to be NaCrF,. Two other crystalline materials have been observed but neither has the optical properties of the Na,CrF, compound previously described. & Thermal effects are observed below the solidus temperature ot 25 mole % CrF,; it is possible that several crystal- line modtfucahons of Na,CrF . exist. THERMAL ANALYSIS OF CHLORIDE AND MIXED CHLORIDE-FLUORIDE SYSTEMS A. B, Wilkerson C. J. Barton Materials Chemistry Division T. N. McVay, Consultant Metallurgy Division Studies of chloride systems have been continued at a somewhat diminished rate during the paost quarter, The systems studied have included a few mixed chloride-fluoride compositions and some binary and ternary systems containing UCI . UcCl,-Uci, Conflicting statements about the UCI,-UCI, sys- tem are found in the literature. It was stm‘ed by Kraus? that the two compounds are only slightly soluble in each other in the liquid state, and he reported a liquidus break for only one composition. The group ot Ames, lowa, !0 stated that there is a eutectic in the system at 88 mole % UCI , with a melting point 14°C lower than that of the UCI . Five compositions in this system were prepared in sealed capsules, and thermal effects were noted on the cooling curves obtained after the mixtures had been heated to about 850°C and the capsules had been inverted to obtain mixing of the contents, Samples of UCI and UCI were also heated in sealed capsules. In uddn‘lon, one UC! -UCI, (80- 20 mole %) mixture was prepared in a mckel con- tainer at atmospheric pressure (helium), but it was not heated to such a high temperature as were the 8[... M. Bratcher and C. J. Barton, ANP Quar. Rep Dec. 10, 1952, ORNL-1439, p 116. %c. A. Kraus, Phase Diagrams of Some Complex 3alts of Uranium with Halides of the Afkali and Alkaline Earth Metals, M-251 (July 1, 1943). mJ. J. Katz and E. Rabinewitch, The Chemistry of Uranium, p 488, McGraw-Hill, New York, 1951, Prog. 45 ANP QUARTERLY PROGRESS REPORT sealed capsules because of the high vapor pres- sure of UC|4. The data are given in Table 4.4. The thermal data give little indication of a eu- tectic but are not easily interpreted because of the anomalous thermal effects noted with the twe components in the sealed capsules. In open con- tainers under an inert atmosphere, only one break was noted on the UCl, cooling curve, at 840°C, and one with UCI,, at 565°C. Both UCl; and UCt, were noted petrogrophically in the fused ucl, probably because of reduction of UCI, by the con- tainer, but this does not seem to account for all the thermal effects noted, Both UCl, and ucl, were found in all the mixtures, except for the 12% UCl, compesition, in which no UCI, was visible under the microscope. It is possible that oxidation of the UCI; may have occurred before the petro- graphic examination could be completed, UCI,-UF, Only one composition was prepared in the uct - UF, system, a 50-50 mixture corresponding to the compound UCL,F, reported in the literature. 1 It gave a single thermal effect at 520°C with an indi- ”J. W. Gates, Jr., G, H. Clewett, and H. A. Young, Preparation of a New Mixed Halide of Tuballoy, CD-454 (duly 20, 1944); J. W. Gates, Jr., et al., A Mixed Tub- alloy Tetrahalide, CD-460 (Aug. 26, 1944), TABLE 4.4, THERMAL EFFECTS ON COOLING CURVYES FOR MIXTURES OF UCI3 AND U(.‘.I4 THEORETICAL COMPOSITION THERMAL EFFECTS (mole % UCI3) Cc) 100 835, 768, 520 80 797, 535, 488 60 798,(9) 540,(%) 405 40 762, 542, 497 20 622,%) 545, 531, 497 20() 670, 555 12 720, 548, 500 0 602, 540,(%) 515, 498 (u)lndicqfes that undercooling cccurred. (b)Questionable thermal effect. ' (C)Prepared at atmospheric pressure. 46 catiori of undercooling, Petrographic examination revealed three crystalline phases: ucl,, Ur,, and a bright greenish-blue phase with refractive indices near 1.747. The latter was the principal phase and is believed to be the UCI,F, campound. Other workers who have attempted to produce the com- pound by the fusion method'? have also failed to obtain a pure product, A probable melting point of 460°C, based upon their work, was reported’3 for the compound. UCI,-UF, Five compositions in the UCI,-UF, system have been prepared in sealed capsules. The thermal data are too meager to permit definite conclusions to be drawn, and petrographic examination of the fused melts was rather difficult, The 50-50 mole % mixture showed only one thermal effect, at 615°C, but it is not known whether the mixture was com- pletely liquid above this temperature, The fused mixture was predominantly a single phase with optical properties different from those of the two components; this suggests the possibility of a compound such as U,CI,F,. Such a compound would be a novelty, since, insofar as is known here, no mixed chloride-fluoride compound of tri- valent uranium has been reported. KCI-UCi, The KCI-UCI, system, studied by Kraus,’ has received considerable attention in this laboratory because thermal data obtained here 46 for a part of the system were at variance with the earlier work, The data, particularly the liquidus breaks in the 30 to 60 mole % UCI, region, have not been very reproducible. The reaction of UCH, with the nickel containers is responsible apparently for some of the difficulty. A crystalline growth, identi- fied as NiCl,, appears in the cooler part of the opparatus above the fused melt; apparently it is more volatile than the UCi, from these mixtures, 12J. C. Warf et al.,, Mixed Uranium Halides, CC.1785 (Sept. 10, 1944); N. W, Gregory, Compounds of Tuballoy Containing Two or More Different Halogen Atoms, R.L. 4,6,905 (Jan. 8, 1945). }3J. J. Katz and E. Rabinowitch, op. cit.,, p 542. 4R, J. Sheil and C. J, Barton, ANP Quar. Prog. Rep. Mar, 10, 1953, ORNL-1515, p 109, 15R. J. Sheil and C. J. Barton, ANP Quar. Prog. Rap. June 10, 1953, ORNL-1556, ¢ 42, I6R. J. Sheil, S. A. Boyer, and C. J. Barton, ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL.-1609, p 58. Petrographic examination of the fused melts indi. cated the probable existence of three compounds in this system, but only K,UCI, has been prepared in sufficient purity to be readily identified. The likely compositions for the other two compounds are KUCI, and KU,Cl,, but since they were not produced free of the other crystalline phases in the system, it seems probable that they melt incongruently. The diagram shown in Fig. 4.3, which is based upon both petrographic and thermal studies of the system, is believed to give a reasonably accurate representation of the phase relationships. No further work on the system is planned at the present time. RbCI-UCI,, The RbCI-UCI, system, for which preliminary data have been reported,!” has been studied con- siderably less than the KCI-UCI, system. Cooling curves with mixtures in the RbCI-UCI system con- taining 5 to 65 mole % UCI, yielded very scattered and confusing data, Petrographic examination of several fused mixtures did not aid materially in the interpretation of the thermal data because none of the crystalline phases observed in the fused melts had been prepared in sufficient purity for definite identification. However, the available dato seem to indicate that melting points ond phase relation- ships in this system are not markedly different from those shown for the KCI-UCl, system in Fig. 4.3. ”C. J. Barton and §. A, Boyer, ANP Quaor, Prog. Rep. Dec. 10, 1953, ORNL.-1649, p 52. OfiNL-! R-DWG 1532 TEMPERATURE (°C) LGl {rmole %) Fig. 4.3. Tentative Phase iDimgmm of KCI-UCH, System. | PERIOD ENDING JUNE 10, 1954 KCi-CuCl.UCI, A eutectic in the KCI-CuCl system that melts at 150°C was reported. A mixture corresponding to the reported eutectic composition (65 mole % CuCl) was prepared by using metallic copper to reduce the CuCl, in the CuCl. The resulting mix- ture, which showed a single thermal effect at 140°C, was colorless but contained some opaque inclusions in the crystalline material. No thermal effects were noted on the cooling curves when 2.5 or 5 mole % UCl; was added. Petrographic ex- amination of the fused mixture was unsatisfactory, however, since it disclosed only a mass of very fine crystals with opaque material in the aggre- gates, It is not certain, therefore, that the UCI, dissolved in the fused KCI-CuCl mixture. XRAY DIFFRACTION STUDIES OF THE NaF-ZrF, SYSTEM - P. A, Agron M. A. Bredig Chemistry Division A high-temperature x-ray diffractometer is being utilized to further understanding of the phase tronsformations that occur in the NaF-ZrF , system. The equipment, which was originally developed '8 for the temperature range of 1500 to 2200°C, was modified to cover the 400 to 900°C range. In order to avoid decomposition of binary complexes by partial vaporization, the solid salt is enclosed in a 15-cc beryllium container, The walls were ma- chined to a 10-mil thickness to allow free passage of the x-ray beams. This flanged container has been successfully gasketed to a stainless steel mounting with a combination of gold and copper rings. (In an initial aftempt to use a gold gasket in contact with the beryllium, there was rapid deleterious alloying at about 500°C,) The sample, about 0.4 g of solid material, is held in a flat, nickel holder. This arrangement was successfully tested on the CsCl transition. A more detailed description of this equipment, which is useful only for the examination of solids, will be given ot a future date. In the region of 33 to 40 mole % ZrF , several polymorphic forms of Na,ZrF, are indicated. At the 60% NaF-40% ZrF, composition the question of whether a solid solution in the cubic form ()} IBM. A. Bredig, Chem. Div. Quar, Prog. Rep. Mar. 31, 1951, ORNL.-1053, p 114, : 47 ANP QUARTERLY PROGRESS REPORT of NQZZrF{) {as in the analogous Na, UF , sys- tem'?) or a new compound exists, or both, has yet to be settled. Nine separate samples from this composition range were heated successively over temperature intervals from 450 to 575°C, A period of 1 hr at a fixed temperature appeared to be suf- ficient for equilibrating the solid phases. An analysis of the x.ray data obtained is pre- sented in Table 4.5, The choice of temperatures was influenced by the results of differential thermal andlyses and the quenching experiments described above, The data indicate that wide ranges of solid solutions exist in the regions examined. Diffuse scattering would have increased the background in the x-ray diffraction patterns if liquid was present along with the solid. It appears that one or two samples approximating 37 mole % ZrF, may have reached the liquidus region at 350°C, the highest temperature explored. Further experiments will be carried out before a phase diagram covering this region is drawn. The phase B-1 was observed for the first time in a high-temperature x-ray pattern and appeared to be easily preserved even on slow cooling, as indicated for the compositions of 33,3, 33.9, and 35.0 mole % ZrF, (Table 4.5). The diffraction pattern of the (X) phase at elevated temperatures appears to be better developed than the pattern observed at room temperature with a quenched sample, ‘T-12"", The agreement of the (Z) phase pattern with the pattern of a quenched ‘‘T.5" sample does not appear to be so good. The x-ray pottern at 495°C for the 40 mole % ZrF, composition indicates the presence of B-] and “Z5", The diffraction lines for the (40) phase are coincident with those of *Z5"" and therefore cannot be positively identified if the (40) phase is present with an appreciable amount of **Z5"". Alse, the presence of three phases would indicate that equilibrium had not been established. At room temperature the pattern was that for the (40) phase, some low-temperature polymorphs of “Z6’’, and a small amount of the **Z5"" phase. The petrographic examinationZ2? of this mixture at room temperature showed the presence of B-1 and the {(40) phase, but the (40) phase here gave a slightly higher index of refraction than that observed on the original sample. The above observations would indicate 195, J. Katz and E. Rabinowitch, op. cit., p 377-382. 2OISampIe examined by H. Insley, Consultant, Metal- lurgy Division. 48 that the (40) phase is not stable in the region of 495°C, Some structural relationships among the com- pounds in the NaF-ZrF binary system have been found to be similar to those in the NaF-UF / sys- tem, The stucture of the compound NaUF . has been given by Zachariasen'? as rhombohedral with 6 molecules per unit cell. The space group is R3 (ng‘)' The positions of the cations are given in general positions, that is, H{x y z) (y z x) (z x y), with the following parameter values: x y z 3 } 9 U A3 43 13 6 2 5 Na 13 A3 43 There is a pseudocubic cell which contains 3.7 metal atoms and 9.2 fluorine atoms, and its struc- ture closely resembles that of fluorite. The phase “75""21 is isomorphous with NaUF .. The relative intensities in the x-ray patterns of the uranivm and the zirconium compounds are similar, It has been noted?? that in kilogram preparations of the NaZrF, and of compositions containing 47 mole % ZrF, the latter nonstoichiometric compo- sition froze into a single glass-like phase, whereas the former invariably gave mixtures of the phases ““Z5'" and "'E2"’, For a discussion of the structure, a comparison of the directly measured density with the density derived from the x-ray data appeared necessary, Therefore the densities of 5 samples of the glass- like phase were measured, 23 and an average value of 4,11 + 0.01 g/cc was found, Sampling from different portions of the melt gave no apparent difference in density value, It had also been noticed that no change in cell size of the ‘‘Z5" structure was apparent in the range from 43 to 52 mole % ZrF4; however, some variations in intensity were observed. On the basis of 6 molecules per unit cell the calculated x-ray density is 4.04 g/cc, In order to resolve the discrepancy between the values 4.04 and 4.11, a structure model of ‘‘Z5" This model showed that it was possible to add another sodium cation at was built and examined. 21, tzsm phase refers to the extensive solid solufion range for the NaZrF5 molecule. 22G. W. Watson, Materials Chemistry Division, 235. l. Cohen, Reactor Experimental Engineering Di- vision, 6y TABLE 4.5. HIGH-TEMPERATURE PHASES OF THE NaF-Zrf, BINARY SYSTEM TEMPERATURE AND PHASES FROM HIGH-TEMPERATURE X-RAY STUDY COMPGOSITION . — , - (mole % ZfF4) Order of Heating Stages a b c d o f 33,3(® 485°C; B-11®) 25°C: B-1 475°C: B-1 503°C: (x)¢<) 575°C; sample major major expanded out of sample holder 33.9 460 to 470°C; 490°C;: B-1 25°C: B-1 530 to 535°C: 550 to 555°C: 520°C: (v) 8.1 major, plus major major (X)(c) (Y){d) unidentified 35.0 (1) 480°C: B-] 25°C: B-1 480°C: B.1 575°C: (z)y®) | a75°%C: 540°C: (¥) major major major plus o (F unidentified 35,0 (11) 485°C: B.1 500 to 503°C: 505°C: (X) 540 to 550°C: 494°C; B.1 plus **Z5'"(?) B-1 and {X)} (X) plus & plus unidentified 36.8 460°C: B-2(9) 495°C: B-1 507°C: () 523°C: 550°C: plus (40)“') plus unidentified plus unidentified unidentified unidentified 37.9 ~450°C; Be2 ~500°C: Bl 517°C: (x) 520°C: (X) 25°C; B-2, B.3,(" ptus (40) plus ““Z5"(?) plus unidentified and unidentified 40.0 460°C: —{40) 495°C; B-1,"Z5". | 25°C: (40) major, plus unidentified and {40)(7) “2610’ Qnd llzsll (a)Scmp'led fram composition C, a large batch preparation of qusz5° Remaining samples were taken from the hydrofivorinated '* X-mas tree’’ pre- parations., (B B} .phase shown petrographically to be bioxially pesitive, R.l. ¥ 1.41, Examined by G. D. White, Ceromics Department. (C}(X) phase related to *'T-12"" phase quenched in this temperature region, Quenches of these compositions reported by R, E. Mcore, Materials Chemistry Division, (d)(y ) phases not identified. {e (Z) phase may be related to "' T-5"" phase; quenched phase identified petrograghically, (f) Q. phase corresponds with cubic Na,ZrF . (Q)B-Z phase may be the polymorphic form of NQZZrFé just below B.l, LE W pattern of this (40) phase is similar to that obtained at room tempearature with a 40 mole % .Z.r'i:;‘ s.arnple.. (i)B 2 and B+3 are two biaxial opticelly negative forms of *'Z8"" normeally found at room temperature, rso6l '_'Ol INAF ONIAN3 QOl3d ANP QUARTERLY PROGRESS REPORT the (0 0 0) position and that it seemed feasible to also place an additional fluorine anion in the F> (00 I/2) positions of the pseudocubic cell with- out causing a noticeable change in the cell size. The density calculated for this model was 4.16 9/cc at 53.8% NafF -46.2% ZrF ;. Maximum thermal stability of stoichiometrically odd compositions of solid solutions was noticed in other cases,24+25 On further additions of NaF, the cell size appears to remain essentially constant, The structural model may explain this, if it is assumed that, below 46.2 mole % ZrF,, substitutions of No™ for Zr*" and vacancies in the anion positions occur and counterbalance each other with respect to cell size. (The introeduction of the larger sodium cation may be offset by the removal of fluoride ions.) The **Z5'" structure is ordered, and as more Naf is substituted into the unit cell the ordered arrange- ment is disrupted. At 60% NaF —~40% ZrF , on the NaF-rich side of a two-phase region the cations are in a disordered arrangement which gives a unit cell similar to what was refeired to above as the pseudocubic cell in the “‘Z5"" lattice. Here that unit cell contains, statistically speaking, 2.4 molecules of NaF and 1.6 molecules of ZrF4. The petrographic obser- vations2® of this phase at room temperature have indicated some slight birefringence. Quenching experiments and high-temperature x-ray diffraction studies give some indication that perhaps both phases exist near this composition, in different temperature regions. An attempt was made to evaluate the effect of changes in composition in the range between 44.2 and 30 mole % ZrF, upon the relative x-ray dit- fraction intensities. Factors were calculated for the ““Z5" structures for both the 6 NaF-6 ZrF (in molecules per unit cell) and the 7 NaF ~6 ZrF4 compositions by considering only the scattering by the cations. However, the omission of the contri- bution by the numerous anions may account for the lack of agreesment between observed diffraction intensities and those calculated from structure factors. A more favorable situation is envisaged for a computation of the fluorite structure with the inclusion of both cation and anion scattering. 25\1. A. Bredig, J. Phys. Chem. 46, 747 (1942). 26\.. G. Overholser, J. D. Redman, and C. F. Weaver, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL -1692, p 56. 50 CHEMICAL REACTIONS IN MOLTEN SALTS F. . Blankenship L.. G. Overholser W. R, Grimes Materials Chemistry Division Chemical Equilibrio in Molten Fluorides J. D. Redman C. F. Weaver Materials Chemistry Division Yalues for equilibrium constants ot 600 and at 800°C for the reduction of UF, in molten NaZrF by Cr° and Fe® were presented in the previous report.?® No comparable study has been completed during the past quarter; however, a number of experiments have been made to answer gquestions suggested by the earlier work. The solubility of FeF, in NaF-ZrF -UF, (50- 46-4 mole %) is 0.7 and 7.5 wt % at 600 and 800°C, respectively. These values, which agree well with those found previously for NaZrF ., show that small quantities of UF, have o negligible effect on the solubility of FeF . Attempts to measure the solubility of Fef, in NaZrF . at 800°C in nickel equipment demonstrated that the reaction 2FeF, + Ni —> Nif, + 2FeF, proceeded nearly to completion. Virtually all the iron appeared in the filtrate as FeF,, along with about 60% of the stoichiometric quantity of NiF,. Examination of the solid residue indicated clearly that the theoretical quantity of NiF, was formed but that solubility of this material at 800°C fimited the quantity which appeared in the filtered liquid. The solubility of Nif, in NaZrF 5 at 800°C appears to be about 2.5 wt %. Experiments at 800°C have shown that the re- action QFer + Fe®—> 3FeF2 proceeds essentially to completion in a short time; When no frivalent iron was found in the filtrate. the reaction 2Fe|:3 + 3Cr° o> 3CrF2 + 2Fe° was attempted, no FeF, and only 200 ppm (1%) of the original quantity of soluble iron which was added were detected in the filtrate. The Cr** concentration which was found approximated the theoretical concentration. Unsuccessful attempts have been made to measure the solubility of CrF, in NaZrF . and the NaF. ZrF (-UF , mixture described above. The molten salts were difficult to filter; a solid phase which clogged the filter or, as is somewhat less likely, a large increase in viscosity of the melt may have been responsible. The analyses of the few filtrates obtained were quite inconsistent. |t does appear, however, that virtually all the Cr3% is reduced to Cr** by reaction with the nickel apparatus. Some data: on the reactions Fe® 4+ 2CrF3—>2CrF2 + FeF2 and ' ' Cro + 2CrF, —> 3GiF, in molten NaZrF, at 600 and 800°C are shown in Table 4.6, The results obtained at 800°C indicate that in each case quantitative reduction to Cr*? occurs, It is not clear from these data what occurs at 600°C, Preliminary aottempts have been made to study the equilibrium Fe + 2UF4*‘.,.:'3 QU'F3 + F@F2 in molten NaZrF, by adding UF, and FeF, rather than Fe and UF | to the melt. Equilibrium concen- trations of the soluble constituents agree quali- tatively with those previously presented. iore detailed study of this system will be conducted in the near future. ‘ Preliminary attempts to determine the equilibrium PERIOD ENDING JUNE 10, 1954 turnings are used to reduce UF, in molten NaZrFs at 800°C have been made. The equilibrium concen- trations are considerably fower than those found when pure chromium is used; the activity of chro- mium in Inconel is apparently considerably lower than unity, Solubility of UF, in NaF.ZrF , Mixtures G. M. Watson C. M. Blood - Materials Chemistry Division The technique used in solubility experiments was, briefly, the following: About 3 kg of an appropriate NaF-ZrF -UF , mixture was prepared and purified by hydrogenation and hydrofluorination in the usual manner. The UF, was reduced to UF, by addition, under an inert atmosphere, of an excess of Zr® or U° to the melt at 800°C, followed by continuous sparging with purified hydrogen, usually overnight, to permit equilibration of the mixture. Samples of the mixture were then taken through filter sticks with the filtration medium consisting of sintered nickel. Equilibration times of at least 1 hr permitted each time the temperature was changed. The results obtained by filtration at seven tem- peratures in the range 600 to 900°C of o mixture containing 50 mole % NaF, 46.8 mole % ZrF,, and with continuous hydrogen sparging were 47 concentrations of CrF, and UF, when Inconel 3.2 mole % UF, are shown in Fig. 4.4. Since the TABLE 4.6. REACTION OF Ee® OR Cr° WITH CrF, IN MOLTEN NquFs(a) PRODUCTS OBSERVED IN FILTRATE REDUCINjG(fi‘)GENT TEMPERATURE Fo (wr %) T 7 N[ USED (°C) A Fe++ Total Cr++ Total {ppm) Fo 800 0.38 0.37 1.50 1.02(¢) 10 800 0.41 0.38 1.45 1.01(e) 10 600 0.07 0.18 0.79 0.53 335 600 0.08 0.08 0.69 0.33 80 Cr 800 1,53(¢) 1.510€) 10 800 1.55(¢) 1.47(€) 10 600 0.96 0.74(d) 90 600 0.75 ~ 0.55( 10 (b) c . )Theorehccl value. (a)'l wt % of Cr3T added as CrF3 to each sample. 5wt % of reducing metal added to sample. d)Low values may be due to low solubility of CrFy at 600°C. 51 ANP QUARTERLY PROGRESS REPORT ORNL-LR=-0OWG 1202 TEMPERATURE (°G) 700°C 600°C I I I URANIUM IN SOLUTION {wt %} o I ] FID N 8.0 8.5 2.0 9.5 10.0 1C.5 1.0 15 4 O ek Fig. 4.4, Solubility of UF, in NoF-ZcF . mixture, as prepared, contained 7 wt % uranium, it is obvious from the analyses that all the uranium which waos added was soluble above about 815°C. A composite sample of the residue which remained after removal of about 10% of the batch as samples contained 6.8 wt % uranium by analysis; it is apparent that the pretreatment and the considerable period at high temperatures affected the sample only slightly if ot all, The accuracy of the data obtained is not sufficient to establish that the curvature of the line below 800°C in Fig. 4.4 is real. Heats of solution as calculated from the slope of this curve range from 8300 cal/mole at the higher temperatures to 5600 cal/mole at the lower temperatures, Previous observations indicated that the solubility of UF, increased with increasing ZrF, concentration in the concentration range near 50 mole % ZrF . Filtration at 600°C of five different preparations by the general technique described above has quantitatively confirmed these obser- vations. The dato obtained are shown in Fig. 4.5. An increase in ZrF | concentration from 45 to 57 mole % apparently at least doubles the solubility of UF,. Since the UF, dissolved by the 57 mole % ZrF | mixture agreed within experimental error with the amount added, the solubility of UF, in this mixture at 600°C must be equal to or greater than 5.9 wt % uranium, 27,28 27F, F. Blankenship et al,, ANP Quar, Prog. Rep. Dec. 10, 1952, ORNL-1439, p 119. 28J. D. Redman et al,, ANP Quar, Prog, Rep, Mar, 10, 1953, ORNL-1515, ¢ 110, 52 ORNL-LR-DWG 11398 o o o o} & O URANIUM {wt 7%} 3 | o g o o 40 Zrf, (mole %) Fig. 4.5. Solubility of UF at 600°C in NaF. ZrF4. The increase in solubility of UF, with increasing ZrF, concentration implies that UF, is not com- plexed strongly by free fluoride ion but that dis- sociation of UF, is aided by the presence of free ZrF . Whether disproportionation of UF, into UF, and U° is likely to become apprecicble at a lower temperature in such melts is, as yet, unknown. Some auxiliary experiments performed in smaller scale apparatus by other members of the ANP Chemistry Group (R. J. Sheil and C. J. Barton) have shown that a mixture prepared to contain 52.7 mole % NaF, 40.9 mole % ZrF, and 6.4 mole % UF, when filtered at 600°C yields a filtrate con- taining only 1.4 wt % vuranium, The residue from this filtration comprised about 20% of the total sample and was predominantly UF. with some yellow crystals that were shown to be mainly NagZr F ., with opaque material which was almost certainly Zr°® or ZrH,, These data, which are in general agreement with those cited above, serve to demonstrate that UF, and not a complex uranifer- ous fluoride is the primary phase in the concens tration range near 50 mole % ZrF . Solubility of UF; in NaF-KF.-LiF Mixtures G. M. Watson C. M. Blood Materials Chemistry Division Two experiments were performed to determine the solubility of UF; in the NaF-KF-LiF eutectic. In both experiments, the UF, was obtained by reduction in situ of the dissolved Ul , with uranium metal, The technique used in obtaining the filtrates for solubility determinations was otherwise identi- cal to that used in the corresponding experiments with NaF-ZrF | mixtures, as described above. When a two-fold excess of metallic uranium was used, potassium metal was distilled from the re- actor at temperatures of 800°C and above. When tilter sticks were introduced into the melt, the filter media immediately became clogged. Filtrates weighing only 1 to 15 g could be obtained in contrast to - the 50.g filtrates obtained in normal operations. At the end of the experiment, a metal- lic alloy was observed on the reactor walls and as a crust on the solidified melt, The formation of potassium is explained when the equilibrium constant of the reaction KF {solution) + %U (crystal). = K (vapor} + %UFS (solution) is calculated from published values of free energies of formation. At 800°C the equilibrium partial pressure of potassium, with the assumption of ideal solutions, may be estimated to be from 15 to 20 mm Hg. The constituents of the metallic alloy were identified as nickel and uranium. The alloy forma- tion may be rationalized from consideration of the U-Ni phase dicgrum,29 which shows that eutectic melting of nickel and uranium can occur at about 738°C, The eutectic phases are UNi and U Ni. In order to avoid or minimize side effects in the second experiment, only the stoichiometric quantity of uranium metal was added with 10% excess of zirconium metal, The mixture was heated to only 750°C, Much mere satisfactory filtrates were ob- tained (an average of about 45 g above 600°C), and no metallic alloy was observed. The concentrations of uranium in the filtrates and in residues of beath experiments are summarized in Table 4.7. Petrographic examination of composite samples of the residue from each experiment faoiled to show the presence:of tetravalent uranium but detected (in quantity) a bright-orange phase believed to be a KF-UF, compound. The tabulated results appear to show essentially constant uranium concentration at the temperatures studied, This effect would be shown if all the uranium were in the liquid phase throughout the temperature range. Unfortunately, the results are not consistent between experimenfs and the filtrate (or residue) analyses do not agree 29The Reactor Handbook, Vol. 3, p 495, Technical Information Service, AEC, 1953, . ................................................................................................................................................................................. PERIOD ENDING JUNE 10, 1954 TABLE 4.7, SOLUBILITY OF UF, IN NaF-KF-LiF EUTECTIC TEMPERATURE | SOLUBILITY OF UF, (wt % U) S R-153* R-154%* 800 15.4 750 17.3 16.0 700 17.3 16.1. 650 17.3 163 600 17.2 16.2 550 164 208 15,9 Residue 20.1 | 17.8 - *22.0 wt % uranium added. **18.6 wt % uranium added. within experimental error with the amount of uranium added. [t is obvious, however, that the solubility of UF, in the NaF.KF-LiF eutectic is higher than 15 wt % uranium, Solubility of UF, in NaF-RbF-LiF Mixtures R. J. Sheil - C. J. Barton Materials Chemistry Division The solubility of UF, in the NaF-RbF-LiF eu- tectic has been determined at three temperatures by equilibration of a melt to which sufficient UF, and excess uranium metal had been added to produce 4 mole % UF,, filtration of a specimen at the desired temperaoture, and chemical analysis of the resulting solution. Because of the high RbF content of these mixtures, they are extremely hygroscopic; great care in handling the samples is necessary fo obtain reliable values for UF, solu- bility. The data obtained are shown in Table 4.8. The values obtained show that 597°C is probably above the liquidus temperature for this composition; ¢ wt % of uranium is probably not the maximum that can be dissolved at this temperature. Petrographic examination of the filtrates indicated that no UF, or complex fluorides containing tetravalent uranium were present, The samples contdined no free UF; no identification of the complex U3* fluoride has yet been made. ' 53 ANP QUARTERLY PROGRESS REPORTY TABLE 4.8, SOLUBILITY OF UF5 IN NoF-RbF-LiF EUTECTIC TEMPERATURE SAMPLE OF UF, ~— ©C) FIL TERED (%) (wt % U) LiF RbF NaF UF3 497 67 4,4 43.8 46.8 8.2 1.2 552 90 6.8 39.6 47.6 10.7 2.1 597 99 2.0 41.1 46,5 9.6 2.84 + + + *Calculated from analyses for Li , Na', Rb, and total U. Chlorination of UF3 G. M, Watson C. M. Blood Materials Chemistry Division Analytical determinations of the UF; dissolved, with or without added UF ,, in fluoride melts gave values considerably lower than those anticipated from petrographic examination or from the stoichi- ometry of the preparation reactions, The descrip- tion3® of the preparation and the properties of UF,Cl suggested that chlorination of the UF, to this compound ond subsequent determinations of Cl~ in the product might be successfully used, Attempts to chlorinate 1- to 2-g portions of a 200-mesh UF;-UF, mixture contained in ceramic boats in a glass-tube reactor by exposure to dry Cl, at 350°C for various intervals of from 2‘/2 to 20 hr yielded a green product of uniform appear- ance. Petrographic examination showed the product to be markedly birefringent and uniaxially (or possibly biaxially) negative, with refractive indices of 1.56 to 1.539. X-ray diffraction revealed no free UF, but did reveal an unknown diffraction pattern unrelated either to UF, or UF,. Chemical analyses of samples chlorinated 21{2 to 41/2 hr indicated 5.7% total chlorine; since this is roughly one-half the expected value for chlorine, it appears that the chlorination was incomplete or that some hydrolysis and oxidation of the sample had occurred, Attempts will be made to improve the contact between the powder and the Cl, by passage of the gos through the solid contained on a sintered-glass filter, |If these attempts are successful, studies with UF ;-bearing fuel mixtures will be attempted. 30) . Warf and F. Edwards in Chemical Research — Generol Chemistry Report for Period of March 10 to April 10, 1944, CC-1496 (May 11, 1944); J. C. Warf et al., Mixed Uranium Halides, CC-1785 (Sept. 10, 1944), 3]\/. S. Coleman, C. J. Barton, and T. N. McVay, ANP Quar. Prog. Rep. Dec. 10, 1952, ORNL.-1439, p 117, 54 Preparation of UF.2Z/F F. P. Boody H, A, Friedman Materials Chemistry Division The compound UF,.2ZrF, was identified®! in 1952 as one of the materials obtained on cooling ZrF ,-bearing melts containing large quontities of UF,. Since large quantities of the compound had not been prepared, the preparation of a 3,5-kg batch was attempted to make available a quantity suf- ficient for measurement of vapor pressure and other physical properties. The method used was a modi- fication of the standard hydrofluorination procedure which involved reduction of a UF -bearing mixture of proper composition with zirconium metal chips at 850°C, The product appeared homogeneous and was a dark reddish-brown in color, Only 655 g was filtered, since unavoidable cold spots in the trans- fer line permitted freezing of this relatively high- melting-point material (740°C). The compound ma- terial, unlike virtvally all other melts observed here, appeared to expand slightly on freezing. Petrographic examination of the product showed it to be approximately 99% homogensous, red- orange, uniaxially positive UF;.2ZrF, and about 1% free UF,. X-ray diffraction analysis confirmed the predominant phase to be UF;.2ZrF,, with small amounts of free UF,. On warming in air, the color of the product changes from orange-red to oliveegreen and the refractive index increases, Structurally the crystals do not appear to be altered by the color change, and no evidence of oxidation other than the color is visible. Preporation of Various Fluorides B. J. Sturm E. E, Ketchen .. G. Overholser Materials Chemistry Division Several of the simple structural-metal fuorides, as well as complexes of these materials with the alkali-metal fluorides, were prepared for corrosion experiments and some physical and chemical property testing., In an effort to prepare the struc- tural-metal fluorides in a high state of purity, attempts were made to refine crude NiF, and CrF, by treatment with BrF,. In a modification of the apparatus originally assembled by the ANP Ana- lytical Group, the crude material was treated with Brf , ot 350 to 400°C for 1to 3 hr, and the product was outgassed by thorough pumping at about 350°C. Analyses of 10 samples indicated that considerable conversion of oxides and/or oxyfluorides was achieved by this procedure; however, the purity of the product was still not of the order desired. Large quantities (1.5 to 2 kg) of K3CrF6 and of K,FeF, were prepared by heating appropriate qucn’rmes of 2CtF,.7H,0 and FeF,.3H,0 with KHF,. Several bctches of FeF,, FeF,, FeCl,, NiF,, NiCl,, CrF,, and CrF, were syn'rhesued by mefhods prewously described. FUNDAMENTAL CHEMISTRY OF FUSED SALTS EMF Medasurements | L. E. Topol Materials Chemistry Division Measurements of decomposition potentials of KCi and of a variety of chlorides in solution in molten KCI at 850°C were attempted, Cathodes of platinum or nickel and anodes of carbon, nickel, zirconium, platinum, electrolytic chromium, Inconel, and stoinless steel have been employed. Morganite (Alzos) crucibles were used to contain the melt in each case, The solutions were prepared by heating 2 moles of KC! with 1 mole of the desired anhydrous chlo- ride to above the melting point of the mixture in a sealed, evacuated quartz tube. More dilute so- lutions were prepared by the addition of KCI to the materials so obtained. The data obtained are tabulated in Table 4.9, If the value of 0.90 volt for electrolysis of KCl with uranium anodes is combined with that of 3.15 volts for electrolysis of this material with inert anodes, E for the reaction x U + 5 Cl, —> UCl_ appears to be 2.25 volts. This value agrees well with the estimated E° (2.26 volts) for UCl,. The value (1.27 volis) obtained by electrolysis of KCli 'PERIOD ENDING JUNE 10, 1954 with an anode of electrolytic chromium is less than that previously obtained (1.5 volts) with an anode of Cr° plated on gold wire.?2 The potentials observed for electrolysis of KCI with Inconel electrodes suggest that Cr° is dis- solved first and that, subsequently, Ni®is attacked; no dissolution of Fe® was detected. However, when the anode is a 300-series stainless steel, it appears that first iron,and then chromium,is dissolved. Three possible reactions of CrCl,, 4 (n 3CCl, - Cr? 4+ 2GCH, = (.84 volt, (2) CrCI2 -3 Cr° + CI2 , E° = 1.36 volts, (3) CrCI + Ni® -—-}NICIZ + Cr = 0, 54 volt, may qualitatively explain the experimen’ral obser- vations on electrolysis of solutions of CrCl,. It appears that with a nicke! anode, reaction 3 occurs at both CrCl, concentrations tested. With the inert carbon anode, reaction 1 is controlfing at high concentrations of CrCl,, while reaction 2 is im- portant in the dilute solutions. The experimental electrolysis of ZrCl, solutions with nickel anodes checks well with E° values for reactions producing ZrCl,, ZrCl,, and Zr° with NiF,, as described prewously. 32” The values ob- ramed with the graphite anodes may well be due to the reactions (4) L, — I 4 2C, (5) Z0, + C—Zr" + CO, , for which the best estimates of E°are 1.9 and 1.24 volts, respectively. Attack on some of the graphite anodes indicates that reaction 5 occurs. The electrochemical behavior of UCI be similar to that of CrCl appears to bt is likely that the reactions (6) 4uCi, — 3UCl, + U° , : E° = 1.08 volts, (7) - 2UCH, —> 2U° + 3CH, | E° = 2.26 volts, (8) 2UCI, + Ni —=3NiCl, + 2U° , E° = 1.44 volts, 32, E. Topol, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 63. _ | 55 ANP QUARTERLY PROGRESS REPORT TABLE 4.9, OBSERVED DECOMPOSITION POTENTIALS FOR REDUCIBLE CHLORIDES IN MOLTEN KCI AT 850°C SOLUTE COEEES‘;TLTGEON ELLECTRODES BLANKETING | POTENTIAL OBSERVED ADDED Cathode Anode ATMOSPHERE (volt) {mole %) None Pt U He 0.89 Pt U H2 0.90 Nij Zr He 1.03 Ni Cr* H2 1.27 Ni Incone! H, 1.85, 2.03 Ni Inconel He 1.85, 2.03 Ni Type 316 H, 1.55, 1.85 stainless steel Ni Type 347 H2 1.43, 1.85 stainiess steel CrCI2 33.3 Ni C H2 0.88 to 0,92 33,3 Ni Ni H:2 0.53 1o 0.57 2.0 Ni C He 1.30 to 1.43 2,0 Ni Ni He 0.65 to 0.69 ZirCI4 33.3 Ni C H2 1.67 to 1.80 33.3 Ni Ni H2 0.50 to 0,57, 0.80, 1,01 25.0 Ni C H2 1.33, 1.80 to 1.90 25,0 Ni Ni H2 0.63, 0.94 to 1.06 UCI3 100 Pt Ni H, 0.80 to 0.84, 1.01 to 1.10 100 Pt Ni He 0.80 to 0.87, 0.98 33.3 Pt C H, 0.78 to 0,85 33.3 Pt Ni H, 0.80 to 0.82 33.3 Pt Ni He 0.83 to 0.85 33,3 Ni C I'12 0.68 te 0.72 33.3 Ni Ni H2 0.78 to 0.82 2.0 Pt Ni H2 1.25 10 1.35 2.0 Pt Pt He 2.10 2.0 Pt C He 1.15, (2.08), 2.48 to 2,56 1.0 Pt Ni He 1.36 *Electrolytically prepared. are responsible for most of the effects observed. The ogreement between volues under He and H suggests that UCI, is only slightly affected by H, in the dilute solutions with relatively poor contact efficiency used, When nickel cathodes were em- ployed, an alloy of uranium and nickel resulted. The general constancy of the potential obtained with both pure UCL, and K,UCI, seems to indicate some solubility of UCI, in ucl,. 56 A zirconia crucible (fabricated by Titanium Alloys Division of National Lead) was heated at 750°C for 4 hr in a hydrogen atmosphers with a charge of NofF-KF-LiF-UF,. The vessel proved to be some- what porous; the only evidence of attack observed was a slight reaction to produce UO, at the cru- cible melt interface., It is hoped that vessels of this material may be useful for electrochemical studies of fluoride melts, Physical Chemistry®? E. R. VYan Artsdalen Chemistry D:wsnon The density and electrical conductivity of the molten system KCl-NaCl have been determined over the entire composition field from temperatures a little above the melting point of the individual mixtures to about 1000°C. Density can be ex- pressed as a linear function of temperature, and the specific conductance can be expressed as a quad- ratic function of temperature. A plot of the molar volume vs the composition shows a slight positive deviation from linearity, with a maximum deviation of about 0.5% at 50 mole % KCl. Equivalent con- ductance, A, may be expressed in the form 1000 1000 C + A —-:I-:———-+ 2302.6B IOg -—-*%‘"“* ’ log A = where T is‘in degrees Kelvin, C is a constant of integration, and A and B are the intercept and the slope, respectively, of the straight line which results from the graphical differentiation of log A vs (1000/T). The possibility that this form of equation is general in nature and applicable to many molten salt systems is being investigated. A plot of equivalent conductance vs mole per cent KCI shows that the system is far from additive but that it resembles the molten KCI-LiCl system re- ported previously.®4 There is a very shallow minimum in equivalent conductance for the KCl- NaCl system at about 85 mole % KCl, The expla- nation for this is apparently the same as the one advanced in the case of the molten system KCI- LiCL An emf cell with transference has been designed for use with fused salts. With such a cell it is possible to make the potential measurements re- quired for obtaining important cross-checks on transference numbers and activity coefficients of ions in melts, An initial study with the molten system NaNO;-AgNO, indicated that useful results can be cbtmned by measuremenfs of cells with transference. However, the attainment of good reproducibility poses a difficult problem. Radiotracer techniques are being used for making measurements of the self-diffusion coefficient of 33For details of this work see Chem. Div. Semiann. Prog., Rep. Dec., 20, 1953, ORNL.1674, and Chem. Div, Semiann, Prog. Rep. June 20, 1954 {to be published). Mg R, Van Artsdalen, ANP: Quor. Prog. Rep. Dec, 10, 1953, ORNL 1649, p 58. _ information - and PERIOD ENDING JUNE 10, 1954 This work has necessitated the development of new techniques for handling molten materials as well as for pre- individual ions in molten salts. cision control of high-temperature furnaces over periods of several days. Measurements of the self- diffusion of sodium ions in molten sodium nitrate indicate that the diffusion coefficient is of the order of 1 or 2 x 10~° sq cm sec™ ! at temperatures a little above the melting point. This work is being pursued actively and is expected to yield important results applicable to transport processes. A high-precision, adiabatic, heat-capacity calo- rimeter has been built for operation in the range up to about 1000°C. AH preliminary tests have been completed and the calibration should be com- pleted soon. It is anticipated that heats and free energies of formation can be determined with good accuracy for a number of important reactor ma- terials. In addlhon\ fundamental information should be obtained concerning the solid state as well as the molten state of salts. CHEMICAL EFFECTS OF FISSION PRODUCTS R. F. Newton ANP Division Other than the absorption of neutrons, the pos- sible undesirable effects of fission products in a reactor are the decrease of heat transfer by sepa- ration of o solid phase on heot exchange surfaces and by change of viscosity or heat capacity of the molten mixture, and the acceleration of corrosion. Any such effects will be dependent on the QUanti- ties of fission products which accumulate. Quantity of Fission Products The fission products were calculated on the assumption that about 0.1 mole of uranium under- goes fission per kilogram of melt at a uniform rate during 1000 hr. The quantities given in Table 4,10 are for the end of the 1000-hr period. The yields for the various products were taken from the smooth curve of Siegel.3® The quantities are as large as or farger than any which occur at earlier times and not greatly different than they would be if the fissions were spread over a greater time period. Separation of Solid Phases Of the elements listed, molybdenum, ruthenium, rhodium, ‘and palladium are expected to separate 35). M. Siegel, J. Am. Chem. Soc. 68, 2411 (1946). 57 ANP QUARTERLY PROGRESS REPORTY TABLE 4.10. FISSION-PRODUCT YIiELD AT END OF 1000 hr PRODUCT YIELD* PRODUCT YIELD PRODUCT YIELD Se 0.4 Ch 7.0 | 1.2 Br 0.25 Me 18.6 Xe 18.4 Kr 3.5 Te 5.1 Cs 16.9 Rb 4.0 Ru 14.0 Ba 6.7 Sr 12.2 Rh 1.0 La 6.5 Y 5.3 Pd 1.3 Ce 26.0 Zr 24.8 Te 1.4 TRE** 53.3 *In atoms per 100 fissions occurring uniformly over 1000 hr. **Total rare eavths. as the metals. Molybdenum and palladium alloy with nicke!, and it is presumed that ruthenium and rhodium do likewise. If they form a solid solution with nicke! or nickel alloy and slowly diffuse into the heat-exchanger wall, they will probably do no harm; if they deposit on the surface and remain there, they may reduce heat transfer becouse of roughening of the surface. The possible separation of either simple or com- plex solid fluorides, because of their poor heat- transfer properties, would be much more serious than the deposition of metal. Rubidium may be dismissed immediately, since it is o desirable constituent of the fuel; chemically, cesium is also desirable. Zirconium likewise may already be present in the fuel, in which case the zirconium formed by fission will change the relative amount of zirconium in the fuel to a negligible extent, If zirconium intentional constifuent, it appears impossible that it could precipitate from a fluoride-based fuel not containing oxygen. fluoride and sodium fluoride form a eutectic at 820°C that contains 35 mole % BaF,. The extrapo- lated solubility of the eutectic is 14 mole % at 600°C, Corresponding data are not available for SrF,, but an extrapolation from the eutectic of CaF,-NaF gives a solubility of CaF, at 600°C of 12.3 mole %. Since the properties of strontium compounds are generally between those of calcium and barium, it is safe to assume a solubility of at least 10 mole % for the strontium compounds, thot is, nearly 100 times the amount produced. is not an Barium The only remaining substances worthy of con- sideration are the fluorides of yttrium and the rare earths. These substances are so similar chemi- 58 that they undoubtedly form a continuous seties of solid solutions and can practically be treated as a single compound. From the phase diagrams of alkali-metal fluorides with LaF, ond with YF3, the lowest solubility (YF3 in KF) ex- trapolated to 600°C is 8 mole %. Since the total of rare earths plus yttrium is of the order of 0.5 mole %, even here, there seems to be an ample factor of cally safety regarding the possibility of precipitation. Effects on Yiscosity and Heat Capacity The replacement of one wuranium atom by two atoms of fission products would be expected to increase the heat capacity per gram, but the density would be decreased at the same time, with very little change in the heat capacity per cubic centi- meter. Similarly some fission products would in- crease the viscosity while others would decrease it. In ony case, it is safe to predict that the changes will be no greater than about 1%, Effects on Corrosion It is apparent that the total number of equivalents of cationic fission products exceeds that of anionic fission products plus the fluoride made available by fission of the uranium (assuming the uranium as UF ) only if molybdenum is included as one of the cationic elements. However, it the relative affinity of molybdenum and nickel for fluorine is considered (Glassner3® gives free energies of formation of ~56 and ~63 kg-cal/gram-atom of fluorine for MoF and NiF,, respectively), it is evident that molyb- 36A. Glassner, A Survey of the Free Energies of For- mation of the Fluorides, Chlorides, and Oxides of the Elements to 2500°K, ANL-5107 (Oct. 22, 1953). denum will appear as the element along with In such cases, the fission of uranium occurring as UF ruthenium, rhedium, and palladium. . 4 will' be accompanied by the solution of a corresponding amount of nickel, chromium, or other wall con- stituent. The imbalance can be corrected by using a fuel in which one-half of the uranium which undergoes fission is UF,. Since it is anticipated that about 20% of the uranium. present will undergo fission, the change required to prevent solution of nickel because of this cation-anion imbalance is to have 10% or more of the uranium present as UF,, From preliminary data available on the solubility of UF,, it appears that even the zirconium-base fuels should be capable of dissolving this amount, while the alkali-metal-fluoride fuels possess a large margin of safety in this regqrcfl PURIFICATION AND PROPERTIES OF ALKALI HYDROXIDES .. G. Overholser F. Kertesz Materials Chemistry Division Purification of Hydroxides E. E. Ketchen L. G. Overholser Materials Chemistry Division Eight batches of NaOH were purified by filtering a 50-wt % aqueous solution through a fine sintered- glass filter to remove the Na,CO, and then dehy- drating at 400°C under vacuum. The product, as in the past, contained less than 0.1 wt % of H,0 and of Na,CO,. Four batches of commercial LiOH were dehydrated at 200°C under vacuvum to yield material containing less than 0.1 wt % of H,0 ‘and 0.2 wt % of Li,CO;. Approximately 5 Ib of potas- siuvm, after bemg dwnded into l/-I‘o portions, was reacted with water, and the resulflng solutions of KOH were dehydrated at 400°C under vacuum, The dehydrated KOH contained approximately 0.1 wt % of Na and less than 0.1 wit % of K,CO, and of H,0. Reaction of Sodium Hydroxide with Metals F. A. Knox H. J. Buttram F. Kertesz Materials Chemistry Division The hydrogen equilibrium pressures developed in the sodium hydroxide~nickel reaction were further investigated with the use of the previously de- scribed quartz-jacketed nickel reaction tube which PERIOD ENDING JUNE 10, 1954 was connected to a manometer. [t was planned that hydrogen would be added to the jacket to reduce the time necessary for establishment of the pressure equilibrium, but in practice this arrangement proved unsatisfactory, The hydrogen added to the jacket intfroduced an imbalance in the system and thus increased - the pressure inside the reaction: tube. It was also recognized that errors were introduced in the previous measurements by hydrogen being removed when the system was evacuated during heating up to 600°C, because the sodium hydroxide was thus subjected fo heat and to lowering of the pressure, - The current measurements are being made by evacuating the system at near the melting point of the hydroxide. The redetermined values, which seem to approximate steady-state conditions, were found to be higher than those reported previ- ously. The redetermined values are given in Table 4,11, These data yield a straight line when log P,, s plotted against 1/T as indicated in Fig. 4.6. In addition to these measurements of the hydrogen pressures developed during the reaction, an attempt was made to study the mechanism involved by determining the amount of metal dissolved. In the first series of experiments, nickel copsules con- taining hydrogen were sealed in quartz envelopes; in the second series, the capsules were unjacketed and the evolved gas was allowed to escape, The capsules were heated for periods which ranged from 2 to 256 hr., It was found that the nickel concenfrqhon in the hydroxide in the capsules with the quartz jackets reached o plateau rather early and remained con- stant during the prolonged heating periods. It may be assumed that the nickel concentration reached an equilibrium value in the jacketed system. The nickel concentration continued to increase with time in the unjacketed capsules, ' TABLE 4.11. HYDROGEN PRESSURE OVER NaOH IN NICKEL TEMPERATURE PRESSURE (°C) {mm Hg) 600 46 700 77 800 | 126 900 | 160 1000 205 59 ANP QUARTERLY PROGRESS REPORT Similar series of experiments with sodium hy- droxide in copper capsules showed similar be. havior. After 256 hr of exposure at 600°C, the copper concentration in the jocketed capsules was about 700 ppm, while it was as high as 8000 ppm in the wnjocketed capsules. For the jacketed capsules the copper concentration rose to about 800 ppm at 700°C and 1100 ppm at 800°C., The unjacketed capsules that were heated at 800°C in a helium-filled dry box had relatively low copper concentrations, but the walls were covered with copper crystals., This is the first time that mass transfer of copper has been observed under these conditions, A larger percentage of the nickel impurity that originally existed in the copper capsule was found in the transferred metal than in the wall material of the capsule. PRODUCTION OF PURIFIED MOLTEN FLUORIDES F. F. Blankenship G. J. Nessle Materials Chemistry Division Laboratory-Scale Production C. M. Blood G. M. Watson F. P. Boody H. A. Friedman Materials Chemistry Division A total of 28 batches of various molten fluorides was processed during the quarter; these materials were used for phase studies, for corrosion testing, for chemical separations research, and for studies in kinetics of high-temperature chemical reactions, The largest usage of these purified materials was in the studies of UF, solubility, Interest has recently developed in the alkali tluoride eutectics as UF, solvents, Since there had previously been some difficulty in preparing the NaF-KF-LiF eutectic in a high state of purity by the hydrogenation-hydrofluorination technique without excessive corrosion of the apparatus, a study of the behavior of these materials during purification was attempted. Two batches of the NaF-KF-LiF eutectic have been prepared successfully in apparatus which was assembled with special care to avoid leaks, The solid fluorides were admitted to the apparatus and dried at 350 to 380°C under flowing helium before elevation of the temperature, The fluorides were mainfained under hydrogen during the heating to 800°C and HF was admitted only after this temper- oture was reached, One batch was treated for 100 60 UNCLASSIFIED ORNL-LR-DWG 1333 TEMPERATURE (°C) 1000 900 800 700 600 300 |- 200 |- 100 | | 2 I T I £ £ L £ | AH=2.3R(2000)] 7 29200 cal | oy Ut — SR - o . 1 20 - - f T } 10 ‘l . ! ! I _____________ 06 07 08 0% 10 11 1.2 1.3 ‘/r(oK)Xios Fig. 4.6. Hydrogen Pressure over Sodium Hy- droxide in Quartz-Jacketed Nickel Capsules. min with HF and stripped with 500 liters of H,; the other batch was treated with HF for 70 min and stripped with 600 liters of H,. The HF concen- tration in the exit hydrogen from the first batch was 2.5 x 104 mole/liter and from the second, 1.8 x 104 mole/liter. The color of the product wes dead white in each case, and the product appeared to be quite homo- geneous, Chemical analyses of the products have not been completed. Although there were deposits of metal around the thermocouple well, on the reactor walls, and as a partial plug in the gas-inlet tube that were shown spectrographically to consist of NaF, KF, and LiF, along with an alloy of nickel and iron, attack on the container seemed to compare faverably with that observed in processing NaF- ZrF , mixtures. Therefore it appears that the alkali- metal-fluoride melts can be prepared in a high state of purity by existing techniques if the pro-~ cedure described above is followed. Experimental Production J. E. Eorgan C. R, Croft J. Truitt J. P. Blakely Materials Chemistry Division A total of 162.7 kg of various fluoride compo- sitions was prepared in the experimental production facility and dispensed to various requesters in the ANP program, The pilot-scale production facility which will somewhat more than double the present pilot-scale capacity is nearly completed and should be available for operations before June 1. The experimental production facility will then be adapted for safe operation with BeF . Large-Scale Production F. A. Doss R. Reid R. G. Wiley ANP Division M. S, Freed Pratt & Whitney Aircraft Division F. L. Daley E. F. Joseph E. L. Youngbiood J. P. Blakely Materials Chemistry Division To meet the markedly increased demand for processed fluorides, the 250-ib production facility was placed on three-shift operation, and a total of 1358 kg of fluorides was processed. Most of this material was transferred to containers which ranged in size from 2 to 25 kg. Proft & Whitney Aircraft Division received opproximately 450 kg of this material, and, except for about 20 kg that was delivered to Battelle Memorial Institute, the re- mainder was dispensed to requesters in the ANP program at ORNL. The increasing demands for purified fluondes necessitated, early in this calendar year, an in- creased output of ZrF4 from the production plant operated by Y-12 personnel. With the available equipment, this production increose could be ac- complished only by decreasing the hydrofluorination time of ZrCl in the converfer and by increasing the temperature of sublimation of the crude ZrF . The quality of the ZrF , was, accordingly, reduced; the iron content of the ZrF , increased from a mean of 250 to about 1000 ppm, while the nickel content rose from about 50 to about 200 ppm. Although it is possible to prepare fluoride mix- tures to rigorous specifications frem this ZrF“, use of this material has reduced the efficiency of the production facility to a level considerably below that obtained during 1953 when the ARE fuel solvent was being processed. - For example, the gas-inlet tubes frequently become plugged by PERIOD ENDING JUNE 10, 1954 crystals of nickel and iron during the hydrogen- flushing operation. Since the amount of iren to be stripped (by reduction of FeF, with hydrogen) is much larger than that previously considered normal, the time required for the hydrogen stripping has increased from cbout 30 to nearly 100 hr, and hydrogen requirements have risen from 10,000 liters to nearly 25,000 liters per 250-Ib batch of fluoride. Since the anticipated demand for fluorides will probably preclude a slower production rate of ZrF, and since improved purity at the present production rate would require expansion of the facility, aftempts are being mode to obtain pure Na,ZrF from commercial sources. While ZrF4 must be added to this material to reach the fuel compo- sitions required, the amounts of ZrF , needed would be well within the capacity of the Y-12 Plant for ZrF , of acceptable purity. It is anticipated that pure Zer (or its equivalent) will be available within the next six to -eight weeks. Should no commercial concern be able to meet purity specificotions (ot a competitive price), it may prove more economical in the long run to use hafnium-free ZrF, for all fuel production with- out regard to intended use. This rather startling conclusion (based on price of production of about 3500 Ib of hafnium-free material} is dve, in part, to the better initial purity of the hafnium-free ZrCl, and, in part, to its higher bulk density which permits larger loadings of converters and sublimers and, accordingly, larger throughput of material in existing ZrF , production equipment, Production of Enriched Fuel for {n-Pile Loop J. E. Eorgan Materials Chemistry Division To provide the fuel for an in-pile forced-convec- tion loop (cf., Sec. 8, “Radiation Damage’’}, two batches of about 6 kg each of o mixture of 62.5 mole % NaF, 12.5 mole % ZrF,, 25 mole % UF were prepared from product-level UF, by Y-12 personnel with some technical assistance from the ANP Chemistry Group. Actual production of the material was accomplished during the interval March 15 to March 29, 1954, 61 ANP QUARTERLY PROGRESS REPORTY 5. CORROSION RESEARCH W. D. Manly Metallurgy Division W, R, Grimes . Kertesz Materials Chemistry Division H. W. Savage ANP Division The static and the seesaw corrosion testing fa- cilities were used for further studies of Inconel, stainless steel, various cermets, ond graphite exposed to fluoride mixiures or liquid metals. Chromium metal additions to fluoride mixtures with and without uranium contained in Inconel capsules have been shown to give some protection to the Inconel. Experiments conducted at various temper- atures confirmed previous indications that the chromium content of a fluoride mixture contained in Inconel increases to a maximum and subsequently decreases with increasing temperature. The es- sentially linear relationship between chromium concentration of the fluoride mixture and the surface area of the exposed Inconel indicates that equi- Jibrium concentrations are not attained when Inconel specimens are exposed to fluoride mixtures in graphite crucibles., Fluoride mixtures containing UF, were tested in Inconel capsules in the seesaw apparatus and, as in thermal-convection-loop tests, no evidence of attack on the Inconel wos found. Experiments are under way to determine the effects of fission products on the corrosion of Inconel by fluoride mixtures. Tests have shown that when Inconel and type 316 stainless steel are exposed in the same system to fluoride mixtures, the steel is inferior to Inconel in its resistivity to attack even when the steel is in the cold zone., Static tests confirmed the re- ported increase in corrosiveness of lithium metal when lithium nitride is added. Impregnation of the surfaces of steels with chromium did not improve the corrosion resistance of specimens tested in sodium and several liquid-metal alloys. Extensive tests of various cermets that are being considered for applications as bearing materials in contact with fluoride mixtures are under woy. The cor- rosion resistance of cobalt-base bonded specimens has been found to be superior to that of nickel- Special tar-impregnated graphite crucibles showed better corrosion resist- ance in a fluoride mixture than in sodium. Reactor- baose bonded specimens. 62 grade {C-18) graphite exposed to sedium in a type 304 stainless steel container was unaltered, where- as o similarly exposed, tar-impregnated graphite was badly cracked and spalled. Thermaleconvection loops have been used for additional investigotions of corrosive attack of fluoride mixtures and liquid metals on Inconel, a special Inconel-type alloy, stainless steel, Hastel- loy B, and Inconel with stainless steel, nickel, or beryllium inserts. It has been shown that it is possible to greatly reduce the attack in Inconel loops when UF, is used with zirconium fluoride- base mixtures instead of the customary UF,, One difficulty with these mixtures is the formation of layers on the hot-leg surface which may be uronium- Inconel alloys. Fluorides with mixtures of UF, and UF , are being investigated, Mass fransfer of chromium was found to oceur in thermal-convection foops with hot-leg temperatures as low as 1200°F, The amount of mass transfer increases with an increase either in operating temperature or in temperature difference between the hot and cold legs. When graphed, the values for depth of attack obtained in a loop operated for 5000 hr fell on a straight line with the values obtained in shorter times. The previously presented data which showed @ rapid initial ottack with a slower secondary attack after 250 hr were confirmed. The increase in depth of attack is about 4 mils for each 1000 hr of exposure. Reductions in depth of attack were obtained with loops of Hastelloy B and a Incone! with a portion of the chromium In almost every loop special replaced by molybdenum, constructed with combinations of Inconel and type 316 stainless steel, evidence of increased attack on the steel was observed. A reduction was usually noted in the attack on the Inconel but mass transfer wos increased. Dissimilar-metal mass transfer was found with beryllium metal inserts in Inconel loops in which sodium was circulated at 1500°F. More mass transfer was found with a hot- leg insert, but it was present even with a cold-leg insert, Two lithium-filled type 316 stainless steel thermal«convection loops operated for 1000 he with hot-leg temperatures of 1445 and 1500°F, respec- tively. Only small omounts of mass transfer oc- curred, and ot no time was there any indication of plugging. These results are in contrast to results obtained with previous loops of this type which would only operate in the neighborhood of 200 to 300 hr prior to plugging. The increased life was probably due to the higher purity of the lithium used and, especially, to a decrease in the lithiom nitride content. ' In the study of the corrosion and mass-transfer characteristics of materials in contact with liquid lead, it was found that quartz thermal-convection loops containing types 410 and 446 stainless steel specimens had a much longer life prior to plugging than similar loops containing pure iron and chro- mium or one of the 300-series stainless steels. Two possible explanations of these findings are being investigated., A tenacious oxide film might act as o diffusion barrier between the container material and the moltern lead, or it may be that « thin film of sigma phase forms and acts as a dif- fusion barrier, Studies are continuing on the identification of the compounds produced by hydroxide.metal re- actions. Current research is concerned with the reaction of NaOM and nickel when hydrogen is allowed to escape from the system. In the past, color changes were observed when NaOH was heated to above 500°C, Present studies indicate that this is not due to a change in concentration of metallic iohs from the container but that it must be due to the existence in thermal equilibrium of some species other than the expected sodium ions and hydroxy! ions, Additives such as oxygen, hydrogen, and water vapor have been studied, The interactions of these materials on hydroxides do not seem to be greatly altered by changing the cation from sodium to cesium, but they do seem to be associated with the anion. FLUORIDE CORROSION OF INCONEL IN STATIC AND SEESAW TESTS H. J. Buttram R. E. Meadows J. M. Didlake F. Kertesz Materials Chemistry Division Effect of Chromium Addition to Fluoride Mel? An investigation was made of the effect of chromium additions to the fluoride mixtures NakF- - PERIOD ENDING JUNE 10, 1954 ZrF“-UF‘4 :(53.5-40.0-6.5 mele %) and NaF-ZrF‘ (53-47 mole %) on the tluoride corrosion of inconel. In a series of seesaw tests the chromium was added as coarse powder and as slugs of three sizes with surface-area ratios of 1:2:4 to determine the effect of varying the surface orea of the chro- mium, Inconel capsules which contained the fluo- ride mixtures with and without the chromium ad- ditions were tested. The capsules without chromium showed light attack to a depth of about 1 mil. A thin metallic loyer was formed on the walls of a capsule tested with chromium powder added to the uranium-bearing mixture; no layer was found on a capsule containing chromium powder and the non- vranium-bearing mixture. The weight losses of the solid chromium speci- mens which were added to the capsules were found to be independent of the size of the specimen used, but there was a marked difference between the weight losses caused by the two fluoride mixtures. The weight losses of the solid chromium speci- mens averaged 0.62 g in the uranium-bearing mix- ture and 0.11 g in the NaF-ZrF ; mixture. It was obvious that the difference between the weight losses in the two types of mixtures, 0.51 g in this test, did not correspond to the amount of reduction according to the equation UF, + C° —>CrF, + 2UF, . A simple calculation shows that the weight lost by the chromium specimen would hove been suf- ficient to reduce all the UF , contained in the 30-g charge. However, the chromium content gained by the fluoride mixture (about 2000 ppm) was much too small for the reduction of the UF ;. Also, an x-ray diffraction study of the solidified mixture revealed only the normal pattern of Na Zr(U)gF,, and none of the extraneous lines which would be expected if there were a high concentration of UF,, Petro- graphic examination showed just a tinge of the vellow, reduced phase. The protective action exerted by the chromium was undoubtedly due to the higher activity of the pure metal in comparison with that of the chromium in the Inconel. Since the chromium content of the melt after the test was much lower than the amount lost by the specimen, the chromium must have been deposited and alloyed with the wall, especially at the cold end of the capsule. ' Effect of Temperature The effect of temperature on corrosion of Inconel in fluoride ‘mixtures was examined previously to 63 ANP QUARTERLY PROGRESS REPORT establish the relationship between reaction temper- in the melt. These studies suggested that the chromium concen- tration of the melt increased to a maximum and subsequently decreased with increasing temper- ature, ature and chromium concentration To check the validity of these rather surprising tentative conclusions, itwo series of tests were conducted, In each test both NaZrF. and NaF- ZrF“-UFz4 (50-46-4 mole %) were used. Both series of tests were made under static, isothermal cone ditions for 100 hr; the test temperatures ranged from 600 to 1000°C. In one series of experiments the fluoride mixtures were contained in graphite capsules to which Inconel specimens were added. The second series of tests was made with the fluoride mixtures contained in Inconel capsules. Metallographic examination of the Inconel speci- mens from the graphite capsules showed no attack after exposure at 600°C, while after exposure af 700°C, light to moderate subsurface-void formation to a depth of about 2 mils was observed. The attack was more marked at 800°C, and some of the voids were associated with carbide precipitation on the alloy. The depth of attack increased ot ?00°C, whereas at 1000°C a slight reduction in void formation was observed. Metallographic ob- servation of the walls of the inconel capsules used in the second series of tests showed considerable attack at the lower temperatures; at the higher temperctures the subsurface-void formation was light and scattered, but some areas of very deep intergranular penetration were observed. It is possible that this general decrease in number of subsurface voids at the higher temperatures is due to a much higher rate of diffusion of chromium under these circumstances and a consequent slight depletion in chromium concentration over a large region rather than to void formation. The results of chemical analyses of the fluoride mixtures for chromium paralleled to some extent the metallographicy observations, The chromium content of the fluoride mixtures in the inconel capsules certainly did not increase in a regular fashion with temperature, as might be expected. While there appear to be slight tendencies toward maximums in the curves of chromium concentration vs temperature, it is likely that the differences observed are within the experimental errors of sampling and specimen analyses. For the samples tested in graphite capsules, however, the chromium 64 content of the fluoride mixture increased with temperature up to 900°C, and a straight line was obtained when the logarithm of chromium concen- tration was plotied against the reciprocal of absolute temperature. The chromium values for the mixtures tested at 1000°C appear to be somewhat lower than the values for those tested at 900°C, The appearance of the maximum values for the tests at 900°C, if it is real, is difficult to explain. The increase in chromium concentration with in- creasing temperature suggests that equilibrium is attained quite slowly at the lower temperatures and "that some stage in the corrosion mechanism has a considerable activation energy. The difference between the behavior of specimens in graphite capsules and those in Inconel capsules may well be due to the formation of a carbide film which reduces the rate at which chromium is available to react with the melt., No effects attributable to the presence of uranium in the fluoride mixture could be observed in these tests. Effect of Surface Area In static, isothermal tests in which fluoride mix- tures were contained in graphite capsules, two different sizes of Inconel specimens were used in identical quantities of fluoride mixture to establish the influence of ratio of corroding surface to fluo- ride volume on the quantity of chromium dissolved. Tests were conducted at temperatures from 600 to 1000°C at 100-deg intervals; the test time was 100 hr in each case, The surface areas of the two sizes of Inconel specimen were in the ratio 1.88:1. Except for the tests at 600°C, in which the chromium concentra- tions were very similar for the two specimen sizes, the larger specimen gave higher chromium concen« trations in the fluoride mixture. The ratio of chro- mium concentration in the fluoride mixtures ranged from o minimum value of 1.5 to a maximum of 2. 1. The essentially linear relationship between chro- mium surface area exposed cannot, at the chromium concentrations observed, concenfration and be ascribed to prior surface oxidation of the speci- mens. It seems likely that equilibrium concentra- tions are not aftained, possibly because of the formation of a carbide film on the specimens, and that the concentrations, which therefore are meas- urements of reaction rate, are contralled by area of surface exposed. Effect of Reduced Phases in Fluoride Melt The fuel containing trivalent uranium which gave low corrosion: values in thermal-convection-loop tests has been tested in seesaw experiments, The UF, was obtained by adding metallic uranium in the melt of NaF-ZrF,-UF, (53.5-40.0-6.5 mole %). After 100 hr of exposure in the seesaw apparatus, the unreduced control batch of the NaF-ZrF -UF, mixture showed intermittent, moderate, subsurface- void formation to a depth of 1 to 1.5 mils at the hot end of the capsule, whereas none of the batches containing UF, showed any meosurable attack. Chemical analyses confirmed the metallographic observations; the chromium concentration of the unreduced control batch was higher than that of the mixtures containing reduced uranium com- pounds. | Effect of Fission Products Preliminary ' experiments were run to determine the effect of fission products on the corrosive properties of NaF-ZrF ,«UF, on Inconel. The total amount of fission products per megawatt day of PERIOD ENDING JUNE 10, 1954 operation was taken as about: 1.1 g or about 6.6 g for four days of operation at 1.5 Mw, Most of the fission products would be present in amounts less than 10% of the uranium undergoing fission; so the maximum concentration of the product in highest abundance should be not more than about 1 ppm because the products will be diluted in about 1500 Ib of the fuel. Such low concentrations are not expected to cause difficulty from the corrosion standpoint. Since it would be difficult to handle such small quantities in the regular corrosion tests which require only a total of 30 g of fluoride mixture, the concentrations of the simuloted fission products were scaled up one-thousandfold. ' In order to mock up a fuel containing fission- product elements, a number of arbitrary substi- tutions had to be made because some of the desired elements are not availoble, Zirconium was not included as o fission-product element because the mixture already contained a large amount of that element, The additions thot were made to the NaF-Zer—UFd mixture are shown in Table 5.1, TABLE 5.1. MATERIALS ADDED TO SIMULATE FiSSION-PRODUCT ELEMENTS IN A ' NaF-ZrF («UF, MIXTURE QUANTITY ADDED T FISSION PRODUCT MATERIAL (meiq/kg) SIMULATED Added Before Hydrofluorination of Mixture BaF, 110 Ba CsF 40 Cs CeF, 425 Rare earths LaF, 110 Rare earths RbF 76 Rb YF 122 Y Added After Hfdrofluorination of M:ixfure* MoBr. 147 Br Te 70 Te RuF4 157 Pt group Nal 83 t *These materials were added after hydrofiucrination because they would have decomposed or volatilized during hydrofluorination. 65 ANP QUARTERLY PROGRESS REPORT Metallographic study of the Inconel-capsule walls after the 100-hr seesaw test showed heavy to very heavy subsurface-void formation 2 to 5 mils deep. In some cases the voids were so concentrated near the grain boundary that the attack appeared at first to be of an intergranular nature, Closer inspection revealed that the generally observed subsurface- void attack was present, ond the voids showed a predominant tendency to form near or at the grain boundaries. Chemical analyses of the fluoride mixtures showed very high chromium concentrations that ranged from 0.23 to 0.25 wt % ond thus con. firmed the severe attock observed metallographi- cally. Tests are under way to determine the effects of the simulated fission-product materials individually in as small amounts as possible. Preliminary re- sults show that after the usual 100-hr seesaw test, yttrium fluoride caused the greatest attack, with the moximum depth of subsurface-void formation being 3 mils, while molybdenum bromide coused the next greatest. Chemical results confirmed, in general, the metallographic observations. Tests are in progress to establish the exact amount of chromium removed from the Incone! by the various quantities of simulated fission products. STATIC AND SEESAW TESTS OF VARIOUS MATERIALS IN FLUORIDE MIXTURES AND LIQUID METALS E. E. Hoffman J. E. Pope W. H. Cook L. R. Trotter Metallurgy Division Dissimilar Metals in NuF-ZrF4wUF4 A series of tests has been conducted to determine the effect of having dissimilar metals (type 316 stainless steel and Inconel) in contact with the fluoride mixture NaF-ZrF -UF, (50-46-4 mole %) in a dynamic system with a temperature differential, Each tube was one-half type 316 stainless steel and one-half Inconel and the joins were made by heliarc welding. The hot zone in the seesaw apparatus used for these tests was at 1500°F and the cold zone was at 1256°F. Both the type 316 stainless steel and the Inconel sections were tested in the hot zone for 100 and 300 hr, The results of these tests are presented in Table 5.2, The hot and cold zones and the welds joining the two materials were examined after each fest. It may be concluded from the results of these tests 66 that type 316 stainless steel is inferior to Inconel in resistivity to attack by molten fluorides even when the steel occupies the cold zone. The heaviest attack found was to a depth of 6 mils in the type 316 stainless steel tube adjacent to a weld; the Incone!l tube adjacent to the opposite side of the weld was attacked to a depth of 1 mil {(Fig. 5.1). Molybdenum-Coated Inconel in NaF-ZrF ,-UF Inconel specimens which were coated with molybdenum by the Crawford Fitting Company were tested in static lithium at 1500°F for 100 hr, Lithium was chosen as the comrodent in order to determine whether the molybdenum coating was continuous, because molybdenum is resistant to attack by molten lithium at this temperature and Inconel is heavily attacked. Examination of the as-received sample revealed a ]/2-» to 3/,’4-mil molyb- denum-rich surface layer. The molybdenum-rich phase penetrated the grain boundaries to o depth of 'lTé mils beneath the surface layer. Examination of the tested specimen showed very heavy inter- granular attack to a depth of 35 mils; therefore the molybdenum surface layer was either not corrosion resistant or not continuous. Stainless Steel in Lithiom with Lithium Nitride Added Static tests were made to confirm the reported increase in corrosiveness of lithium metal when lithium nitride is added. Type 316 stainless steel tubes were exposed at 1600°F for 100 hr to lithium metal with 0.1, 0.25, and 1% of lithium nitride added. The test temperature of 1600°F was chosen because lithium nitride has a melting point of ap- proximately 1553°F., In all the tests the 35-mil walls of the type 316 stainless steel containers were completely penetroted intergranularly (Fig, 5.2). In standard tests under similar conditions, the maximum intergranular attack of type 316 stain- less steel by lithium (without lithium nitride) is S mils or less. Chromalloyed Steels in Liquid Metals Seesaw tests of chromalloyed stainless steels in ligquid sodium and in several liquid alloys were run to determine whether the chromium-rich layer improved the corrosion resistance of the base material, The surfaces of somples of type 304 stainless stee!l and type 1035 carbon steel were PERIOD ENDING JUNE 10, 1954 TABLE 5.2, RESULTS OF SEESAW TESTS OF DISSiMILAR METALS WELDED TOGETHER AND EXPOSED TO NaF»Z.r-F4-'UF4 (50+46.4 mole %) - EXPOSURE . MATERIAL TEMPEORATURE TIME DEPTH Oi ATT CK (FF) (he) {mils)} Inconel (hot zone) 1500 100 0.3 Type 316 stainless steel (cold zone) 1256 2 Inconel (hot zone) 1500 300 0.5 Type 316 stainless steel (cold:zone) “1256 1 Type 316 stainless steel (hot zone) 1500 100 2 Inconel (cold zone) 1254 G Type 316 stainiess steel (hot zone) 1500 300 ; 2.5 incone! (cold zone) ' 1256 0 Fig. 5.1. Attack Adjacent to Weld Joining Type 316 Siuinl@ss Steel and Inconel After Exposure to NnF-ZrF4-U‘F4 (50-46-4 mole %) in Seesaw Apparatus for 100 hr. During the test the temperature in the weld zone was approximately 1382°F. 500X. fi impregnated with chromium by the process which uses a reaction between a gaseous chromium com- pound and iron as a means of exchanging iron atoms for chromium atoms at the surface of the The parts to be treated were packed in o compound containing energizer, and an inert material, steel. powdered chromium, an Both the type 304 stainless steel and the type 1035 carbon steel after being chromalloyed were tested for 100 hr with hot- and cold-zone tempera- tures of 1500 and 1220°F in sodium, 38% Pb-62% Sn, 82% Pb-18% Cd, 32% Cd-68% Sn, 60% Bi-40% Cd, and 45% Pb-55% Bi. In all cases the chro- miumerich' layer was found to have cracked, and 67 ANP QUARTERLY PROGRESS REPORT | UNCLASSIFIED 1 Y- 10604 , e Fd vt -, Q.01 tNCH 002 | Q03 Fig. 5.2. Intergranular Penetration of Type 316 Stainless Steel Exposed ot 1600°F for 100 hr to Static ‘Lithjum with 0.1% Lithium Nitride Added. Top edge exposed to liquid. 100X, therefore the corrosion resistance of the base material was unimproved. Figure 5.3 shows photo- micrographs of a chromalloyed type 304 stainless steel specimen before and after testing in 45% Pb-55% Bi. Cermets in NGF-Z.!F4»UF4 Various cermets that are being considered as bearing materials were exposed to NaF-ZrF ,-UF , (50-46-4 mole %) in seesaw apparatus. The speci- mens were restricted to the hot zone of Inconel tubes and were exposed to the fluoride mixture for 100 hr. During the tests the hot zones of the tubes were at 1500°F, and the cold zones were at ap- proximately 1292°F, The results of the tests are presented in Table 5.3. Of the various cermets of titanium carbide with nickel binder, one specimen with 20% of binder and another with 30% of binder had the best resistance to attack. However, tests of duplicate specimens showed slightly heavier attack. This discrepancy may be due to variations in the specimens. Kennaometal D4675 (tungsten carbide plus 2.5% cobalt} showed very good resist- ance, and additional tests are planned to confirm this result. Since Corboloy X3505, which is the same as Carboloy 608 except that it has no nickel 68 binder, was not attacked, it appears that the nickel binder in Carboloy 608 (Cr,C, plus 2% WC and 15% Ni) was attacked preferentially {(Fig. 5.4), A second group of cermet samples submitted by Kennameta! lnc., and the Sintercast Corp. of America, was exposed under the some conditions, The exact compositions of one of the Kennametal samples and all the Sintercast samples are yet to be obtained from these firms. All these specimens had fair to good resistance to the fluoride mixture, as may be seen in Table 5.4, In general, the cor- rosion resistance of the cobalt-base bonded speci- mens was equal to or slightly superior to that of the nickel-base bonded specimens, In all cases the attack appedred o be along the titanium car- bide particle surfoces that separate the particles from each other and from the binder, The corrosion along the surfaces of the titanium carbide particles might have been the attack of a fabrication reaction product the titanium carbide and the binder. !t has been reported that a finely dispersed reaction product is usually found in the metallic areas of cermets containing cobalt or nickel.' between IC. C. McBride, H. M. Greenhouse, and T. 5. Shevlin, J. Am, Ceram,. Soc. 35, 29-30 (1952), PERIOD ENDING JUNE 10, 1954 Fig. 5.3. Chromalloyed Type 304 Stainless Steel Before and After Exposure to 45% Pb-55% Bi in Seesaw Apparatus. (a) As received. Specimen nickel plated prior o examination to prevent rounding of edge during polishing. (b) After exposure. 500X. ST A Fig. 5.4. Carboloy 608 (Cr,C, plus 2% WC and 15% Ni) Before and After Exposure to NaF-ZrF ,-UF (50-46-4 mole %) for 100 hr at 1500°F in Seesaw Apparatus. (o) As received. (b) Affer exposure. 250X. 69 ANP QUARTERLY PROGRESS REPORT TABLE 5.3. RESULTS OF SEESAW TESTS OF CERMETS EXPOSED TO NuF-Zer-U Fa (50-46-4 mole %) AT 1500°F FOR 100 hr MATERIAL COMPOSITION Kentanium K-150-A TiC + 10% Ni Kentanium K-151-A TiC + 20% Ni Kentanium K-151-A TiC + 20% Ni Kentanium K-152-B TiC + 30% Ni Kentanium K-152-B TiC + 30% Ni Attacked to a depth of 4 mils A zone of slightly affected material to a depth of 1 or 2 mils Macroexamination revealed many blisters on the surface of the sample; attack extended to a depth of 6 mils Attacked in only a few areas to a maximum depth of 2 mils; much less attack than on either K<150-A or K-151-A Attacked somewhat erratically to a depth of 4 to 7 mils; did not resist attack so well as the other sample of K-152-B Kentanium K-162-B TiC + 25% Ni and 5% Mo Kennametal D4675 WC +2.53% Co Firth Sterling 27 TiC +7% Cra(:2 and 50% Ni Carbolay 608 Cr3C2 +2% WC and 15% Ni Carboloy X3505 CryC, +2% WC {no nickel binder) Metamic L T-] (not heat treated) 77% Cr--23% A|203 Metamic LT-1 (heat treated) 77% Cr~23% Al203 Macroexamination revealed a surface which had the appearcnce of a plated specimen with portions of the plating removed; attacked to a depth of 9 mils Subsurface-void attack to a depth of 1 mil Attacked to a depth of 2 to 5 mils Attacked to a depth of 4 mils No attack detected; both as-received and tested specimens very brittle and similar in appearance Completely penectrated Completely penetrated These limited tests indicate that corrosion resist- ance is low for the specimens with the greatest binder content, whether nickel or cobait. Titanium boride, zirconium boride, and molybde- num boride samples were also exposed to NafF- ZrF ,-UF , (33.5-40.0-6.5 mole %) at 1500°F for 100 hr in seesaw apparatus, The specimens were restricted to the hot zone of Inconél tubes. The results of the tests are presented in Table 5.5, The evidence of attack on the zirconium boride specimen may be seen in Fig. 5.5. Graphite in NaF-ZrF UF, and in Sodium Special tar-impregnated graphite crucibles were exposed to NaF-ZrF4-UFd (50-46-4 mole %) and to sodium at 1500°F for 100 hr in static tests. As can be seen in Fig. 5.6, the sodium completely penetrated the walls of the crucible and caused them to crack and crumble. The crucible containing 70 the fluoride mixture had a weight loss of 0.012%, and the surfoce which was in contact with the fluoride had an etched appearance, The fluoride mixture did not penetrate the crucible walls, Static tests were also run fo compare the cor- rosion resistance in molten sodium at 1500°F for 100 hr of two types of graphite in type 304 stain- less steel containers. The types of graphite tested were C-18 (reactor-grade) and a special tar-impreg- nated graphite that was impregnated ond fired 16 times during its preparation. The purpose of the repeated impregnation and firing was to produce o high density and a tough skin that could possibly help to reduce penetration by various liquids., The C-18 graphite was practically vnaltered by the test (Fig. 5.7a). The surface of the specimen had an etched appearance and the sharp corners had been rounded slightly. The tar-impregnated cylinder was cracked badly and had spalled, as may be seen in Fig, 5.7b. PERIOD ENDING JUNE 10, 1954 TABLE 5.4. RESULTS OF SEESAW TESTS OF TITANIUM CARBIDE CERMETS EXPOSED TO ' MaF-ZrF («UF , (50-46-4 mole %) AT 1500°F FOR 100 hr MATERIAL COMPOSITION METALLOGRAPHIC NOTES Kentanium K-138-A Kentanium K-153-B Kentanium K-161-B Sintercast No. 1 Sintercast No, 4 Sintercast No. 3 Sintercast No, 8 Sintercast Ne. 10 65% TiC—20% Co-15% CbTiTaC, 60% TiC~40% Ni TiC + Ni and Mo TiC + Co TiC + Co TiC + Ni TiC + Ni TiC + Ni Attack appeared to be on binder to a depth of 0.5 to 1 mil Attacked to a uniform depth of 3 mils; attock was along the TiC particle surfaces that separate the TiC particles from each’ other and from the binder No attack; tested surface slightly more irregular thon as- received surfoce Attacked to a depth of 0.5 to 2 mils; many voids throughout specimen probably formed during fabrication Affucked to a depth of 1 mil; some TiC particles had a second phase within them No attack; binder appeared to be firmly attached to the TiC particles throughout the specimen Attacked to a depth of 2 mils; small section on corner of speci- men not uniform; appeared to be binder without TiC particles and did not appear to be attacked Attacked to a depih of 5 mils along the TiC particle surfaces; aftack just enough to separate particles from the binder GO0 INCH 0002 G003 Fig. 5.5. Zirconium Boride Before and After Exposure to NaF.ZrF -UF {53.5-40,0-6.5 mole %) for 100 hr of 1506°F in Seesaw Appuratuso {@) As received. (b) After exposure. 1000X. : 71 ANP QUARTERLY PROGRESS REPORT TABLE 5.5, RESULTS OF SEESAW TESTS OF VARIOUS BORIDES EXPOSED TO NaF«ZeF «UF , (53.5-40.0-6.5 mole %) AT 1500°F FOR 100 hr DEPTH OF ATTACK SAMPLE (mils) METALLOGRAPHIC NOTES TiB:2 2 As-received sample had what appeared to be a pressing crack; tested sample had a cracked corner ZrB, 2 Irregular shaped dark and light grains; as-received sample cracked and porous MO2B 16 Corrosion was definite and macrescopically visible in the polished metallographic sample; sample porous and very brittle UNCLASSIFIED Y-11097 ¥-12009 Fig. 5.7. C.18 Graphite and Tar-lmpregnated Graphite After Exposure to Static Sodium in Type 304 Stainless Steel Containers at 1500°F for 100 hr. (a) C-18 graphite. (b) Tar-impregnated graphite. FLUORIDE CORROSION OF INCONEL IN THERMAL-CONVECTION LOOPS G. M. Adamson Metallurgy Division Fig. 5.6. Special Tar-lmpregnated Graphite Effects of UF; and Mixtures of UF; and UF, Crucibles After Static Tests in Seodium ond in in Fluoride Fuels NuF.Z:F -UF, (50-46-4 mole %) at 1500°F for An extensive experimental program is under way 1C0 hr. ?a) Tested in sodium. (b) Tested in flueo- to determine the effect on corrosion of Inconel ride mixture, thermal-convection loops by circulating fluoride 72 fuels containing UF; and mixtures of UF; and UF . it has been determined that fluoride fuels con- taining uranium as UF3 are less corrosive fo Inconel than UF ;-bearing fuels. However, UF is not sufficiently soluble in the fluoride mixtures presently being considered for use in the Reflector- Moderated Reactor and mixtures of UF, and UF, are therefore being investigated, An effort is being made to determine the amount of UF; required to ‘eliminate corrosive attack by a fluoride mixture containing UF ; and UF ;. A few experiments have been completed with zirconium-base fluoride mix- tures ond tests with alkali-metal fluoride mixtures are planned. : The fluoride mixture NaF-ZrF ,-UF, was circu- lated for 500 hr in a standard Inconel thermal- convection loop with a hot-leg temperature of 1500°F (AT = 195°F). The uranium concentration was low, being only 2,8 wt %.: Examination of the loop after 500 hr of operation showed no subsurface voids or intergranuvlar type of attack. The hoi-leg PERIOD ENDING JUNE 10, 1954 surface was partially covered by a thin, unidentified layer, but no change in the uranium analysis was found. A mixture with a higher, but still low, uranivm concentration was circulated in a second loop. The uranium concentration was 5.8 wt % and to obtain this high a concentration, a mixture of UF, and UF ; was required, Estimates and cnalyses of the relative amounts of the two compounds have varied from 100% UF, to 100% UF,. While a definite figure cannot be obtained, a reasonable value for the UF, and three-fourths of the total vranium. This loop also showed no aottack after 500 hr of operation with o hot-leg temperature of 1500°F. Layers were present in both the hot and the cold legs, but the layer in the hot leg was thicker than the layer in the cold feg (Fig. 5.8). . The average uranium concentration of the fluoride mixture after circu- lation was 5.3 wt %, with none of the values as high as the value reported for the mixture before seems to lie between one-half Fig. 5.8. Layer on Surfoce of Hot Leg of an Inconel Thermal-Convection L00p After Circulation for 500 hr ot 1500°F of a NaF-ZrF , Mixture Containing UF and UF . 1400X. 73 ANF QUARTERLY PROGRESS REPORT circulation. It is possible that the deposited ma- terial is a uranium-Inconel alloy formed by the dissociation of the UF, from the fluoride mixture. This is the main question that must be answered before a decision can be made as to whether these fluoride mixtures will be useful as reactor fuels circulating in Inconel, An intermediate uranium mixture of 2.7 wt % UF, and 8.6 wt % UF, or o total uranium concentration of 8.7 wt % was circulated in another loop. These proportions were obtained by calculations, but the analytically determined uranium concentration of 8.63 wt % makes the calculated value appear to be reasonable. After 500 hr of circulation, a moderate concentration of subsurface voids with a maximum penetration of 4 mils was found. No layer was found in either the hot or the cold leg. The final analytical values have not been received. Effect of Hydrogen Fluoride Small additions of hydrogen fluoride gas were made to a batch of NaF-ZrfF,-UF (50-46-4 mole %). Portions of this batch were then circulated for varicus times at 1500°F in Inconel thermal-con- vection loops in order to determine the effect of hydrogen fluoride on the corrosiveness of the fuel mixture, The data obtained are summarized in Table 5.4. The doubling of the depth of penetration in loop 415 in comparison with that in loop 421 confirms the adverse effect of small amounts of hydrogen fluoride on the initial rate of attack. In comparison with the maximum penetration of 17 mils found in loop 417 after 2000 hr of operation, an estimate based on results obtained previously indicates that the maximum penetration in a loop operated for 2000 hr with the as-received fluoride mixture would be 12 mils. The effect of the hydrogen fluoride addition on maximum penetration oppears to be about the same after 2000 hr of operation as after 500 hr of operation, EHect of Temperature The effects of temperature drop and of maximum hot-leg temperature on mass transfer in Inconel thermal-convection loops are being studied. The use of thermal-convection loops in the study of the effects of temperature gradients is not entirely satisfactory because the flow in a loop is a function of the density of the fluid ond therefore the flow rate changes with changes in the temperature drop. The thermal-convection-loop resulis are of some value, however, because the flow rates are so low that the changes in flow rate ore not so important as the direct changes in temperature drop. Two series of loops in which the cold-leg temper- ature was varied while the hot-leg temperature was held at 1500°F were operated with NaF-ZrF -UF , (50.46-4 mole %) as the circulated fluid, The variation in the cold-leg temperature was accom- plished by adding insulation or by blowing air on the cold leg. The data from these tests are pre- sented in Table 5.7, of attack is caused by impurities and therefore The first, more rapid stage TABLE 5.6, EFFECT OF HYDROGEN FLUORIDE ADDITIONS ON CORROSIVENESS OF NaF-ZrF -UF, IN INCONEL THERMAL-CONVECTION LOGPS CORIGINAL FLUORIDE- 7 LOOP HYDROGEN FLLUORIDE MIXTURE IMPURITY |CIRCULATION MAXIMUM NO CONCENTRATION {ppm) TIME PENETRATION HOT-LEG ATTACK " | (moles/liter of purging gas) (hr} {mils) Ni Fe 421 1.2 x 105+ <10 110 500 5 intergranular, moderate to heavy 415 3.4 x 108 300 10 Intergranular, moderate to heavy 416 6 x 108 <10 345 1000 12 Heavy intergranular 417 6 x 10° <10 195 2000 17 Intergranular, moderate to heavy with large voids *Normal concentration of hydrogen fluaride in as-received NqF-Zer-UFd mixture, 74 would not be expected to be greatly influenced by the temperature drop. The differences found appear to take place in the mass-transfer portion of the attack which usually occurs after the first 250 to 500 hr of operation. Although no definite con- clusions should be drawn from such meager data, the rate of mass transfer seems to vary quite markedly with changes in temperature drop. A series of loops filled from one batch of NoF- ZrF ,-UF , was operated for 2000 hr in an effort to determine the effect on rate of mass transfer of PERIOD ENDING JUNE 10, 1954 data obtained are presented in Table 5.8. The increases .in depth of attack and incidence of cold- leg deposits indicate that the rate of mass transfer increases with increasing hot-leg temperatures, L.oops operated previously for 500 hr with maxi- mum hot-leg temperatures of 1200°F showed maxi- mum penetrations of 3 to 5 mils, and in a loop operated for 500 hr with a maximum hot-leg temper- ature of 1250%F the maximum penetration was 3 mils. The previously operated loops were not filled from the batch of fluoride mixture used for filling variations in maximum hot-leg temperatures, The the loops listed in Table 5.8, but the small vari- TABLE 5.7, EFFECT OF TEMPERATURE GRADIENT ON MASS TRANSFER IN INCONEL : THERMAL-CONYECTION LOOPS CIRCULATING NaF-ZrF ,-UF, L ooP MAXIMUM OPERATING MAXIMUM , . NO Ar TIME PENETRATION HOT-LEG ATTACK : (°F) (hr) {mils) 447 65 1000 5 Intergranular, light to mederate 443 195* 1000 10 General and intergranular, moderate to heavy; the deepest penetrations were scattered 446 220 1000 12 Infergranular, moderate to heavy 422 130 500 5 Intergranular, moderate to heavy 423 140 2000 6 Intergranular, moderate to heavy 442 195* 500 10 Intergranular, moderate to heavy *Standard temperature drop. TABLE 5.8. EFFECT OF VARIATIONS IN MAXIMUM HOT-LEG TEMPERATURE OF THERMAL-CONVECTION LOOPS OPERATED FOR 2000 hr HOT-LEG MAXIMUM : ngp TEMPERATURE | PENETRATION HOT-LEG CONDITION COLD-LEG CONDITION ’ (“F) (mils) 391* 1200 7 "Very heavy general attack with No deposit small veoids : 449 1250 6 Intergranular, moderate to No deposit heavy 392 1400 8 ‘Heavy general attack to 4 mils Intermitfent " with oceasional deeper metallic deposit intergranular penetratiens 393 1600 13 Antergronular, moderate to Thin but continuous heavy with large voids metallic deposit *This run wos terminated after 1313 hr when circulation was stopped for the second time by a power failure. 75 ANP QUARTERLY PROGRESS REPORT ations encountered in comparing different batches are not sufficient to invalidate the comparison between the previously operated loops and loops 391 and 449. The difference between the maximum penetration in loop 391 and the maximum penetration in the loops operated previously indicates that mass transfer takes place even at a hot-leg temper- ature as low as 1200°F, Effect of Exposure Time A maximum penetration of 27 mils was found in loop 344 which circulated NaF-ZrF -UF , (50-46-4 mole %) for 5000 hr at 1500°F, The fluoride mixture in this loop was from the same batch as that used for a series of tests reperted previously. When graphed, the values for the depth of attack in 5000 hr fell surprisingly close to the straight line for the second stage of attack (after 250 hr), as shown in Fig. 6.6 of the previous report,? The average increase in depth of attack is about 4 mils for every 1000 hr of operation ofter the first 250 hr. A layer about !{‘-in. thick of dendritic chromium metal crystals was found in the trap at the bottom of the cold leg of the loop operated for 5000 hr (Fig. 5.9). Another series of experiments is under way to confirm the time curve and to provide samples from which a reaction rate may be obtained by deter- mining the volume of the subsurface voids. The data obtained for the time study (tabulated in Table 5.9) confirm the data obtained previously,? including the change of slope of the curve af around 250 hr, However, the depth of attack found in the loop that operated for 1000 hr is out of line both with the other data obtained in this series of experiments and with the data from previous experi- ments., 2G, M. Adamson, ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL-1609, p 79, esp. Fig. 6.6. Fig. 5.9. Dendritic Chromium-Meta! Crystals in Trap at Bottom of Cold Leg of an Inconel Therma!. Convection Loop That Circulated NoF-ZrF -UF, (50-46-4 mole %) for 500 hr at o Hot-Leg Tempera- ture of 1500°F. 250X. Reduced 19.5%. TABLE 5.9. EFFECT OF OPERATING TIME ON DEPTH OF ATTACK BY NaF-ZrFA-UF4 CIRCULATING IN INCONEL THERMAL-CONVECTION LOOPS AT A HOT-LEG TEMPERATURE OF 1500°F LoOP OPERATING MAXIMUM NO. TIME PENETRATION HOT-LEG ATTACK (hr) {mils) 431 10 2 Light with intermittent moderate areas 432 50 3 Moderate 433 100 4.5 Light to moderate 434 250 8 Moderate to heavy, primarily intergranular 435 500 8 Heavy, primarily intergranular 436 1000 14 Heavy, intergranular 445 1500 13 J Heavy, intergranular 76 THERMAL-CONVECTION-LOOP TESTS OF VARIOUS MATERIALS i‘lm:l"'-iz.fF“-Ul‘:4 in Special inconel G. M. Adamson Metallurgy Division A special series of alloys with slight changes from the usual Inconel analys:s were vacuum cast af ORNL ond drawn into /-m. tubing with a 0.045- in, woll by the Superior Tube Co. The first lot of tubing was made into a loop which circulated NaF- ZrF4-UF4 (50-46-4 mole %) for 500 hr at a hot.leg temperature of 1500°F, A portion of the chromium normally found in Inconel had been replaced in this special alloy by molybdenum. An analysis of the special alloy showed 6.0% Cr, 10.0% Fe, 74.4% Ni, and 9.8% Mo. Examination of the tubing after 500 hr showed light attack with a maximum pene. tration of 3.5 mils, Standard Inconel loops of commercial ,‘/f‘,-in. tubing developed moderate to heavy attack 4 to 5 mils deep under the same conditions.® Although the attack on the special alloy was somewhat less than the attack on com- mercial Inconel, the special alloy was not as cor- rosion resistant as had been expected, Additional loops are to be operated for Jonger periods to study the effects of mass transfer. NaF-ZrF -UF, in Hdsfelloy B G. M. Adamson Metallurgy Division A loop of lé-in. Hastelloy B tubing was fabri. cated by seam welding and then solution annealed for 1 hr at 2150°F and aged 30 hr at 1950°F, The fluoride mixture NaF-ZrF ,-UF, (50-46-4 mole %) was then circulated in this loop for 500 hr at a hot-leg temperature of 1500°F, Examination of the tubing after 500 hr showed that the inner surface of the hot leg was rough and had light subsurface voids to a depth of 2 mils. The weld bead at the seam showed much less aftack than the tubing (Fig. 5.10). The cold leg showed similar attack except that there were fewer subsurface voids. NuFfsz“-UF“ in Stainless Steel G. M. Adamson Metallurgy Division A second type 430 stainless steel loop? was operated with. NaF-ZrF UF, (50-46-4 mole %) for 500 hr at a hot-leg temperature of 1500°F. The PERIOD ENDING JUNE 10, 1954 hot leg developed very light, widely scaitered subsurface voids to o maximum depth of 7 mils. A phase change of the stainless steel to what appeared to be martensite was noted in the vicinity of the voids. [n an area up to 10 mils deep from the surfoce, the carbides had apparently gone back into solution and grain growth had occurred: (Fig. 5.11). Particles of metal had deposited on the surface of the cold leg, and metallic crystals were found in the odhering fluorides, While the attack was less than that found in Inconel loops, more metallic crystals were visible, The attack was also slightly greater than that found in the first type 430 stainless steel loop.* NGF-ZIF4-UF4 in Inconel with Stainless Steel or Nickel Inserts G. M. Adamson Metallurgy Division Mass transfer of dissimilar metals was investi- goted in a series of loops with various combinations of Inconel and type 316 stainless steel, The fluo- ride mixture NqF-Zer-UF‘t. (50-46-4 mole %) was circulated in the loops for 500 hr ot a hof-ieg temperature of 1500°F, Inconel loop 315 had a 6-m. insert of type 316 stainless steel in the upper portion of the hot leg.? The lnconel above the insert showed very light, widely scattered attack to a depth of 3 mils.: The surface of the type 316 stainless steel insert was very rough and there was attack to o depth of 12 mils, There was a 0.3-mil-thick deposit on 'rhe Inconel cold leg. In inconel loop 430, the lower half of the hot leg was replaced with type 316 stainiess steel. The Inconel showed moderate attack to o depth of 8 mils, At the welds the Inconel was rough but showed no other evidence of ottack. The stainless steel had severe intergronular voids to a depth of 4 mils. There was no deposit and no attack in the cold leg. Loop 429 was constructed with a hot leg of type 316 stainless steel and o cold leg of Inconel. The surface of the hot leg was very rough and showed severe intergranular penetrations to a depth of 4 35, M. Adamson, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 71. 4G. M. Adamson, ANP Quar. Prog. Rep. Dec. 10, 1953, ORNL-1649, p 74. . 56, M. Adamson, ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL-1609, p 77. 77 ANP QUARTERLY PROGRESS REPORT Fig. 5.10, Hot-Leg Surface and Weld Bead of o Hastelloy B Loop After Circulation of NaF.-ZeF -UF, (50-46-4 mole %) for 500 hr ot o Hot-Leg Temperature of 1500°F. 250X, mils. At the welds the stainless steel was heavily attacked to a depth of 3 mils. At the lower weld the Incone! was unattacked, but at the upper weld there were many subsurface voids to a depth of 3 mils in the Inconel. A scattered metallic deposit was found on the Inconel cold leg. In loop 428, the cold leg was type 316 stainless steel and the hot leg was Inconel. The Inconel hot leg showed heavy intergranular attack to a depth of 10 mils. rough and there was a light concentration of sub- surface voids to a depth of 0.5 mil. The stainless steel at the welds showed severe but intermittent attack to a depth of 2 mils. The surface of the stainless steel cold leg was very rough with heavy intergranular aftack to a depth of 2.5 mils. An all- Inconel loop (435) filled from the same batch of fluorides showed heavy hot-leg attack to a depth of 8 mils. At the welds the Inconel surfoce was 78 pat Wl s | e e T T a : o % » i L . s - _ // o . - ’ 4 ’ 'j//‘\»"‘“'- “ * " = a 5 . S P e i A t ; * 1 < - = 4 . ) ) .. : . S ~ “ . - - Lo - . " P a : - . ¢ N B - T - e ks b ¢ .. : s S g s oL LSS e R A g - s [ " ’ ™~ . P - - Cooh—ed - a, 1o N T e T s - Fig. 5.11. Hot-Leg Surface of Type 430 Stain- less Steel Loop After Circulation of NaF.ZrF - UF4 (50-46-4 mole %) for 500 hr at o Hoi-Leg Temperature of 1500°F. 250X. Reduced 35.5%. in every loop tested, there was much deeper attack of the type 316 stainless steel insert;espe-~ cially when it was in contact with Inconel in the hotter portion of the loop, than there would have been in an all-steel loop. The effect of the stain- less steel on the Inconel is not so definite, but in most cases there seemed to be a reduction in the attack on the:Inconel, Since the data showed rather large variations in wall thickness, the thicknesses are being rechecked, but it seems that considerable reductions in wall thickness occurred in all the type 316 stainless steel tubing. A nickel insert was also placed in the cold leg of an Inconel loop. With this combination neither material seemed to be affected. This experiment confirms a previous experiment® in which a nicke! hot-leg insert in an inconel loop was also shown to have practically no effect, Sodium in Inconel with Beryllium Inserts G. M. Adamson Metallurgy Division Two Incone! loops were constructed with cylin- drical hotspressed beryllium inserts about 6 in. long. Both loops circulated sodium for 500 hr at a hot-leg temperature of 1500°F. In one loop the insert was in the upper portion of the hot leg. The outer surface of the insert, which was separated from the Inconel by a 0,020-in, annulus filled with slow-moving sodium, showed subsurface-void at- tack to o maximum depth of 14 mils. The inner surface of the insert did not have the void type of attack, but it was rough, and some even removal had probably occurred.. The Inconel adjacent to the beryllium was rough, and there was a metallic- appearing deposit on the surface. A black scale was present in all sections of this loop, and metallic crystals were found both in the annulus and in the cold leg. These crystals were predomi- nantly nickel and beryllium. The black scale produced a very complicated diffraction pattern which could not be identified. In the other loop the beryllium insert was in the fower cold leg. No attack was found in the Inconel hot leg of this loop. A thin metallic deposit was visible on the surface of the Inconel below the insert but not on the surface above it. The insert showed some attack on both faces as surface 6(5. M. Adamson, ANP Quar. Prog. Rep., Mar. 10, 7954, ORNL-1692, p 73. PERIOD ENDING JUNE 10, 1954 roughness, intergranular penetrations, ond voids to a maximum depth of 2 mils. Again, some even removal had probably occurred. No metallic crys- tals were found in this loop, but the same black scale was found in all sections. Three additional loops are now operating with similar hot-leg in- serts, These loops are cnrculahng sodium at 900, 1100, and 1300°F. Lithium in Stainless Steel E. E. Hoffman W. H. Cock J. E. Pope L. R. Trotter Metallurgy Division Two lithium-filled type 316 stainless steel ther- mal-convection loops completed 1000 hr of oper- ation. One loop operated with a hot-leg temper~ ature of 1500°F and a cold-leg temperature of 1220°F. The other loop operated with a hot-leg temperature of 1445°F and a cold-leg temperature of 1337°F. At no time during the tests did the loops give any indication of plugging. There was very little difference in the appearance of the hot leg and the coid leg in either loop. There was no macroscopic evidence of mass-transferred crystals except at the bath-level line below the fill pipe where there was a fine ring of crystals in both loops. Metallographic examination of these Ioops is not yet complete. ' FUNDAMENTAL CORROSION RESEARCH G. P. Smith Metallurgy Division Mass Transfer in Liquid Lead J. V. Cathcart Metallurgy Division The investigation of corrosion and mass transfer in liquid lead has revealed that small quortz thermal-convection loops containing types 410 and 446 stoinless steel test specimens require from two to three times longer to plug than similar loops confaining pure iron and chromium or one of the 300-series stainless steels.” An effort has been made to find an explanation for this phenomenon. As in previous experiments, all tests were con- 7ANP Quar. Prog. Rep. Sept. 10, 1953, DRNL.1609, p 80; ANP Quar. Prog. Rep. Mar. 10, 1953, ORNL-1515, p 128; ANP Quar. Prog. Rep. June 10, 7953, ORNL- 1556, p 64. 79 ANP QUARTERLY PROGRESS REPORT ducted in small quartz thermal-convection loops.® Two possible explanations for the increased resistance to mass transfer of the 400-series stain- less steels in liquid lead have been considered. First, the influence of the very thin oxide films on the specimens was investigated. It was thought possible that the oxide formed on the 400-series stainless steels possessed a structure which made it especially resistant to ottack by liquid lead. In such an event a thin oxide film would present a barrier to the solution of material from the speci- men walls and thus reduce the rate of mass trans- fer. It has been demonstrated’ that relatively thick oxide films (approximately 1000 1&) greatly reduced mass transfer in loops containing types 304 and 347 stainless steel specimens. An attempt was olso made to remove completely the oxide film from type 446 stainless steel speci- mens mounted in a loop. A small quartz tube was sealed into the loop ot the bottom of the bend in the cold leg and, before the loop was loaded, the specimens were heated to 950°C while hydrogen, which had been dried over magnesium perchlorate, was allowed to flow through the loop. After 1 hr of this treatment the side arm was sealed off with a hand torch, and the loop was immediately loaded with liquid lead. The loop plugged ofter 768 hr of operation with hot and cold-leg temperatures of 805 and 525°C, The long time required for the plugging of this loop was taken as an indication that the presence of an oxide film on the specimens was not the correct explanation for the relatively high resistance to mass transfer in liquid lead exhibited by type 446 stainless steel. Of course, it is possible that the oxide on the test specimens was only partially reduced or that, even if it were completely reduced, encugh oxide remained in the lead to reoxidize the surface of the test specimens. In such an event, the results of this test would be inconclusive; however, it is believed that the corrosion characteristics of type 446 stainless steel cannot be explained on the basis of the influence of thin oxide films on the surface of the test specimens, The second opproach made to the problem in- volved the assumption that a thin layer of sigma phase formed near the surface of the types 446 and 8ANP Quar. Prog. Rep. Dec. 10, 1952, ORNL-1439, p 148. 9ANP Quar, Prog. Rep. Dec. 10, 1953, ORNL.-1649, p74. 80 410 stainless steel hot-leg specimens and served as a barrier to solution of the specimens. The presence of such a layer has not yet been revealed either by metallographic exomination of the test specimens or by an x-ray study of filings taken from the surface of a type 446 stainless steel hot- leg specimen, If this explanation is correct, hows- ever, only a very thin layer of sigma phase would be required to produce the increased resistance to mass transfer and it may not be surprising that such o layer has not been found. In order to test the effect of sigma phase on resistance o mass transfer in iron-chromiumenickel alloys, a special alloy having a composition of 16% Ni—~37% Cr—47% Fe was prepared. This alloy was chosen because its composition corresponds to that in the sigma-plus-austenite region of the nickel-chromium-iron phase diagram. Two sets of specimens were prepared; one set was annealed for 80 hr at 810°C to facilitate the precipitation of the sigma phase; the other set was annealed at 1200°C and water-quenched to leave only ferrite and austenite phases in the alioy. Two loops containing sigma-phase (sigma and austenite} specimens of this alloy were operated. The first loop plugged ofter 456 hr of operation with hot- and cold-leg temperatures of 820 and 510°C. The second loop plugged after only 306 hr with hote and cold-leg temperatures ot 805 and 550°C. Metallographic examination of the second loop has not yet been completed, but a transverse section of the attacked region on the hot-leg speci-~ men from the first loop is shown in Fig. 5.12. As may be seen, the attack was very irregular; thin filaments of the original metal extend into o lead Close examination revealed that each filament consisted of a core of sigma-phase ma- terial surrounded by a thin film of matrix metal, The sigma-phase regions of the test specimens appeared to have suffered much less attack than the austenitic regions. matrix, The nonsigma-phase (ferrite and austenite) speci- mens of the alloy were also tested in two loops. The first loop plugged after 287 hr of operation with hot- and cold-leg temperatures of 810 and 520°C. The second loop plugged after 101 hr with hot- and cold-leg temperatures of 810 and 510°C, A transverse section of the hot-leg specimen from the first loop is shown in Fig., 5,13, The lack of repreducibility in the plugging times for both the sigma- and nonsigma-phase loops can PERIOD ENDING JUNE 10, 1954 0,004 Fig. 5.12. Transverse Section of a Special Sigma-Phase Aflpy (16% Ni-37% Cr-47% Fe) Specimen from the Hot Leg of a Quartz Thermal-Convection Loop Which Circulated Liquid Lead. 750X, probably be accounted for in terms of differences in the degree of transformation which occurred during the heat treatments of the specimens., The hot-leg temperature during the operation of the loops was very close to the sigma-to-ferrite trans- formation temperature; thus it is not unlikely that an appreciable amount of sigma phase was either formed or destroyed during the operation of the loops and therefore produced the variation in plugging times which was observed. The results obtained with the 16% Ni-37% Cr-47% Fe alloy indicate a definite increase in resistance to mass transfer for sigma-phase alloys in comparison with the nonsigma-phase alloys. However, no definite conclusions can be drawn as to the role of the sigma phase in the resistance to mass transfer exhibited by the 400-series stainless steels until a sigma-phase layer is actually ob- served in the test specimens from a loop containing one of these steels, | ............................................................................................................................................................................................................. Products of Hydroxide-Metal Reactions H. L. Yakel, Jr. G. P. Smith ‘ Metallurgy Division Studies are being made to identify the compounds produced by hydroxide-metal reaction and to deter- mine their properties. Previous studies were con- cerned with the action of lithium and sodium hydroxides on nickel in the presence of oxidizing agents.'® Current research is concerned with the reaction which occurs between sodium hydroxide and nickel when hydrogen is allowed to escape from the system, _ fi Miller and Williams'' have shown the existence 19 D. Dyer, B. S. Borie, Jr., and G. P. Smith, Alkali- Metal Nickel Oxides Containing Trivalent Nickel, ORNL- 1667 (Feb. 26, 1954). ”Reported by D. D. Williams and C. T. Ewing in Sixth Progress Report on Thermal and Related Physical Properties of Molten Materials, Naval Research L.ubora- tory Problem No. 32C11.06 (1953) and preceding reports in this series. 81 ANP QUARTERLY PROGRESS REPORT Fig. 5,13. Tronsverse Section of a Special Nonsigma-Phase Alloy (16% from the Hot Leg of a Quartz Thermal-Convection Loop Which Circulated Liquid Lead. 750X. of an equilibrium of the type Ni + NaOH = H, + un- known products. They reported that when this equilibrium was displaced by evacuating at 900 to 1000°C until hydrogen evolution ceased, an “amorphous’’ product was obtained which had a gross composition corresponding to 1 mole of sodium oxide and 1 mole of nickelous oxide. Woltersdorf!? found that mixtures of sodium oxide and nickelous oxide react at temperatures as low as 250°C to produce various crystalline sodium nickelate(ll) compounds. He was able to establish that sodium orthonickelate(l!) hos the empirical formula Na,NiO,. Kruh'3 found that a mixture of NoOH and nickel heated to 700°C under flowing argon gave x-ray powder patterns which did not 126, Woltersdorf, Z. anorg. Chem. 252, 126 (1943). 13R’eporfed by R. F. Kruh, Report for the Period April 1, 1953 through June 30, 1953, p 5, University of Arkansas, Institute of Science and Technology, Fayette- ville. 82 UNCLASSIFIED & Y-12420 0.0G4 0.002 I 0 Z 0.003 0004 10,005 Ni—37% Cr.-47% Fe) Specimen correspond to any patterns in the literature. Kertesz and Knox'# induced a reaction between sodium hydroxide and nickel by slowly removing a relatively small amount of hiydrogen gas. X-ray powder patterns of the resulting material did not show any identifiable lines. During the past few months the reaction which takes place when hydrogen is rapidly evacuated from the NaOH-Ni system has been studied. The primary product of this reaction is a crystalline sodium divalent nickelate of an as yet undetermined formula. In these studies the reaction chamber was a 1.27-cm-00D, 44-cm-long nickel tube (low-carbon grade), The tube was welded closed at the bottom end ond fastened at the top end to the rest of the all-glass apparatus by means of a standard, ground ball joint. The nickel tube could be evacuated g, Kertesz, private communication. through either of two circuits by turning a stopcock. One circuit was relatively short and allowed fast pumping speeds. The other circuit contained a ltquid-nitrogen cold trap. The pressure of the system was measured by means of a Pirani gage and an Octoil-S manometer similar to that described by Biondi.'®> This combination of instruments provided pressure measurements from 1 to 1000 p with a precision of 1/5 u at the low-pressure extreme and 20 p at the high-pressure extreme. The vacuum was produced by a Kinney CVM 3534 mechanical pump. Dry carbon-dioxide-free argon could be admitted to the apparatus when desired. The nicke! reaction tube was heated by a tube furnace. Temperatures were measured with chromei-alumel thermocouples which were wired to the nickel tube. The method used for opening the reaction tubes at the end of an experiment did not make it practical to weld the thermocouples to the tube. Hence, the highest temperatures reported (950°C) may have been in error by as much as +20°C, The nickel reaction tube was loaded in air with about 10 g of cp-grade sodium hydroxide and then quickly attached to the apparatus. Dehydration of 150, A. Biondi, Rev. Sci. Instr. 24, 989 (1953). PERIOD ENDING JUNE 10, 1954 the hydroxide was accomplished by slowly heating it to about 400°C under a vacuum. The circuit containing the cold trap was used to remove the water evolved, The course of the dehydration was followed by observing pressure changes, and some observations were made that are described in the following section. | f After the dehydration step the temperature was increased to a predetermined value while the gases evolved were pumped off through the short circuit. After reaction had proceeded as far as desired, the furnace was rapidly cooled, Dry carbon-dioxide- free argon was then admitted to the apparatus, and the reaction tube and its contents were removed to a dry box for examination. Reactions were carried out over the temperature range of 770 to 950°C, and. the reaction products were examined., At 770°C the pressure (assumed to be due to hydrogen) rose to a maximum of 200 and slowly decreased to 70 p over a period of about 7 hr. The course of another reaction is shown in Fig. 5.14, It had been intended that this reaction would be carried out at 950°C, However, as may be seen from Fig.5.14, the pressure passed through its maximum before 950°C was reached. The shaded area connects the extremes of pressure flucty- ations, | | UNCLASSIFIED ORNL-LR—~DWG 1346 1000 T - 1000 T T ITT 1 : . ’ TEMPERATURE i ! 800 e e — ] 80O ’:,"‘ P £ & 5 600 b e e e et 4 800 £ L & £ £ Z 3 a ) qa, D400 o X e e e 400 = x : = . PRESSURE 200 b dopln MTm"_m— ----------------- el 200 0 L 0 D 10 20 30 40 50 60 70 g0 a0 100 TIME (min) Fig. 5.14. Gas Evolution in the Reaction of Nickel with Sodium Hydroxide in a Continuously Evacu- ated System, R AT e R A LR S 8 4 8y S AR e R R LS 4S8 e At e ................................................................................................... 83 ANP QUARTERLY PROGRESS REPORT In Fig. 5.14, an area A taken under the curve for any time interval Atz is proportional to n, the number of moles of gas evelved during the time interval Af, The pumping speed s is defined to be s =dV/dt, where dV is the increment of gas velume passing through a given cross section of tubing in the time interval 4z, By applying the ideal gas law in its usuval form it is seen that sP dn = —— dt RT If s is constant over the pressure range under con- sideration, s Ly sA n o= P dt = | RT RT and therefore n ~ A For the pressure range represented in Fig. 5,14, the pumping speed varies by about 20% for the Kinney CVM 3534 pump. Application of this analysis to Fig. 5.14 shows that the reaction NeOH + Ni = nickelate + H, cannot be said to cease at any definite time under the experimental conditions. The effect which the accumulation of the reaction product has on the diffusion of sodium hydroxide to the nickel wall is undoubtedly the reason for this. The nickel tubes containing the reaction products were opened in a helium-filled dry box., Microscopic examination showed that the reaction products consisted of fibers and a powder, which were subsequently separated. The fibers, when very thin, were transparent green, They showed no recognizable crystal faces and were frequently bent or portially split along the fiber axis. They dissclved readily in T N hydrochloric acid, had an oxidizing power of less than 0.88 meq/g (probably due to cobalt and man- ganese impurity), and reacted with moisture and carbon dioxide in the air to form a crust of sedium carbonate monohydrate. The chemical data sug- gest that the compound is a sedium nickelate(ll). A single fiber was sealed in a glass capillary in the dry box immediately after the nickel reaction tube was open. This fiber, which was found to be a single crystal, was used for preliminary structure determination work. Rotation, Weissen- berg, and precession photographs of the single- crystal fiber show orthorhombic symmetry (Laue 84 symmetry mmm). Unit cell dimensions were meas- vred from all three types of photograph, and the suitably averaged results are: 4 - 828,A 100114, b = 10.140 A + 0,008 A , o c = 281, A + 0.008A , V - 236.67 x 107 cc . The extinctions observed in the single-crystal diffraction pottern are consistent with the three space groups Amam, Amae2, and Amo2,, Piezo- electric or pyroelectric experiments might distin- guish between the centrosymmetric space group Amam ond the other noncentrosymmetric possibili- + It ties, but these experiments have not been performed because of the lack of experimental equipment, A statistical averaging of the observed intensities may also permit a distinction to be made between the possible space groups. If the compound is assumed to consist only of Na*, Ni2* and 0?2, chemical analysis for sodium and nickel of alcohol-washed products gives a composition which cannot exist. If, on the other hand, it is assumed that nickelous hydroxide is present, the chemical arnalysis is quantitively satisfied by a mixture of Na,NiO, with 40 wt % Ni(OH),. Such o mixture would be plausible. It is not unlikely that sodium orthonickelate(ll) is the original product of the reaction, that it reacts with atmospheric water to form nickelous hydroxide and sodium hydroxide, and that the sodium hydroxide is removed by the alcohol washing. Dehydration of Sodium Hydroxide G. P. Smith Metallurgy Division It is well known that sodium hydroxide can be dehydrated by heating to 400°C under o vaccum, However, the details of this process have not previously been reported. In studies of the re- action of sodium hydroxide with nickel, as de- scribed above, it was possible to measure the evolution of water during the course of dehydration. It was found that virtually all the woter evolution occurred near the melting points of sodium hy- droxide monchydrate and sodium hydroxide. This result is not surprising in view of the enhanced diffusion rates in the liquid as compared with the solid. However, the quantitative extent of this effect is greater than might have been predicted. Figure 5.15 is a plot of pressure and temperature vs time during one of four similar dehydrations. As was shown in the preceding section, an area under the pressure-time curve for a specified time interval is proportional to the amount of gas evolved during that time interval, Hence, most of the water was-evolved within the relatively narrow limits of two peaks or bands, the first of which corresponds to the melting of sodium hydroxide monchydrate and the second of which corresponds to the melting of sodium hydroxide., The shaded area on the second peak connects the upper and lower limits of rapid pressure fluctuations which were probably caused by the bursting of water- vapor bubbles, Color Changes in Fusea Hydroxides C. R. Boston M. E. Steidlitz Metallurgy Division Color changes observed in fused hydroxides indicate the existence of species in thermal equi- librium other than the expected sodium ions and hydroxyl ions. Several years ago it was noted in this laboratory that colorless fused sodium hy- droxide in a nickel container under an atmosphere of hydrogen turned green when heated to above 500°C and again became colorless when cooled to below 500°C, This color change was at first thought to be due to a changing concentration of PERIOD ENDING JUNE 10, 1954 nickel ions with changing temperature. Subse- quently it was found that the same color change took place ‘when the hydroxide was contained in silver. These observations indicated that the color was not due to the presence of heavy-metal ions. This qualitative study of color changes has been greatly extended. The experimental method used consisted of placing the hydroxide in a crucible which was con- tained in a quartz tube which could be evaocuated with a mechanical pump and to which various atmospheres could be admitted. The crucible of hydroxide was heated by a small pot-type furnace in which the hydroxide could be observed. The gases were either dried by passing them through a liquid-nitrogen cold trap or they were saturated with water vapor by bubbling them through water. In most of the experiments the crucibles used were pure sintered aluminum oxide (Morganite) to ensure the absence of heavy-metal ions. Morganite is very slowly and unifoermly attacked by fused hy- droxides. However, of the varicous ceramic cru- cibles available, Morganite was by far the most resistant to corrosion. In a few tests, nickel crucibles were used. The lithium, sodium, and potassium. hydroxides used were cp grade. The rubidium and cesium hydroxides were supplied by L. G. Overholser of the ORNL Materials Chemistry Division. UNGLASSIE(ED ORNL LR —DWE 1347 400 300 - PRESSURE microns) ( I 400 ! ! . s st TEMPERATURE - e T T o) 300) \S//M s —_ ‘ 4/ / } é} - A ‘ = 4 200 b e e e e e X e e g 200 & - 5 G ; ‘ a r | 2 //, - L[] T R et mmaaaanan e ____,,A___,,_,,__[___________,___,_“____ 100 PRESSURE ' ! e Q 50 60 70 80 90 TIME {(min) Fig. 5.15. Water Evolution During the Dehydration of Sodium Hydroxide ot a Slow Heating Rate, 85 ANP QUARTERLY PROGRESS REPORT A series of experiments was conducted in which the hydroxides were heated to approximately 600°C under specified atmospheres. At 6009C the initial atmosphere was evacuated and various atmospheres were successively admitted and evacuated. Finally the hydroxide was cooled to room temperature, During each step the color changes were noted. The colors of the hydroxides (contained in Mcrganite) which developed on standing in specified atmospheres at 600°C are shown in Table 3.10. For each combination of fused hydroxide and atmosphere which was studied, the hydroxide was colorless below approximately 500°C and distinctly colored at 600°C. The colorless-colored transition was reversible with changing temperature. When the system was evocuated or filled with a specified gas, the color changed sometimes with startling abruptness after a brief induction period, Lithivm hydroxide did not lend itself well to testing because of its tendency to decompose with the evolution of water. Sodium hydroxide, after being heated for a long period in either air or vacuum, changed in color from yellow to a deep red-orange. This color, however, disappeared promptly on cooling or when the atmosphere was changed, Several experiments were conducted with sodium hydroxide contained in nicke! crucibles, The colors found were the same as those observed with a Morganite container except when an atmosphere of air wos used. In this instance, reaction with the nickel quickly turned the hydroxide black. Experiments were performed in which various sodium compounds were added to sodium hydroxide and the mixtures were heated under a vacuum fto a temperature of about 400°C, These additives TABLE 5.10, COLORS OF HYDROXIDES IN VARIOUS ATMOSPHERES AT 600°C were of commercial purity. Metallic sodium and sodium hydride additions gave, at first, a blue color which, in time, turmed to green. Sodium peroxide additions imparted an orange color. With sodium oxide additions the hydroxide was yellow. It should be noted that these colors were formed ot a temperature at which the pure hydroxide would have been colorless. All the colors were retained vpon cooling the hydroxide to below its melting point. The results of these experiments are still too incomplete and too qualitative to permit specific conclusions as to the species giving rise to the various colors, Nevertheless, a number of tentative speculations can be made. First, various simple atmospheres such as oxygen, hydrogen, and water vapor can interact with the fused alkali-metc! hydroxide to produce different chemical species which can be characterized, at least to a limited extent, by their color, These interactions do not seem to be greatly altered by progressively chang- ing the cation from sodium to cesium, Hence, they seem to be associated with the anion. The extent of interaction is rapidly established and thereafter seems to change only very slowly with time at constfant temperature, The extent of interaction seems fo increase markedly with incregsing temper. ature. It is significant that these hydroxide-gas inter- actions take place in the absence of free metal, It is well khown that the atmosphere plays a dominant role in controlling the mass transfer of nickel, It is not entirely unlikely, then, that the chemical reactions that cause mass transfer do not take place directly between nicke! and the hydroxyl ion, as is generally believed, but rather between nickel HYDROXIDE Air Vacuum LiOH Yellow NaOH Yellow, Yellow, orange orange KOH Yellow Yellow RbOH Yellow Yeliow CsOH Yellow Yellow 86 H H Water Vapor 2 € in H2 or He b - Green Yellow Colorless Green Graen Green and a chemical species whose concentration is dependent on an interaction between the hydroxyl ion and the atmosphere. Preliminary measurements have been made of the electrical conductivity of fused sodium hydroxide contained in silver exposed to the air at temper- atures up to 550°C. The results are in fair agree- PERIOD ENDING JUNE 10, 1954 ment with those reported by Aradt and Ploetz'® who give data up to 450°C, It is interesting to note that the conductivity was found to rise rapidly over the temperature range for which a color chonge is first noted. 161, Arndt and G. Ploetz, Z. physik. Chem. 110, 237 (1924). 87 ANP QUARTERLY PROGRESS REPORT 6. METALLURGY W. D. Manly Metalturgy Division The studies of the mechanical properties of [nconel in contact with the fused fluorides have shown that material stressed under uniaxial con- ditions has a much longer rupture life than material tested under multiaxial conditions. Since in ony ANP type of reactor the material will be under multiaxial stresses, the testing of alloys for this service should be performed by using tube-burst tests. The uniaxial test conditions are much easier to control and will give a much better in- sight into the effect of environments, but for ob- taining actual design data, the tube-burst tests or other multiaxial-stress tests ore preferred. It has been found that the load-carrying abilities of Inconel, as well as the corrosion, are also quite dependent on the ratio of surface area to volume of fluoride mixture. As the rotio of surface area of container material to volume of fluoride mixture decreases, the rupture life for a given siress is also decreased. Investigations of materials svitable for high- conductivity fins have been primarily concerned with the protection of copper from oxidation and with the prevention of diffusion between the cladding material and the copper. This work has shown that copper clad with types 310, 446, or 430 stainless steel or with Inconel is quite satis- tactory in the unstressed condition if a suitable diffusion barrier is provided. In oxidation tests of the clod material under the stress, however, it was found that high stresses greatly increase the oxidation of the material. Additional tests will be made at lower stresses and for {onger times. In the search for container materials for fluid fuels, Hastelloys B and C were found to have very poor extrusion characteristics. A comparison with the results of extrusion experiments with nickel- molybdenum special alloys indicates that the diffi- culty in extruding Hastelloy B was probably caused by the impurities in the material and the poor melting practice used in its preparation, rather than poor lubrication during extrusion. The work- hardening characteristics of columbium were de- termined, Sheets of high-boron-content material will be 88 required for shielding between the moderator and the heat exchanger and between the heat exchanger and the pressure shell of the Reflector-Moderated Reactor. Powder compacts of boron carbide blended with copper and with silver powders were hot pressed, and boron densities which approached the density required were obtained. It is thought that the required density can be obtained by warm- pressing compacts of boron carbide and copper, High-temperature oxidation-resistance tests of several brazing alloys at 1500 and 1700°F for 200 and 500 hr have shown that the majority of the nicke! and nickel-chromium-base alloys are suit- able for service in an oxidizing atmosphere at 1500°F and that severa! of the alloys retain this resistance at 1700°F, A new semiautomatic heliarc-welding process for production of tube-to-header joints is described and its use in the construction of a prototype sodium-to-air radiator is illustrated. The fabri- cation of a radiator with high-conductivity fins requires a brazing alloy that has good strength at 1600°F and flowability at approximately 1900°F. Coast Metals alloy No, 52, which will svitably wet stainless steels ond flow in dry-hydrogen atmospheres, has been found to be satisfactory for this use. In addition to having a suitable alloy, the furnace temperature must be very care- fully controlled so that there will be a minimum of thermal variation over the assembly. To test the brazing alloys and te check the controls of the furnace, a prototype radiator with high-con- ductivity fins was constructed. The successful fabrication of this prototype radiator indicates that such complicated configurations can be con- structed, An assembly consisting of a beryllium plate canned in an Inconel assembly in which the be- ryllium will be heated by electrical resistance and cooled by sodium passing through holes in the beryllium plate was fabricated. This unit will be used for determining the effect of thermal stresses and thermal cycling on beryllium metal and for studying the compatibility of sodium and beryllium, STRESS-RUPTURE TESTS OF INCONEL R. B. Oliver D. A. Douglas J. W. Woods Metallurgy Division The previous report! presented results for a series of stress-rupture tests of tubular Inconel specimens under multiaxial stress in contact with NGF-ZfF“*UF; (50-46-4 mole %). Some additional dota which compare fime to rupture for multiaxially stressed tubular specimens with that for uniaxially 1R, B. Oliver, D. A. Douglas, and J. W. Woods, ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 84. PERIOD ENDING JUNE 10, 1954 stressed flat specimens in the same fluoride mix- ture are presented in Fig. 6.1. The previous re- sults were for specimens tested with the fluoride mixture inside the tube and purified orgon gas outside. Metallographic examination of the tubes after rupture did not show the void formation which characterizes the type of corrosive attack found on the straight tension tests. The surface ap- peared uneven, but there was no evidence of inter- granulor attack., With the fluoride mixture inside the tube, there is a ratio of about ten units of surface area to one of volume, It was thought that the change in the corrosion rate could be CRNL-LA-DWG 1587 104 , - j ‘ 1 ( { ’ J J R | TS A S S O T Y R A , - ot L1 — e 5 em m e rm i e b kb e saaed o aiiqgaaen L e barmee mmmmmnnf e — e """"""""" T \ T 2 . : w . o = \ . - - o 0.010-in.~WALL TUBING WITH — ‘/r--0.062—|n.—THICK SHEET IN TENSION ’5.‘:’ osr — FLUORIDE MIXTURE INSIDE -y : \fi— 1 IN FLUORIDE MIXTURE 7 : - \ o b I e e L e i 0.010 -in.- WALL TUBING WITH FLLUORIDE MIXTURE OUTSIDE - 1 : ! l 10° A L 10 2 5 10% 2 5 10° 2 5 10% TIME TO RUPTURE (hr) Fig. 6.1. Stress vs Time to Rupture for inconel Sheet and Inconel Tubing in NaF-ZrF -UF, {50-46-4 mole %) ot 1500°F. 89 ANP QUARTERLY PROGRESS REPORT explained on the basis that the surface-to-volume ratio was too high to sustain the corrosive attack. With this point in mind, tests were run with the fluoride mixture outside the tube and argon inside, with a resulting ratio of one unit of surface area UNCLASSIFIED Y-12495 - Fig. 6.2. the Fluoride Mixture Was Inside the Tube and the Resulting Surface-to-Yolume Ratio Was 10 to 1. etched. (b) Etched. 250X. . ' - UNCLAS&?F!ED ¥-12513 Fig. 6.3. Inconel Tubing After Exposure to NaF.ZrF -UF to eight units of volume. The types of corrosive attack on these specimens are illustrated in Figs. 6.2 and 6.3. The attack shown in Fig. 6.3, which is similar to that found in uniaxial tests, differs radically from the attack found when the fluoride - TUNCLASSIFIED Y-12530 Incone! Tubing After Exposure to NaF-ZyF -UF , ot 1500°F in Stress-Rupture Test in Which (@) Un- ’ fi*‘ '&g“f&xngt‘f l“&xi‘flf o o SigTh ,;? AP UNCLASSIFIED +; : M.‘ Ao Y174 .’3 e . . N } e e fiv'“ *h t.‘l?j}w % “‘h\ N“x’-‘*\'t ‘ w‘fi ’fi(:éw*} at 1500°F in Stress-Rupture Test in Which the Fluoride Mixture Was Outside the Tube and the Resulting Surfoce-to-Volume Ratio Was 1 to 8. (a) Unetched. (b) Etched. 250X. 90 mixture was inside the tube (Fig. 6.2). It is therefore concluded that the surface-to-volume ratio is as importont in stress-rupture tests os it is in corrosion tests. It appears that under these test conditions a certain minimum volume of fluo- ride mixture in relation to the surface area is required for corrosion to proceed. Since stress- rupture tests in fluoride mixtures are being run at a number of testing laboratories, some of which will be in support of the efforts of ORNL, it seems imperative that the best surface-to-volume ratio for reproducibility be determined. Otherwise, a comparison of the data from the various labe- ratories may lead to confusion or erroneous con- clusions. ’ Another variable which should be considered in analyzing tube-burst data is the wall thickness of the specimen being tested. The data presented in Fig. 6.1 are for tubular specimens with 0.010- in.-thick walls. Naturally, corrosive attack will have more effect on this type of specimen than on a thick-walled tube. Specimens with wall thick- nesses of (0.020, 0.040, and 0.060 in. are being tested so that the effect of this variable can be studied, [t is expected that all the tubular speci- mens will fail in shorter times than will corre- sponding specimens tested in. straight tension, One reason for this prediction is that in the tube- burst test there is a tangential-to-axial stress ratio of 2 to 1. It is generally recognized that this stress distribution will result in the minimum ductility for any given material. No strain meas- urements are made during the tube-burst test; how- ever, after-teést measurements do not show meas- urable deformation that would be evidence of the brittle behavior of Inconel in this type of test. The lowered ductility of the multiaxially stressed specimen is believed to be the prime reason for its early failure. : Since the available data from tube-burst tests indicate that multiaxial stresses reduce rupture life in comparison with rupture life under the ‘uniaxial stresses used in the more conventional straight-tension tests, it appears that more em- phasis should be placed on the tube-burst tests. Information on the behavior of Inconel tubing under multiaxial stress will be particularly pertinent to ANP-type reactor design. ' PERIOD ENDING JUNE 10, 1954 HIGH-CONDUCTIVITY METALS FOR RADIATOR FINS E.S. Bomar ‘H. lnouye J. H. Coobs R. W. Johnson Metallurgy Division Investigations of fin material have been primarily concerned with the protection of copper from oxi- dation and with the prevention of diffusion between copper and the cladding material. The materials that were found to be satisfactory as cladding were types 310, 446, and 430 stainless steel and In- conel, provided a suitable diffusion barrier was placed between the cladding ond the copper. Since the purpose of the investigation was to develop a high-thermal-conductivity fin, a large ratio of copper to cladding was obviously desirable. With the roll-cladding techniques employed, the minimum clad that could be applied successfully was about 2 mils thick, although clads as thin as 1 mil have been made successfully by other methods. The minimum cladding thickness is apparently de- pendent only on ability to fabricate the composite. Under the conditions that the fin material will be used it must be assumed that there will be stresses. | herefore tensile-strength tests of the composites under oxidizing conditions are being made. _ - A few exploratory tests were made on 8-mil-thick type 310 stainless-steel-clad copper composites; the cladding was 2 mils thick on each side of the copper. The composites made by the General Plate Compony were tested in the as-received condition. The results of a few tests are pre- sented in Table 6.1. Additional tests at different stresses and temperatures are contemplated.. No definite conclusions can be made at this time, but it is evident that low stresses caused failures that were not evident in tests under no-load con- ditions. SPECIAL MATERIALS RESEARCH E. S. Bomar H. Inouye J. H. Coobs R. W. Johnson Metallurgy Division Hastelloys B and C Seven rod extrusions of as-cast Hastelloys B and C were made to determine the optimum temper- atures and the general fabrication characteristics 91 ANP QUARTERLY PROGRESS REPORT of these alloys. The extrusion data are presented in Table 6.2. These extrusions were successful in that they demonstrated that the material could be extruded. However, in nearly every instonce the extruded rod cracked at the leading edge, and in one instance the extrusion shattered. At the present stage of investigation it is thought that the cracking of the leading edge is due to inter- mittent welding of the alloy to the die (‘‘rattle- snaking'’) because of insufficient lubrication. Fractures of a different kind occurred near the center of the extruded rod that probably resulted from faults in the billets — the melts were made in air. Shattering is believed to be caused by low-melting phases in the grain boundaries. These extrusion difficulties may be solved by (1) making the melts in vacuum and thus mini- mizing atmosphere contamination and probably the low-melting constituents, since purification might also be accomplished, (2) changing the die shape to cause more working in the leading edge, (3) canning the billet in such a manner that the alloy TABLE 6.1. TENSILE STRENGTH TESTS OF TYPE 310 STAINLESS-STEEL-CLAD COPPER AT 1500°F IN AIR STRESS TEST DURATION ELONG;\TION REMARKS {psi) (ki) (% in 2 /2 in.) 500 500 5 Surface blistery but otherwise sound 1000 500 20 Siringers of oxide on surface; a few nodules of copper oxide 1500 500 47.5 Numerous copper oxide nodules on surface; arsas oxidized throughout; considered to be a failure 1800 336 50 Sample ruptured 2000 96 39 Sample ruptured TABLE 6.2. EXTRUSION DATA ON HASTELLOYS B AND C Die: CHW* preheated to 572°F, 45-deg cone Lubrication: glass wool in container Ram sleeve: 3]/6 in., preheated to 692°F Billet: 3% in. long, 3 in. OD, heated in Houghton's No. 1550 salt bath EXTRUSION HEATING EXTRUDED RAM PRESSURE ({tons) BILLET TEMPERATURE TIME SIZE °F) (he) (in.) Start Run Hastelloy B 2300 % 1Y 548 400 2300 1y, 1 550 420 2100 17% 1% 560 390 2100 20 17, 615 450 Hastelloy C 2100 18 1‘/2 530 390 2100 20} 1, 565 460 2100 4 1 570 450 *Manufactured by Latrobe Electric Steel Company. 92 is placed where lubrication is known to be ade- quate, that is, by using a dummy in front of the billet. ' : The short billets (3]/2 in.) were extruded at high velocities, and in many instonces the tools were broken when the ram struck the die. This diffi- culty in stopping the ram movement at a precise moment was prevented by using a brass dummy block behind the hillet so that the ram could be allowed to move until the press stopped. Nickel-Molybdenum Alloys Two vacusm melts of nominal composition 24% Mo~76% Ni and 20% Mo-80% Ni have been made. Attempts to extrude the 24% Mo-76% Ni alloy ot 2350°PF at reductions of 13 to 1 and 9 10 1 failed because of tool breakage. In the attempt at ex- trusion at a ratio of 9 to 1, the die insert fractured. A solid die was made, and attempts to extrude the 20% Mo—-80% Ni alloy with the new die failed because of mandrel breakage. A straight mandrel and a 25-deg cone die were used. The extruded material that was obtained before the mechanical failures occurred had good surfaces and ne evi- dence of ‘‘rattlesnoking.”’ Stainless-Steel-Clad Molybdenum and Columbium Molybdenum tubing (0.620-in.-OD, 0.098-in.-wall) was clad on the outside with 0.065-in.-thick type 321 stainless steel. The composite tube was successfully bent 105 deg on a 3-in. radius at temperatures around 400°F and made into the shape of a conventional thermal-convection loop. Attempts to weld the molybdenum were unsuc- cessful because of porosity and extreme brittle- ness. Another compliceting problem was that, ot the temperatures required to fuse the molybdenum, the cladding melted ond alloyed in the weld. These difficulties were circumvented by assem- bling the loop with threaded molybdenum joints and welding the stainless steel cladding. Two such loops were made. Type 310 stainless-steel- clad columbium tubing has been bent into the shope of a conventional thermal-convection loop, but the joint has not yet been welded. The bending was accomplished at room temperature by using Cerrobend inside the tube. Columbium Fansteel columbium sheet 0.250 in, thick was reduced to 0.193 in, in thickness and annealed in PERIOD ENDING JUNE 10, 1954 vacuum to an ASTM grain size of 6 to 7. The work-hardening characteristics were determined by making Tukon hardness measurements at various stages of reduction up to 89.6% in thickness. The results of these tests are given in Table 6.3. The data show that hardening in cold-worked co- lumbium does not increase rapidly, and therefore the metal may be severely deformed without danger of fracture. A similar investigation? showed some- what higher hordness values for the same re- duction., ' TABLE 6.3. WORK-HARDENING CHARACTERISTICS OF COLUMBIUM SHEET AT VARIOUS STAGES OF REDUCTION OF THICKNESS THICKNESS REDUCTION iN HARDMESS Gn) THICKNESS VPN (%) ' 0.193 Annealed 72.4 0.170 1.9 98.2 0.152 21.2 100.2 0.135 30.0 108.3 0.115 40.3 111.1 0.095 50.2 115.5 0.077 60.1 123 0.058 70.0 127 0.040 79.3 . 134 0.025 87.0 144 0.020 89.6 150 Boron Carbide Thin high-boron-content shielding is required for the heat exchanger of the Reflector-Moderated Re- actor. This shield will be in two ]/s-in.—thick layers, one inside and one outside the heat ex- changer, and it should have as high a boron content as practical. A concentration of 1.5 g/cc of boron, equivalent to B,C compacted to 80% density, would be sufficient. The temperature of the shield, which will be about 1500°F, severely limits the choice of bonding materials for use in consolidating the 2w. o. Alexander, Observations on the Fabricating Properties of Columbium Sheet, BR-627 {July 12, 1945}. 93 ANP QUARTERLY PROGRESS REPORT boron carbide. lron and nickel both react with B,C to form eutectics and brittle intermetallic phases, but they have been used to bond B,C by hot-pressing or high-temperature sintering. Copper and silver are more attractive since they do not react with boron carbide and would therefore retain their strength and conductivity after being formed into a composite. However, silver is somewhat objectionable because of its relatively high cross section and associated gamma activity, Hirsch® has reported fabricating strong compacts of boron bonded with 15 vol % of copper or of silver with horon densities as high as 1.52 g/cc by warm-pressing at 500°C with 50-tsi pressure. In the hope of producing satisfactory compacts by duplicating this work with high-grade B,C, compo- sitions containing 15 vol % of copper or 20 vol % of silver were prepared. Fine, annealed electro- lytic copper and precipitated silver powders were used with blends of various particle-size fractions of B,C. These mixtures were then pressed at 350°C with 40-tsi pressure, and the density was determined from the measured volume. The maximum boron density obtained from these compositions thus far is about 1,25 g/cc for high- grade B4C bonded with 20 vol % of silver. Den- sities attainable by using sized fractions of metal- lurgical-grade B,C are 1.1 g/cc of boron with 20 vol % of copper binder and about 1.2 g/cc with 20 vol % of silver binder. By using compositions of high-grade B,C (>80% boron) with copper and by pressing at 500°C, it should be possible to obtain compacts with boron densities at least as high as 1.3 g/cc. This density is somewhat lower than desired but may still be considered. The attainment of higher densities will require one of three alternatives: 1. the use of metallic boron bonded with copper by warm-pressing, as reported by Hirsch, 2. the high-temperoture hot-pressing of B, C, with or without iron or other reactive bonding agent, 3. the high-temperature sintering, at 1900 to 2000°C, of carefully prepared B,C-Fe compo- sitions, for which boron densities as high as 1.45 g/cc have been reported.? A wmore complete investigation of the warm- pressing technique is to be made. For this in- SH. Hirsch, J. Met. Ceram., TID-67, p 7 (May 1949). YKnolls Atomic Power Laboratory, Report of Metal- furgy Section for December 1952 and Janvary and Feb- ruary 1953, KAPL-894. 94 vestigation, 15 to 20 vol % of copper in BAC will be pressed at 500°C with 40- to 50-tsi pressure. A new die of high-alloy steel will be required for satisfactory service at this temperature. In addition to the problem of fabricating a satis- factory shield, the possible effects of boron dif- fusion must be studied. Since the shield will be in contact with Incone! at operating temperatures of 1500°F, extensive diffusion of boron may occur. Samples of Incone! coated with Al,O, are being obtained for compatibility tests with boron-con- taining compacts, WELDING AND BRAZING C. E. Shubert P. Patriarca G. M. Slaughter K. W. Reber B. McDowell J. M, Cisar Metallurgy Division Brazing Alloy Development The extent and rate of high-temperature oxidation of brazing alloys during service are among the important factors which will determine the useful life of ANP radiators. Inconel T joints brazed with 15 different high-temperature brazing alloys were subjected to static air at 1500 and 1700°F for periods of 200 and 500 hr. The results of these tests, as determined by metallographic examination (100X magnification) in the as-polished condition, are presented in Table 6.4. It appears that the majority of the nickel and nickel-chromium-base alloys are suitable for service in an oxidizing atmosphere at 1500°F, aond several are suitable at 1700°F. However, since G-E alloy No. 81 has shown considerable promise for tube-to-fin joints and as a backup for tube-to-header joints, its unexpected behavior ot 1700°F is being investi- gated further. |t was encouraging to find that the Coast Metals alloy No. 52 was very oxidation resistant, since its favorable flow point makes this alloy useful for joining high-conductivity fin materials, A prototype sodium-to-gir radiator utilizing In- conel-clad-copper high-conductivity fins was fabri- cated by a combination heliarc-welding and dry- hydrogen-brazing technique. The assembly, shown in Fig. 6.4, contains 180 heliarc-welded tube-to- header joints and approximately 4300 brazed tube- joints. As an added precaution against undesirable stress concentrations and leaks during service, the tube-to-header joints were back brazed. to-fin §6 TABLE 6.4. OXIDATION RESISTANCE OF DRY-HYDROGEN-BRAZED T JOINTS (O)Very slight, tess than 1 mil of penetration. Stight, 1 to 2 mils of penetration. Moderate, 2 to 5 mils of penetration, Severe, greater than 5 mils of peretration. Complete, fillet completely destroyed, )y oid formation caused by internal oxidation at braze alloy fillet~Inconrel interface. OXIDATION IN STATIC AIR(Q} COMPOSITION BRAZING S > BRAZING ALLOY (wt %) TEMPERATURE At 1500°F A+ 17007 F e o F) For 200 hr For 500 hr For 200 hr | For 500 hr Commercial Alloys i ! f r Nicrobraz 7O Mi-14 Cr—6 Fe-58-45i-1C 2150 Slight Slight { Stight j Stight L ow-melting-point Microbraz | 80 Ni—6 Fe--5Cr-55i-3 B-1C 1920 Slight Stight | Stight Stight Coast Metals alloy No. 52 89 Ni-5S5i—-4 B-2 Fe 1840 Very slight Slight Slight Slight ~Mond Nickel Co. alloy 64 Ag~33 Pd-3 Mn 2150 Severe internal | Severe internal | Complete oxidation oxidation , ¢ = s Copper 100 Cu 2050 Complete | E Experimentui Nickel-Bose Alloys GE alloy No. 81 66 Ni-19 Cr-10Si-4 Fe-1l Mn 2150 Very stight Slight Severs®! Severe(b) Ni-Ge 75 Ni~25 Ge 2150 Stight Slight ! Moderate Moderate Ni-Ge-Cr 65 Ni-25 Ge=10 Cr . 2140 Yery slight Stight Modercfe{b) Severe Electroless Ni-P 88 Ni-12 P ] 1740 Very slight Slight 80 Ni=10 P=10 Cr 1830 Very slight Slight 68 Ni-32 5n 2150 Slight Moderate Severe Complete 40 Ni=60 Mn 1920 Complete | Precious-Metal-Base Alloys Au-Ni 82 Au~18 Ni 1830 Very stight R!ight Moderate Moderate Au-Cuy 80 Au—20 Cu 1740 | Moderate Complete Pd.al 92 Pd-8 Al 2020 Very slight Very slight Very slight Sligh-t #S6L ‘Ol ANNT ONIAON3 QONRI3d ANP QUARTERLY PROGRESS REPORT Fig. 6.4. Sodium-to-Air Radiator. All tube-to-header joints were heliarc welded by a semidutomatic process. In this process, a com- mercially ovailable heliarc torch is rotated around the tube periphery by a variable-speed motor-driven offset-cam mechanism which can be odjusted to any desired diameter of swing. A photograph of this equipment is shown in Fig. 6.5. Preliminary experiments were conducted on typical tube-to-header samples to investigate the effects of arc current and welding speed upon weld size, hole constriction, and weld penetration. It was shown in these tests that welds comparable to those produced manually by a skilled operator® could be made semigutomatically. satisfactory welds having a penetration of at least one tube-wall thickness could be made ot a weld time of 7 sec over a welding current range from 40 to 50 amp. The hole consiriction resulting from welds made over this current range did not exceed S%. For example, 3p, Patriarca, G. M. Slaughter, and J. M., Cisar, ANP Quar, Prog. Rep. Dec. 10, 1953, ORNL-1649, p 86, Fig. 7.3 96 UNCLASSIFIED Y.12494 Fig, 6.5. Semicutomatic Tube-to-Header Welding Machine. As would be expected, welds made at other welding speeds required corresponding changes in the welding variables. Further experiments are now being conducted to study more thoroughly the relationships between the joint parameters and the welding variables. Since it is expected that the maximum weld penetration will be dependent to a significant extent on the radius of electrode swing, experiments will be conducted to determine this effect. The fabrication of a radiator containing high- conductivity fins requires close control of the furnace temperatures during brazing fo assure a minimum of thermal variations over the assembly, Experiments were conducted which indicated that the large Globar pit furnace available in the Welding Laboratory would permit a temperature control of $5°F over the assembly at 1870°F and hence would be satisfactory for this fobrication, Coast Metals alloy No. 52 (Table 6.4) was se- lected as the brazing material because of its superior oxidation resistance and its relatively low flow point of 1840°F. A preheat temperature of 1520°F was utilized in this dry-hydrogen-brazing operation to permit an equalization of temperatures over the assembly before brazing began. The prehecot portion of the therma! cycle aided fillet formation by minimizing selective flowing of the brazing alloy to the fins. Boron diffusion from this alloy was not a problem during preheating because very little sintering occurs at this temperature, It also appears that the fin distortion on the assembly was kept to a minimum as a result of the preheating and the careful temperature control during brazing. Good flowability of the alloy on both the tube- to-fin and tube-to-header joints was obtained during brazing for 30 min at 1870°F. Since the copper exposed by the fin-punching operation can be protected adequately by the brazing alloy, it is believed that the oxidation resistance of the tybe-to-fin joints should be satisfactory. Beryllium Test Assembly An Inconel assembly for determining the effects PERIOD ENDING JUNE 10, 1954 of thermal stresses and thermal cycling on be- - ryllium was fabricated. This unit, which will be used to evaluate the temperature distributions in the beryllium, as well as the compatibility of sodium and beryllium, consists of 46 tube-to- header joints and extensive manifold welding. The beryllium block is enclosed in the center section of the assembly shown in Fig. 6.6. All joints were manually heliare welded by%quol-ified operators and were leak-tight to helium. Since heating will be achieved by the electric resistance heating of sodium, it was necessary to fabricate two lominated-copper bus bars and to join these to the Inconel assembly. A silver- copper eutectic brazing alloy was used to join the copper laminations into a solid electrical con- nection. However, since this silver brazing alloy will not wet inconel satisfactorily in the available dry-hydrogen atmosphere at the brazing temperature of 1510°F, it was necessary to copper-braze nickel sheets to the Inconel cans before subsequent joining of the bus bars to the test assembly with the silver-copper eutectic alloy. 97 98 ANPF QUARTERLY PROGRESS REPORT Fig. 6.6. Beryllium Test Assembly. UNCLASSIFIED Y-12546 o T b PERIOD ENDING JUNE 10, 19-54 7. HEAT TRANSFER AND PHYSICAL PROPERTIES | H. F. Poppendiek | Reactor Experimental Engineering Division The enthalpies and heat capacities of the ARE fuel NoF-ZrF -UF, (53.5-40.0-6.5 mole %) were detfermined; the heat capacity in the solid state over the temperature range 260 to 490°C was found to be 0.19 cal/g-°C, and.the heat capacity in the liquid state over the temperature range 590 to 920°C was found to be 0.23 cal/g-°C. Enthalpy and heat-capacity measurements for K;CrF in the solid state were also obtgined. Density ond viscosity measurements were made for molten RbF-LiF (57-43 mole %); the viscosity varied from aobout 8 centipoises {cp) ot 550°C to about 3 cp at 750°C. The thermal conductivity of molten RbF-LiF (5743 mole %) was determined to be about 1.2 Btu/hr-sq ft (°F/ft}. The thermal con- ductivity of solid MNaF-KF-LiF (11.5-42.0-46.5 mole %) was found to be about 3 Btu/hresq ft (°F/ft). Electrical-conductivity measurements of molten NaOH were obtained over the temperature range 625 to 1490°F. Additional forced-convection heat-transfer meas- urements of molten NaF-KF-LiF were made at higher Reynolds numbers than -had previously been obtained. Some thermal-conductivity information on K,CGrF, o wall-deposit material, was obtained. A Lucn‘e model of the circulating-fuel reflector- moderated reactor has been fabricated ond will be used to study the hydrodynamic structure in that system. The results of a mathematical analysis of convection dare presented for the case of forced flow between parallel plates which are ducting fluids with volume heat sources; this analysis is useful in estimating the temperature structure in the flow annuli of reflector-moderated reactor cores. A study has been initiated to investigate the heat-transfer and fluid-flow characteristics' of a NaDH-moderated circulating-fuel reactor system, PHYSICAL PROPERTIES MEASUREMENTS: Heat Capacity W. D. Powers - G. C, Blalock Reactor Experimental Engineering Division The following enthalpy and heat-capacity meas- urements were made with Bunsen ice calorimeters: NaF-ZrF -UF ; (53.5-40.0-6.5 mole %) Solid (260 to 490°C) Hy — Hpoe = —4.1+ 0,197 C, = 0.19 +0.02 Liquid (590 to 920°C) ; Hyp ~ Hypoe = 34 + 0.23(5)T C, = 0.23(5) + 0.03 ‘K CrF Sohd (50 to 830°C) Hi = Ho. = —4 + 0.228)T | - p 0.22(8) * 0.006 NaF-KF-LiF (11.5-42.0-46.5 mole %) Solid (60 to 455°C) H -«.H\ = «-26+027T+098x10“"12 T C, = 0.27 + 1.96 x 1077 In these expressions H is the enthalpy in cal/g, Cp is the heat capacity in cal/g-°C, and T is the temperature in °C. The enthalpy of K CrF, (an insoluble deposit material) is shown in th 7.1, The enthalpy and heat-capacity data for NaF-KF-LiF (11.5-42.0-46.5 mole %) in the solid state corre- spond to o much greater temperature range than was previously reported. ‘A copper calorimeter has been constructed that will be used to supple- ment the Bunsen ice calorimeters now in use. The heat content of the sample will be measured by the temperature rise of the copper housing. Density and Viscosity S. |, Cohen T. N. Jones Reactor Experimental Engineering Division Density and viscosity measurements were made on two fluoride mixtures. The viscosities of RbF-LiF (57-43 mole %) were determined with modified Brookfield and capillary viscometers, ond the values varied from about 8.0 ¢p at 550°C to about 3 cp at 750°C, The density measurements of RbF-LiF in the molten state are represented by the relation p = 3.39 — 0.00085T , 99 ANP QUARTERLY PROGRESS REPORT UNCLASSIFIED ORNL-LR—-DWG 41414 - 200 - R ENTRALPY {cal/q) 400 o 00 800 1000 TEMPERATURE (°C) Fig. 7.1. Enthalpy-Temperature Relationship for K,GF . where p is in g/cc and the temperature range is 500°C < T < 700°C. The density and viscosity values! for NaF-KF-LiF (11.5-42.0-46.5 mole %) that were determined in the early stages of the ANP Project were checked, and the recent determinations were in substantial agreement with the early ones. The viscosities for the mixture vary from about 8 cp at 500°C to about 3 cp at 750°C, and the molten densities are represented by the relation 2.53 ~ 0.00073T , 600°C < T < 900°C . A report? published recently lists all the density measurements that have been made on molten fluoride mixtures for the ANP Project and includes a correlation of the data that can be used to pre- dict densities of molten fluoride mixtures of known composition. p"'_.‘ Thermal Conductivity W. D. Powers S. J. Claiborne R. M. Burnett Reactor Experimental Engineering Division The Deem-type apparatus has been used to meas- ure the thermal conductivity of RbF-LiF (57-43 mole %). The conductivity was determined to be 1.2 VANP Physical Properties Group, Physical Property Charis for Some Reoactor Fuels, Coslanis, and Miscel- A.'lanec;us Materials, 3d ed., ORNL CF-53-3-261 (Mor. 20, 953). 100 Btu/hr-sq ft (°F/ft) at an average sample temperc- ture of about 1050°F. A flat-plate method for determining the thermal conductivities of solid salts has been developed in which the salt is cast in the form of a slab, The heat that is passed through the sample is measured by the rise in temperature of the water that circulates through the cooling plate, The temperature drop through the sample is also meas- vred. The conductivity may be calculated casily from this information and the physical dimensions of the slab. The preliminary thermal conductivity value obtained for NaF-KF-LiF (11.5-42.0-46.5 mole %) by using this method was about 3 Btu/hr-sq ft (°F/f1). Additional thermal- conductivity measurements have been made on solids by using the transient cooling or heating technique. An improved method of casting nonporous spheres has been developed which invelves cocling the mold from the bottom at a very slow rate. Improved methads of cooling and heating the spheres have also been developed. The preliminary thermal-conductivity value obtained by using this method on NaF-KF-LiF (11.5-42.0-46.5 mole %) in the solid state was about 2.7 Btu/hr-sq ft (°F/&). Electrical Conductivity N. D. Greene Reactor Experimental Engineering Division A preliminary determination of the elecirical con- ductivity of molten NaOH at temperatures of up to 1500°F has been made with the platinum conduc- tivity cell, This determination has extended the temperature range of the electrical-conductivity measurements of NaOH to approximately 450°F higher than the maximum temperoture reported in the literature. The experimental conductivity data are shown in Fig. 7.2, together with the published measurements? for low temperatures, Checks on the platinum conductivity cell have been made by studying three molten salts for which conductivity values have been published. The salts investigated, together with the deviation between ORNL and literature values and the tem- 25, 1, Cohen and T. N, Jones, A Summary of Density Measurements on Molten Flucride Mixturas and a Cor- relation Useful for Predicting Densities of Fluoride Mixh.):res of Known Composition, ORNL-1702 (May 14, 1954). 3!nfernafiona! Critical Tables, vol 6, p 149, HNCLASGIFISD CRNL-LR -DWG 1415 7.00 oo 6.00 1 E 5 5.00 E ~ 4.00 D l:'. = 3.00 [ ‘ g g 2.00 P/ e bl ..__J....--- W e ORNL EXPERIMENTAL VALUES ‘ &4 LITERATURE VALUES 100 —_ .]._.._..{ - ...}.........r_........\....._,. - [ 4 e — O e ,A,._i. ......... ‘. ............... .J ................... l...,.._., 500 800 1000 ' 1200 1400 1600 TEMPERATURE {°F) : Fig. 7.2. Electrical Conductivity of Moliten NaOH. perature ranges, are listed in the following: ‘Mean Deviation Temperature Range | (%) (°F) NaOH +2.7 625 to B35 LiMC\3 +4.4 534 to 862 KNO, -3.3 875 to 900 The measurements are difficult to make becouse of polarization within the cell, and therefore some alternate method of measurement has been sought. It is felt that a hemispherical graphite crucible in which it would.be possible to position leads so that both current and potential measurements would be obtained simultanecusly might be useful. Since the potential measurement would be independent of the degree of polarization at the current-carrying electrode ond an alternating current would be used, it is estimated that good accuracy would be ob- tained. An. attempt to verify the tesults of the measurements made with the platinum conductivity cell will be made with this alternate method during the coming guorter, The conductivities at various pertinent temperatures of some of the salts which are of interest to the ANP Project will also be determined. Vapor Pressures R. E. Moore C. J. Barton Materials Chemistry Division In an attempt to resolve some anomalies, ob- served in another laboratory,? in vapor pressures of some NaF-ZrF , mixtures, the method and apparatus PERIOD ENDING JUNE 10, 1954 of Rodebush and Dixon® have been applied to a NaF-ZrF, mixture contoining 75 mole % ZrF,. Preliminary values obtained for thissmaterial are 37, 54, 66, 74, 83, 115, and 152 mm Hg at 769, 786, 796, 805, 816, 828, and 844°C, respectively. A straight line through these values on a log P vs 1/T plot intersects the plot of vapor pressure of ZrF, ot 762°C. Since the primary phase at 75 mole % ZrF, in this system is ZrF,, the inter- section should correspond to the ligquidus tempera- ture for the mixture, which thermal-analysis data indicate to be 755°C, The difference between the two tempemtures is only slightly iarger 'rhcm the expected error of either method. It was discovered that there was a consistent temperature error in the vapor pressure data given in o previous report® for the 50 mole % NaF~50 mole % ZrF, mixture. Therefore the correct pres- sures ore Jower than those given before. The corrected equation for this composition is 7425 log P H = e e og P (mm Hg) = oK) +- 7.889 . FUSED-SALT HEAT TRANSFER H. W, Hoffman Reactor Experimental Engineering Division J. Lones Previous NaF-KF-LiF heot-tronsfer experiments made in systems in which no tube-wall deposits were formed were limited to rather low Reynolds numbers because of the temperature limitations inherent in the use of nickel tubes. It had been observed that molten fluorides could be contained in type 316 stainless steel for short periods of time without serious corrosion results or the forma- tion of the insoluble fluoride deposits, such as K 3CrF , that are found in Inconel systems. There- fore in order to get data at higher Reynolds numbers it was decided to conduct a heat-transfer experi- ment of short duration in type 316 stainiess steel tubes, The results of the experiment are shown in Fig. 7.3. As was to be expected when there was no wall deposit, the data fell on the curve for normal turbulent-flow for ordinary fluids, that is, the curve expressed by j = 0.023 Re=02, The 44, R. Nelson and R. W. Dayton, Progress Report for January 1954, BMI-902 {(Feb. 1954). 5W. H. Rodebush and A. L. Dixon, Phys. Rev. 26, B51 (1925). SR, E. Traber, Jr., R. E. Moore, and C. J. Barton, ANP Quar, Prog. Rep. Dec. 10, 1953, ORNL-1649, p 99. 101 ANP QUARTERLY PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 1416 0.040 - _ ] - . - . 0.009 —— e — - l e L_ }— - 0.008 |- — — i — : | . 0.007 L | | a NICKEL TUBE | ‘ # TYPE 316 STAINLESS STEEL TUBE 0.006 |——o s — ] T \ z 0.005 [—— L T - e : - g J 3 Re~ O ‘ = 0.004 ;i - .- | e . JR ‘1*%‘ ) . b * l T Lol | : 1 | ™~ 0.003 \ ‘ & | o i - INCONEL TUBE (WALL DEPQOSITS} Q O 0.002 - Afe 1 ){.& - F - L - — o . J ’ ’ T / iy 1 /,; 1 ! J | : 1 ‘ ‘ 1 1 . ,‘ J . i ‘ o004 L— —— ,W‘,J_‘,_,,J ,,,,, . [ J | e [ [ 103 2 5 10% 2 5 103 REYNOLDS NUMBER Fig. 7.3. Colburn j-Function vs Reynolds Number for Molten MaF.KF-Lif, previously obtained nickel-tube and Inconel-tube data are also shown in Fig. 7.3. Some preliminary determinations show the thermal conductivity of K,CrF ( to be about 0.13 Btu/hr.sq ft (°F/ft) for the temperature range 100 to 200°F. The salt was hydrostatically molded into the shape of a cylinder, approximately ]/2 in. in diameter and 4 in, in length. The density of the molded salt was 96% of the theoretical value, The cylinder was heated to a uniform temperature and then suddenly exposed to a cool air stream where it was transiently cooled. The thermal conductivity was calculated from the time-temperoture measurements obtained in this experiment, REACTOR HYDRODYNAMICS J. O. Bradfute L. D. Palmer G. M. Winn Reactor Experimental Engineering Division A Lucite model of the reflector-moderated reactor core has been fabricated, The fluid velocity dis- tribution in the voriable-cross-section flow annulus is to be determined by photographing tiny particles which will be dispersed throughout the flowing 7L. D. Palmer and G. M. Winn, A Feasibility Study of Flow Visualization Using a Phosphorescent Particle Method, ORNL CF-54-4-205 (Apr. 30, 1954). 102 water, An improved method of increasing the light output of the Strobolume (the flashing fight source) has been devised. The timing unit for the Strobolume has nearly been completed, An investigation was made of a technique’ for visual observation of the nature of fluid flow in tronsparent duct systems. It was found that the features which determined the average velocity profile, as well as the velocity fluctuations, under turbulent flow conditions could be observed, A beam of ultraviolet light was flashed through o region of the flowing fluid (containing suspended phosphorescent particles) for which the velocity profile was to be observed. The beam of phos- phorescent particles generated in the flowing fluvid by the wltraviolet light ‘was slowly distorted by the flow into a curve which exactly described the fluid velocity profile. It is believed that this technique can be used to visualize flew character- istics in reactor cores, HEAT-REMOVAL STUDY OF BSF REACTOR D. C. Homilton F. E. Lynch Reactor Experimental Engineering Division A study has been initiated for determining the heat-removal rate that can be provided by free con- vection in a bulk-shielding-facility type of reactor. The problem divides naturally inte two coses de- pending on whether or not nucleate boiling is per- mitted. When boiling is not permitted, the limiting fuel-plate temperature will be equal to the satura- tion temperature. If nucleate boiling is permitted, the exit temperature of the water may not exceed the saturation temperature. The analysis is somewhat similar to on cmoinis performed on a thermal convection harp.? There are two distinct regions in the flow circuit of the reactor, the reactor region and the jet or stack region. The reactor region presents a problem in flow and heat transfer between parallel plates with a known but not uniform wall-heat-flux dis- tribution. At lower reactor powers, the flow is superposed - free-and-forced viscous convection, At higher powers, the Reynolds modulus may approach the lower limit for isothermal turbulent flow; in either case, the heat-transfer and friction data are not available in the literature. The manner of mixing of the jet with the surrounding pool will determine the contribution of the jet to the flow through the reactor. An accurate consideration of the jet contribution will be difficult. HEAT TRANSFER IN CIRCULATING-FUEL REFLECTOR-MCDERATED REACTOR H. F. Poppendiek L. D. Palmer G. M. Winn Reactor Experimental Engineering Division - A heat-transfer analysis? was made for the case of forced convection between parallel plates of infinite extent, which are ducting fluids that contain uniform volume-heat sources, with heat transferred uniformly fo or from the fluids through the parallel plates. The laminar and turbulent flow data in terms of dimensionless wall-fluid temperature dif- ferences are plotied in Fig. 7.4 as a function of Reynolds moduli for the case of no wall-heat transfer. 1f it is desired to obtain the solution for the composite case, that is, a flow system with wall-heat transfer as well as volume-heat sources, 8p. C. Hamilton, F. E. Lynch, and L. D. Palmer, The Nature of the Flow of Ordinary Fluids in a Therma! Convection Harp, ORNL-14624 (Feb 23, 1954). H. F. Poppendiek and L. D. Palmer, Forced Con- vection Heat Transfer Between Parallel Plates and in Annuli with Volume Heot Sources Within the Ffurds, ORNL-1701 (May 11, 1954). PERIOD ENDING JUNE 10, 1954 the data given in Fig. 7.4 must be superposed on the known data for the case of wall-heat tronsfer with no volume-heat sources. It will be noted that the data shown in Fig. 7.4 pertain to liquid metals as well as to ordinary fluids and may also be used to estimate heat transfer in annulus systems such as the reflector-moderated reactor for which inner- to-outer radius ratios do not differ significantly from unity. A series of experiments was conducted recently to determine the influence of electrical-current flow on the hydrodynamic characteristics of a liquid flowing in o tube under both laminar and turbulent flow conditions. This problem arises in convection experiments in which the volume heat sources are generated electrically. No in- fluence of the electrical-current flow on the hydro- dynamic structure has yet been detected. HEAT TRANSFER IN NaOH-MODERATED CIRCULATING-FUEL REACTOR J. O. Bradfute M. F. Poppendiek Reactor Experimental Engineering Division A study has been initiated to determine the feasibility, of @ NaOH-moderated circulating-fuel reactor in terms of heat tronsfer and fluid flow, In such a reactor, the molten sodium hydroxide would be the moderator as well as the coolant necessary to reduce the high wall temperatures in the fuel region. Because of the corrosion difficulty that arises when NaOH is used with the typical structural materials of reactor systems that operate at temperatures above 1000°F, it is further neces- sary to cool the fuel-tube wall sufficiently (by forced circulation of the NaOH) to yield a NaOH- wall interface temperature of below 1000°F. There- fore it is necessary to know whether this cooling can be accomplished with NaOH with reasonable pumping powers and pressure drops. Some pre- fiminary heat- and momentum-transfer analyses have been made for the case in which fuel circulates through tubes that are force-cooled by molten NaOH which circulates in a direction parallel to that of the fuel; the number of tubes was varied in the analyses. The preliminary results suggest that the pumping powers and pressure drops are not prohibitively high. Further calculations on multitube, as well as annulus and channel, systems are being made. 103 ANP QUARTERLY PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 180A 10° 2 5 10® 2 5 104 2 5 105 2 5 1o} Fig. 7.4. Dimensionless Differences Between Wall ond Mixed-Mean Fluid Temperatures os Functions of Reynolds and Prandt] Moduli for o Paralle! Plate System (Walls Insulated). 104 PERIOD ENDING JUNE 10, 1954 8. RADIATION DAMAGE J, B, Trice Solid State Division A, J. Miller ANP Division The program of irradiating fused-salt fuels in the MTR continued with mixtures containing UF_- and UFA-»bearing fuels. The fuel-lnconel interface temperature is being maintained ot more nearly 1500°F than in previous tests with UF, fuels, A re~examination was made of a group of Inconel capsules from the earlier runs along with a study of the temperature control system used and the thermal histories. It was concluded that in mest capsules the interface temperature was in the region of 100°F higher than the desired 1500°F and that 2 to 4 mils of intergranular corrosion occurred. Excessive grain growth and deep penetration of the [nconel occurred only in a few capsules and those capsules had been heated to much higher tempera- fures. ' The first in-pile circulating fuel loop was in- stalled in a horizontal beam hole of the LLITR but difficulties developed before start-up of the reactor which caused the loop to be removed for inspection. Development continued on smaller loops for iin- sertion in LITR and MTR beryllium A-pieces. Creep-test equipment was revised to permit creep testing in the MTR, and development of in-pile stress-corrosion equipment continved, Metallo- graphic studies were made for Pratt & Whitney Aircratt Division on sandwich-type UO,~type 347 stainless steel fuel elements irradiated in the MTR, RADIATION STABILITY OF FLUORIDE FUELS G. W, Keilholtz M. T. Robinson J. G. Morgan W. R. Willis H. E. Rebertson W. E. Browning C. C. Webster M. F. Osbome Solid State Division As previously reported,! some of the Inconel capsules containing fluoride fuel which were irradiated in the MTR were found to have developed large inconel grains and deep intergranular cracks. Since such corrosion effects indicated excessive heating, a re-examination of the thermal conditions ]G. W. Keilholtz et al., ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 102, that had been imposed on the capsules during the irradiations was undertaken. Temperature control had been based on data from several thermocouples welded to the outside surface of the capsule and was aimed at maintaining this outside temperature at a level that would provide a 1500°F fuel-Inconel interface temperature, ' | In order to evaluate the accuracy of the control method, a series of tests was made to determine the relationship between the temperature of the outside surface of the capsule as indicated by a thermocouple to the actual wall temperature in an equivalent portion of the capsule that was unper- turbed by the presence of a thermocouple, The indicated temperature could have low values be- cause of the thermocouple junction being raised slightly above the surface of the capsule and because of conduction of heat by the cooling air stream from the very small amount of uninsulated exposed wire. A review was. also made of the chart temperature records of the experiments to detect runs in which the temperature had risen to experi- mentally undesirable magnitudes. A bench-test apparatus used to mock up the conditions of the MTR capsule tests is shown schematicaily in Fig. 8.1, The electrically heated test section which simulates the capsule consists of a lin, section of an actual Inconel fuel capsule. The annulus for cooling air corresponds to that used in the reactor, and the pyrex portion serves as a sight glass fo permit accurate temperature measurements with an optical pyrometer at a point 180 deg from the thermocouple being tested, Typical data from one of the 30 thermocouples tested are shown for three different airflow rates in Fig. 8,2, in which T is the unperturbed sQrfuce temperature determined with the pyrometer at any position along the length of the capsule. On each of the three curves the temperature indicated by the thermocouple 7, is shown at its contact po- sition. For airflow rates of 189, 220, and 250 fps, it can be seen that the thermocouple readings are fow by 75, 88, and 105°F. The maximum difference between T‘S and T[ for 30 thermocouples was 200°F 105 ANP QUARTERLY PROGRESS REPORT UNGLASSIFIED ORNL --LR-~DWG 515A :- ™ \\ ‘ J,/Y = ;\:“‘\\ \\\ \\ %y N - — ANNULUS O BODY TEST SECTION—--»—F-—- - THERMOCOUPLE A N I N - SIGHT GLASS TN 1 7 / o 2 ELECTRODE -t /" 7 7 Fig. 8.1, Thermocouple Test Apparatus, 106 5500932 _DWG. 233734 1700 ’ e .i.____. e !,,, e [ —_— : i \ CAPSULF CENTER | | ! LINE ] ] ENSOO 7 _ o ' T =1a85°F N ) : n - f;[_ 250 fps % ! 2201 ! = . I e pPs | W k_ 1500 ;¢—§=146|°F Ko : [ ENTRANGE ‘ v AR // HBY fps LT L / / ‘ e ! | / | ! : a0 oAl ] ¢ e g Png CAPSULE LENGTH (in) ig. 8.2, emperature Measurements on Fig. 8.2, Temperature M t MTR Capsule Mock-up, and the average difference at the medium airflow rates used in the reactor was T100°F, This error, together with several others of lesser magnitude, led to the conclusion that the tempera- ture-control system employed in the reactor experi- ments tended to produce fuel-lnconel interface temperatures averaging 1620°F and never higher than 1750°F at the hottest point of the capsule rather than the desired 1500°F., These findings are in agreement with calculations made by H. F. Poppendiek prior to the test experiments. Temperature record charts of the 13 MTR cap- sules irradiated so far in the MTR were examined. In every instance of deep intergranular corrosion and abnormally large grains, the temperature pattern for the fuel-lnconel interface was found to be either too high or too unsteady, For the cases in which the capsule temperature-time profile was within the limits originally set for the experiment, depth of attack (2 to 4 mils) and grain growth were not grossly different from those observed in bench tests at 1500°F, The pattern of corrosion was, however, still intergranular rather than that con- sisting of subsurface voids as in the case in out- of-pile tests. An additional aid to interpretation of results from the in-pile capsule tests was afforded by a study of the effect of high temperatures on the grain structure of Inconel. Samples of capsule material were subjected to temperatures of 1800 to 2350°F for various periods of time. Metallographic exami- nations of these samples are now being made. Preliminary results indicate that farge grains may be found in Inconel which has been heated for periods of time of less than 8 min duration at temperatures of the order of 2100°F, The Inconel capsules presently being irradiated in the MTR contain both UF,- and UF 4-bearing fuel mixtures. The recent mdlcahons in outsofe pile tests that the use of UF, might result in less corrosion suggested in-pile testing of UF -beurmg mixtures. Temperatures of the capsules are being carefully controlled so that the Inconel-fuel inter- face temperature will remain approximately ]SOOOF throughout each test. LITR FLUORIDE-FUEL LOOP W. E. Brundage M. T. Morgan C. D. Baumann A. S. Olson F. M, Blacksher W. W. Parkinson C. Ellis 0. Sisman R. M. Carroll Solid State Division The design of fluoride~fuel loops for use in the LITR horizontal-beam facility was described in previous reports.? The fabrication of components for three of these loops was completed. The first loop was inserted in the LITR but was removed before start-up of the reactor because of a leak in the fuel system. All fuel material was retained within the loop apparatus. Examination to dezter- mine the cause of the breakdown is in progress. Development continued on in-pile pumps :for loops to be operated in the horizontal~beam fa- cilities of the LITR and MTR (cf., Sec. 2, ““Experi- mental Reactor Engineering’’). Several models are being constructed. A smaller uranium investment and less dilution of fission products would be attained by. using such pumps in place of the present pumps which are operated at the reactor face. IN-PILE STRESS CORROSION AND CREEP W. W, Davis J. C, Wilson N. E. Hinkle J. C. Zukas Solid State Division Since intergranular corrosion of Inconel by molten fused salts of the type NaF:ZrF -UF, has been 2W. E. Brundage et al., ANP Quar. Prog. Rep. Moar. 10, 1954, ORNL-1692, p 105, _ PERIOD ENDING JUNE 10, 1954 observed in bench tests in which a strain was introduced into the Inconel, it is conceivable that thermal stress may be associated with the inter- granular type of attack seen in in-pile tests. To investigate this possibility, an in-pile stress-cor- rosion apparatys for the Inconel-fused salt systems was designed. Several prototypes of this apparatus are being bench tested. Creep tests of Inconel in the ORNL Graphne Reactor and in the LITR under conditions pertinent to ANP reactor designs have been reported previ- ously. Apparatus for tests in the higher flux of the MTR is being built. Much of the mechanical assembly is completed and the unit is almost ready for wiring., It is planned to use Bourdon tube extensometers rather than thermal-expansion (bi- metal) units since the former are in a more ad- vanced stage of evolution. . "REMOTE METALLOGRAPHY A. E. Richt W. Parsley E. Schwartz R. Ramsey M. J. Feldman Solid State Division Solid, sandwich-type, UO,-type 347 stainless steel fuel elements irradiated in the MTR: were submitted to the Laboratory by Pratt & Whitney Aircraft Division for metallographic examination, These studies, which were at first intended to yield information of interest only to the supercriti- cal-water~cycle aircraft reactor, are also related to certain power-package reactor designs because of the similarity between the two types of fuel elements both in materials and in required operating conditions. Both reactors have operating pressures of the order of 5000 psi, temperatures in the range of 300 to 700°F, and high power generation rates. Both fuel element prototypes are made as solid sandwich-type elements clad with stainless steel or Inconel. A typical Pratt & Whitney fuel element test is run with the element immersed in water in a sealed capsule at one of a series of fixed temperatures ranging from 100 to 700°F and with pressures of 3000 to 5000 psi. The power density of the fuel runs was as high as several thousand watts per cubic centimeter in the MTR., Results from the five capsules examined so far provide information about the effects of temperature and irradiation on core density changes, bonding of core to cladding, core hardness, and corrosion of the stainless steel 107 ANP QUARTERLY PROGRESS REPORT cladding., No measurable core density changes have been found and general corrosive attack on the cladding is absent, Hardness measurements taken on the core indicate a dependence between particle size and hardness change. Cores with particles less than 3 p in diameter show greater hardness changes than do those in the 15- to 44-u range. In general, it can be said that hardness 108 changes, expressedin DPH units, as great as ratios of 2 to 3 occur for iradiation periods of several hundreds of hours and for test temperatures as high as 700°F. Cracking of the core in bending tests {(1.5-in. radius bend) was found only in the less than 3-p core somples. It is expected that this work will be completed during the next several months, PERIOD ENDING JUNE 10, 1954 9. ANALYTICAL STUDIES OF REACTOR MATERIALS C. D. Susano Analytical Chemistry Division J. M. Warde Metallurgy Division R. Baldock Stable Isotope Research and Production Division Research and development in analytical chemis- try for the ANP Project was concentrated on the problem of separating UF, from UF, in NaZrF - base fuels and fuel solvents. Methods were in- vestigated for the conversion of the fluorides to chloride salts and the simultaneous extraction of the chlorides into nonagqueous solutions. Boron trichloride in a solution of dioxane appeared to be promising for this purpose, since UCI, is soluble in dioxane, whereas UCl, is essentially insoluble. The separatioen is also feasible in the presence of chromium chlorides. Chromium dichloride is soluble in dioxane in contrast to CrCly, which is insoluble. However, the conversion reaction was found to proceed slowly. A method was devised for the determination of oxygen in NaZrF .-base fuels and fuel solvents in which the oxides react with' anhydrous hydrogen fluoride to produce water, quantitatively; the water materially increases the conductivity of the hydro- fluoric acid. It has been calculated that the conductivity of anhydrous hydrogen fluoride is doubled by the addition of 1 ppm of water, In the study of the oxidation of trivalent uranium by hexavalent uranium it was shown that the competi- tive oxidation with hydrogen ion could not be completely eliminated. Stability tests of trivalent uranium in hydrochloric acid solutions were con- ducted. Although evaluation of the data is not complete, potentiometric measurements indicated that the transition in color from purple to green, which is considered to be evidence of oxidation, is not accompanied by corresponding changes in potential. A 1% aqueous solution of the tetra- sodium salt of ethylenediaminetetraacetric ocid effectively removed a film of calcium metasilicate from Inconel. No evidence of corrosion was found by chemical tests. The sulfur content of natural gas was determined to be 0.5 mg/cu ft of gas. Petrographic examination of NaF-ZrF -UF, fuels reduced with metallic zirconium, with metallic uranium, or with hydrogen has shown that the pre- dominant phase in the reduction complexes ob- tained is always a solid solution of reduced U** in NagZroF .. In the RbF-UF, system, there appears to be a compound, 3 RbF-UF3, which is tan and isotropic, with an 7= 1,438, The NaF-UF, system contains a compound of unknown compo- sition which is lavender and uniaxially positive, with refractive indices of approximately 1.552, Examinations of melts in the NaF-CrF, system were made, and four compounds were observed. ANALYTICAL CHEMISTRY OF REACTOR MATERIALS J. C, White Analytical Chemistry Division Oxidation States of Chromium ond Uranium in NquFs-Busie Fuels A. S. Meyer, Jr, D. L. Manning W. J. Ross Analytical Chemistry Division In extensive studies, covered in previous re- ports,! no solvent has been discovered that will selectively dissolve UF, or UF, from NaZrf .- base fuels. Certain organic solvents preferentially dissolved UCH, trom UCI,, but the equi!ibripm of the reaction UIV + Cr” @Ulll + Cr!ll could conceivably be altered during the conversion of the metallic fuel components by the method of fusing the sample with NaAICl . Therefore, an attempt is being made to dissolve the UCI, info an organic solvent during the conversion of the fuel to the chloride salts. ' Dioxane, which has been reported to dissolve UCI, from U(:l3,l was selected as the most promising solvent. |t may offer an additional ad- vantage in that aliphatic ethers are reportgd2 fo V). C. White et ol., ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL-1692, p 107 ff, 2N. Y. Sidgwick, The Chemical Elements and Their Compounds, VYol. Il, p 1024, Oxford University Press, Oxford, 1950. 109 ANP QUARTERLY PROGRESS REPORT dissolve CrCl, from CrCl, to yield stable solutions which are void of reducing power. Thus, if the chlorination of the sample could be carried out in the presence of dioxane, UCi, and CrCl, would be dissolved, while UCl, and CrCl, would remain in the residue, It is hoped that the rate of reaction between the solid chlorides is negligibly slow at room temperature. Preliminary tests were carried out by stirring 25-mg samples of UF,, UF ,, NaZrF, and NaZrF,- UF, in solutions of aluminum chloride in dioxane for periods of 4 to 24 hr, Visual observations and qualitative tests showed no evidence of dissolution of any of these samples. Similar negative results were obtained when the tests were repeated with solutions of aluminum chloride in ethyl ether, tetrahydrofuran, and benzene, respectively. Thermodynamic calculations indicated that BCI is superior to AICI, as the reagent for the meta- thetical chlorination of the fluorides of wranium, chromium, and zirconium. In qualitative tests, 25- mg samples of UF, and NaZrF .-UF , fuel were not completely dissolved after 16 hr of stirring in a solution of dioxane that had been saturated with BCl,. When UF, was stirred with liquid BCl, at 0°C, there was no apparent dissolution of the sample. Analysis of the residue for chloride indi- cated that the reaction 4UF, + 3BCl, — 4UCI, + 3BF, was only about 20% complete. Since the free- energy change of this reaction is -110 kcal, it appears that o slow reaction rate is responsible for the incomplete dissolution of the samples. Attempts were made to increase the rate of reaction by treating the UF, and the fuel samples with BBr, at reflux temperatures of 91°C. Quantitative results of these experiments are not yet available, but portions of the unreacted samples were visible after refluxing for 3 hr, The reaction appeared to be accelerated by addition of AICI; to the BBr,. When 60 mg of UF, was stirred at room tempera- ture for 6 hr with 60 ml of dioxane, which had been saturated with BC|3, about 25% of the salt dis- solved to form a green solution. The concentration of uranium in solution did not increase significantly when aluminum chloride was added and the stirring was continued for 90 min. No tests of the solu- bilities of CrF, and CrF, have yet been carried out. 110 Since this method is the only approach which appears to offer possibilities of the separation of tetravalent uranium from a fuel containing another element for which the oxidation potential is near that of the U3 —> U4 couple, efforts will be made to accelerate the reaction to rates suitable for analytical applications, Because ethereal so- lutions of BCI, are decomposed by heating, it will probably be necessary to find a catalyst for the chlorination reaction. Experiments are also being carried out fo determine whether UCI, may be sublimed by heating the fuels in an atmosphere of BC|3. Determination of Oxygen in NaZrF ;-Base Fuels A. S. Meyer, Jr. J. M, Peele Analytical Chemistry Division Studies have been made to determine whether microgram quantities of oxygen in NaZrF -base fuels can be determined by measuring the water generated by the reaction of the mefallic oxides with anhydrous hydrofluoric acid. Calculotions of the free energies for the reactions of the type M,0, + 6HF (liquid) ~> 2MF, + 3H,0 (liquid) have been carried out for the oxides which might be present in NaZrF _-base fuels, These calcu- lations indicate that the reaction is quantitative for the following oxides: ZrOz, CrZO Fe203, UQ,, NiO, and Na,0. The conductivity of hydrogen fluoride is known 3' to be very sensitive to traces of water. Literature values?® for the equivalent conductivity of water in hydrogen fluoride were used for calculations which showed that the conductivity of anhydrous hydroflueric acid is doubled by the addition of 1 ppm of water, An apparatus is being constructed with which it will be possible to measure the changes in conduc- tivity of hydrofluoric acid that result from the water produced by the fluorination reaction of oxides in the fuel samples, The apparatus in- cludes a drying still, a reoction vessel, and a fluorothene conductivity cell. Provisions can be made for carrying out the reaction with liquid hydrotluoric acid at room temperature or with the gaseous reagent at elevated temperatures. 3). H. Simons, Fluorine Chemistry, Vol. 1, p 240, Academic Press, New York, 1950. Oxidation of Trivalent Uranium by Hexavalent Uranium A. S. Meyer, Jr. D. L. Manning Analytical Chemistry Division Experiments were continued on the oxidmion:of trivalent uranium by the urc:nyl ion UO to the reaction _ 203" + Uo,"™ + 4HT —3U%" 4+ 2H.0 . In previous tests! the extent of the reaction was determined by measuring polarographically the concentration of the unreacted uranyl ion. This measurement can also be made by titrating the U4* formed with the ceric ion Ce?®, Several tests were conducted in which known amounts of ucl, were added to 100 ml of Q.01 N uranyl acetate under varying conditions of acidity and tempera- ture. After dissolution of the UCI,;, the concen- tration of the U4 was determined by titrating with 0.1 N Ce**. The end point was obtained potentio- metrically. The solutions were swept thoroughly with helium and titrated under an inert-gas blanket, Quantitative oxidation with the uranyl not achieved in any of the tests. occordin_g ion was indeed, when solutions more acidic than 1 N were used as sol« vent media, evolution of hydrogen was observed. The reaction is favored, however, when the uranium is added as the chloride instead of as the fluoride. This is probably due to the greater dissolution rate for UCI, in acid (several minutes) as compared with the slow dissolution rate (several hours) for UF, in this medium, Efforts to eliminate the competitive reaction for the oxidation of trivalent uranium in aqueous so- {utions, _ 3% ¢ HY —> U4t 4 UH, have to date been unsuccessful. Future work will involve investigation of possible systems for the oxidation by uranyl ion. nonaqueocus Stability of Trivalent Uranium in Hydrochlorlc Acid Solutions A. S, Meyer, Jr. W. J. Ross D. [.. Manning Analytical Chemistry Division Katz and Rubinowitch4 stated that the stability of UCl, is greater in hydrochloric acid than in 4J. J. Katz and E. Rabinowitch, The Chemrsfry of Uranium, p 458, McGraw-Hill, New York 1951, PERIOD ENDING JUNE 10, 1954 ““characteristic’ purple color An investigation.of the sta- water and reported a for the trivalent ion. bility of trivalent uranium in hydrochloric acid media was made to ascertain the feasibility of titrating U2, The potential of a solution of ucl, in 1.0 N HCI at 0°C was f-ollowed with a plahnum calomel electrode system. The initial potential upon dissolution was +250 mv vs S.C.E., and the potential decreased only 15 mv in 10 min. No evidence of a purple color was observed ini this experiment; the solution turned green upon dis- solution. ' The experiments were repeated with concen- trated hydrochloric acid at both 0 and -20°C., In both cases, instantaneous formation of a .purple- colored solution occurred, and the purple color remained for 10 to 20 min before it changed to green, The purple color remained longest in the solution tested at the lower temperature., The initial potential in all tests was of the order of +150 mv, and the potential did not change appreci- ably during the transition in color from purple to green. The potential was essentially constant ot +150 mv for 30 min. The constancy of the potential of these solutions contradicts the usually accepted hypothesis that the chonge in color from purple to green represents oxidation to the U4" state. Some quantitative data were obtained on the concen- trations of tri- and tetravalent uranium in these solutions by titration with ceric ion. In 0.1 N HCI solutions, two distinct breaks in the potential curve were noted, while only one break was ob- served in concentrated acid solutions. The data have not yet been thoroughly evaluated. The anomaly of the change in color of the solutions without an accompanying chcnge in potential will be mveshgated Removal of Film from Inconel Tubing J. C, White Analytical Chemistry Division The tetrasodium salt of ethylenediaminetetra- acetic acid was tested as a possible reagent for removing ‘a film, probably calcium metasilicate, which was found to be deposited upon the walls of Inconel tubing following a wash with a detergent Conklene. A 1% solution of the reagent in water, heated to 75 to 80°C, effectively removed the film rom a specimen of Inconel in a few minutes, The metal had a bright finish after it was rinsed with water. Chemical analysis of the reagent solution 11 ANP QUARTERLY PROGRESS REPORTY after contact for 60 hr at 75°C with Incone!l showed less than 1 ppm of chromium and nickel. Determination of Sulfur in Natural Gas G. Goldberg Analytical Chemistry Division A determination was made of the sulfur content of the natural fuel gas received at the Y-12 Plant because it is planned to use natural gas as a source of heat in future corrosion tests. The Referee® method was used in which the gas is burned in an ammonia-rich atmosphere and the condensable gases are collected in a tower filled with glass balls and ammonium carbonate, The reaction is SO, + 2NH; + Hzo—%,(NH.fl)zSO3 . The solution of ammonium sulfite is oxidized to sulfate, and the suifur is determined as barium sulfate., The sulfur content was found to be 0.5 mg/cu ft of gas, which is equivalent to 0.8 grain per 100 cu ft. This concentration is essentially the same as the sulfur concentration (1 grain of sulfur per 100 cu ft) of the propane which is re- ceived at the Y-12 Plant, PETROGRAPHIC INVESTIGATIONS OF FLUORIDE FUELS G. D. White T. N. MeVay, Consultant Metallurgy Division With the shift in emphasis to reduced fuels, more effort was devoted to the determination of the optical properties of compounds containing UF,. In NaF-ZrF -UF ; fuels reduced with metallic zir- conium, with metallic uranium, or with hydrogen, the predominant phase in the reduction complexes obtained is always a solid solution of reduced U4* in NaqzraF“. from greenish-yellow, to greenish-tan, to a brilliant yellow. The optic figure remains the same as that for unreduced solid solution, but there is a definite lowering of the refractive indices. Evidently the amount of optic change in the crystal depends upon the amount of reduction. In addition to the reduced solid solution, minor amounts of UF,.2ZrF ,, which is orange, uniaxially positive, and has refractive indices of approximately 1.55, and an olive-drab phase, which is pleochroic with refractive indices The color of this complex varies SL. M. Dennis and M. L. Nichols, Gas Analysis, p 359, MacMillan, New York, 1929. 112 of about 1.54, are found. The composition of the olive-drab phase is unknown. Sometimes a trace of an isotropic yellow phase with n = 1.44 |s present., In the RbF-UF, system, there appears to be a compound, 3RbF.UF ., which is tan and isofropic, with an n = 1,438, The NaF-UF; system contains a compound which is lavender and uniaxially posi- tive, with refractive indices of approximately 1.552. The composition of this compound is not known. Examinations of melts in the NaF-CrF, system were made and four compounds were observed. The compound Na,Crf, was previously reported as isotropic and green, with n == 1.411. In composi- tions with high CrF3 content, two phases were present. The predominant phase was pleochroic, with the fast direction yellow-green and the slow direction blue-green, uniaxially positive, with 0 = 1.452 and E = 1.460, and polysynthetically twinned. A second phase that was present in smaller amounts was green and biaxially negative, with o small optic angle and high dispersion. |t had a moderate birefringence, and its refractive indices were near 1.52. A third phase was present in melts with compositions near 50% NaF--50% CrF,. It was yellow-green and uniaxially negative, with O = 1.444, and it had an estimated birefringence of 0.008. The optical properties of the several compounds that hove been synthesized were compiled and published in an ORNL report.® SUMMARY OF SERVICE ANALYSES J. C. White W. F. Vaughan C. R. Williams Analytical Chemistry Division Analyses of samples from corrosion tests of the NaZrF ;-base fuel continued to be the major portion of the work in the service laboratory, Samples of fuel preparations containing mixtures of UF, and UF, were received. In addition to the determi- nations for the corrosion products iron, nickel, and chromium, the concentrations of UF, and UF , were determined, The method’ for hydrogen evolution 6T. N. McVay and G. D. White, The Optical Properties of Some Inorganic Fluoride and Chloride Compounds, ORNL-1712 (May 5, 1954). ’D. L. Manning, W. K. Miller, and R. Rowan, Jr., Methods of Determination of Uranium Trifluoride, ORNL - 1279 (Apr. 25, 1952). has been used for the determination of UF,, and following the determination of total uranium, the content of UF4 is calculated from the difference in concentration between total uranium ond UF3. Samples from corrosion tests of prospective fufure PERIOD ENDING JUNE 10, 1954 fuels which consist of fluorides of sodium, lithium, and potassium or rubidium were also analyzed. During this quarter the laboratory received 1163 samples and reported 1060 which involved 9357 determinations (Table 9.1). ' TABLE 9.1. SUMMARY OF SERVICE ANALYSES REPORTED NUMBER OF SAMPLES NUMBER OF _DETERM!NATIONS Reactor Chemistry 610 4435 Experimental Engineering : 370 4116 Fuel Production - 64 671 Miscelldneous 16 135 1060 9357 113 Part Il SHIELDING RESEARCH 10. LID TANK FACILITY C. L. Storrs "GE-ANP Jo M, Miller Physics Division - J. B, Dee Pratt & Whitney Aircraft Division G. T. Chapman D. K. Trubey The Lid Tank Facility (LTF) has been used primarily for studying special attenuation problems which arise in aircraft shield design. One of these studies which is of interest for crew-shield design has been the penetration of slant-incident neutrons through a hydrogenous shield. Inanother study, comparison was obtained of secondary (neutron-induced) gamma-ray produchon in lead cnd in bismuth, SLANT PENETRATIONS OF NEUTRONS THROUGH WATER G. T. Chapfian Physics Division The experiments for determining the slant pene- tration of fission neutrons through water have been continued at the LTF. In o previous experiment’ the neutron flux at the end of the duct was low and it was possible to make measurements with only small amounts of water between the detector and the duct., For the present work, a new duct, larger in diameter and shorter in length (Fig, 10.1), was used, and the flux was increased to the extent that measurements could be made 45 cm from the end of the duct. The distribution of the fast-neutron dose around the end of the duct is plotted in Fig. 10.2, The dose was integrated along planes at different slant thicknesses (¢ ) of water and at angles (a) of 0, 30, and 60 deg to the incident flux. The results are given in Fig. 10,3, Owing to the large size of the source, the distance to the integration plane is not well defined at other than 0 deg, and therefore the relative magnitudes of the curves are in some doubt; however, the slopes are probably significant. The observed relaxation length for the 0- and 30-deg measurements was 7.2 cm. For the incident angle of 60 deg, the relaxation length increased from 7.8 to 12,5 cm as the slant thickness in- creased. - This increase in relaxation length is presumably due to the *‘short-circuiting’’ effect, that is, the scattering of neutrons in a direction which permits them to penetrate the shield along a nearly normal path rather than along o slant path, YC. L. Storrs et al., ANP Quar. Prog. Rep. Dec 10, 1953, ORNL 1649, p 120. ORNL~LR-DWG 1045 ¢ 6 D_UI\ST SOURCE \ \ \ ! e . __-PLANE OF f >\ e INTEGRATION ~ "»\-141 aQ° £\ I~ ] \/f’ % %G T ; \ /“* a,-"'\ : ; / 90-cm DUCT 7. 30-cm DIA -~ | 87.9cm \ L W y | |-—--"'28—in. DIA SQURCE "" Duct Arrangement for Slab Penetration Fig. 10,1, Experiment. 117 ANP QUARTERLY PROGRESS REPORT orn IR ., 135 | T T oo . e 1T mMrep/hr ) @Ok ,i ——=ny P R — ' - —— i e ! i ; ; : IO-‘ mrep/hr * 176 —— - \ . . i - z i 5 =20 [ Ly o o § s ] 1 \ : = 5 N % 106 |- - ,1,” e e Y SN N S 2 ; i w /'- N S mrspsar | \\ i oo !-- - e b -\‘K_\\_ S N \ 1G mrep/hr - ! \ \ | I - 4, . I : ":‘10 "3 mrepsn- . \\ : \\ o TSN NN ‘ ‘~ NN N 90 ~— ,,,,,..,.,.,.‘,,,,,,,,.__,__,,A_u_,,,,‘,_.,.__\.‘_ L — L I 60 55 50 45 40 35 30, 5 20 15 10 5 Q \ 5 \ L ¥.DISTANCE FROM SQURCE CENTER LAvE (cm) ; | " 7.3-cm DA X Y A 90-~em DUCT Fig. 10.2. Fast-Neutron Dose Distribution Be- yond End of Duct, DOSE UNITS x cos 4; SLANT THICKNESS {cm) Fig. 10.3. Slant Penetration of Fast Neutrons in Water. SECONDARY GAMMA-RAY STUDY D. K. Trubey Physics Division A study of secondary gamma-ray production in materials considered for use as gomma-ray shields has been undertaken ot the LTF., Measurements have been made behind 6 in. of lead or bismuth placed in borated water (1.1 wt % boron) at various 118 distances from the fission source plate (Fig. 10.4). The 6-in. thicknesses of metal amounted to 145 g/sq cm for the bismuth and 161 g/sq cm for the lead. The borated woter was contagined in a 36-in.-long aluminum tank placed adjacent to the source, [he detector used primarily was o l-in. anthracene scintillation crystal and o photo- multiplier tube, The gamma-dose measurements taken in the plain water behind the borated-water tank in which the lead or bismuth slabs were placed ot various distances (d) from the source are shown in Figs, 10,5 and 10.6. It can be noted that the dose be- comes less sensitive to slab position as & in- creases. In both Figs. 10.5 and 10.6, curves showing the gamma dose in plain water in the Lid Tank and in plain water behind the boroted- water tank have been added for comparison. The slopes of these iwo curves differ because the 2,2-Mev hydrogen-capture gammas produced in the plain water are attenuated more rapidly than the harder-source gammas predominant in the water behind the borated-water tank, In Fig. 10,5 the effects of lead ploced 3.2 ond 8.4 cm from the source in plain water are also shown, A comparison of the dose measurements behind lead and bismuth as a function of the position of the metal, with the detector placed 120 cm from the source, is shown in Fig, 10,7, Measurements ot any other detector position would be similar. The difference in the two curves for 4 = 47.6 cm is ORNL-LR-DWG 933 ‘7 ,,,,,,,,, o _ 1/ ALUMINUM TANK METAL SLABS <% | . S T TR ST e § A NN\ \ H,0 — & i BORATED Hy0 ! 2 (14% B) | ., | — 2= 35 cm - . - Fig. 10.4. Arrangement for Measuring Secondary Gamma-Ray Production in Materials Considered for Use as Gamma-Ray Shields. 2 ‘ ORNL -flkDWG 1004 GAMMA-RAY DOSE {(mr/hr} N 11% BORATED H,0, TANK AGAINST SOURCE | 90 100 110 120 130 140 150 180 z, DISTANCE FROM SOURCE TO DETECTOR (cm) Fig. 10.5. Effect of Lead ofi Gamma Dose as o Function of Distance in Waoter from Fission Source. presumably due to the difference in the densities of the two metals. The fact that the curves become flat for d > 47 cm indicates that the slabs placed at greater distances from the source were in such a small neutron flux that the secondary gamma production had become almost negligible, The gamma-ray dose behind the slabs at d = 47.6 cmiis therefore considered to be coused by source gamma rays, including those originating in the water near the source becouse of hydrogen capture of thermal neutrons. The source dose can be subtracted from the dose behind thé slabs at other positions to determine the secondary gamma-ray dose at a given distance from the source. The resulting secondary doses for lead and bismuth are given in Figs. 10.8 and 10,9, respectively, as a function of distance from the front face of the slabs, since it was considered that the origin of these gamma rays was associated with the position of the slabs, PERIOD ENDING JUNE 10, 1954 GAMMA-RAY DOSE (mr/hr) 90 100 MO 120 130 140 150 150 2, DISTANCE FROM SOURCE TO DETECTOR (cm) Fig. 10,6. Effect of Bismuth on Gamma Dose as a Function of Distance in Water from Fission Source. ' ' A comparison of the doses from secondary gamma- ray production in lead and in bismuth is presented in Fig, 10.10, in which the dose at 100 cm from the front face of the metal is plotted as a function of the metal position. The experimental accuracy does not justify concluding that the slopes are different., Normalized thermal-neutron and fast- neutron curves (measurements made in plain water) have been included in Fig. 10.10 to show that the secondary-gamma dose curve has a slope charac- teristic of neutron curves. | From this experiment little difference between lead and bismuth secondary gamma-ray production The data are to be extended in the near future with the hope that it will be possible to obtain a more complete understanding of fthe phenomena, ' can be seen, 119 ANP QUARTERLY PROGRESS REPORT ORNL—LR—COWG 930 10 L T - . _ _ 1] o/ — ] | b o] I { i ! | S e - i — e i ij — § ;6 in. OF Bi i — - — — - GaMMA - RAY DO3SE (mr/hr) | \ FOUR 1'/2-in. Bi OR Pb SLABS IN 36-in. TANK OF 1.1% BORATED H,O; TANK AGAINST SOURGE 2 - e Y B _ e w2l ) 10 20 20 40 50 60 d, DISTANGE FROM SOURGE TO FRONT FACE OF METAL (cm) Fig. 10.7. Gomma-Ray Dose 120 cm from Source as a Function of Position of Metal., DRNL!E—DWG 9321 2 W07 e T : - T T T j ol B e | L----—_- Lol p ‘ - FOUR 1V2~in. Bi SLABS IN 36-in. TANK OF 11 [ | BORATED HZO; TANK AGAINST SQURLCE. 4‘ 2 e ’ ‘ e i — — L 10 A 5| = = £ oLl w w o &y — - o o L % 5: g : = 3 2 - 1o 10_2 . . i ‘ L i 70 a0 HO 130 150 < — d. DISTANCE FROM FRONT FACE OF METAL TO DETECTOR (ecm) Fig. 10.9. Dose from Secondary Gomma-Rays (Source Gamma-Roys Substracted) as o Function of Distance from Bismuth Slabs, 120 ORNL-LR-DWG 932 2 ey 1o T LTI T B 5 - C e e e Lo - ] FOUR 1Y%-~in. Pb SLLABS IN 3&-in. TANK N i OF 1.1% BORATED H,0; TANK AGAINST SQURCE | 2 b -y §T e ' . e L —t GAMMA-RAY DOSE (mr/hr) | | drr—e e en oL 70 80 90 100 110 120 130 140 150 160 z— 0, DISTANCE FROM FRONT FACE OF MZTAL TC DETECTCR {om) Fig. 10.8, Dese from Secondary Gamma-Rays (Source Gomma-Rays Substracted) as o Function of Distance from Lead Slabs. ORNL—-LR—DWG 923 10 —_— T 4 = T = 7~ FOUR i%2-n. Bi OR Pb SLABS IN 36-in — 5 TANK OF (4% BORATED H0; TANK \\fi AGAINST SOURCE. - \ i e T \””’7" ”7:7' T " . 77'7!' T - T 2 - \\ ————— — THERMAL -~ NEUTRON FLUX N -1 PLAIN H0, niy X 107 | T ~—F— FAST-NEUTRON DOSE IN — 7777 e - PLAIN H0, mrep/hr X 107 -——— 6 e RI NSUSSRRESEL [ S S — | i, AN ,+ i e ] — - GAMMA DOSE BEHIND 4 —— 6 in. Pb, mr/hr ,,,77‘, B ) _— [ 107 |- } ~—GAMMA DOSE BEHIND - & in B mesbr T 5 L — \ N i - ; i 10 1 ek o 10 20 30 40 50 60 o, DISTANCE FROM SOURCE TQ FRONT FACE OF METAL {cm) Fig, 10,10, Dose from Secondary Gamma«Rays (Source Gamma-Rays Substracted) 100 cm from Front Face of Meta!l as o Fuaction of Position of Metal. PERIOD ENDING JUNE 10, 1954 11, BULK SHIELDING FACILITY . G. Cochran E stabrook . Flynn . Haydon Henry ARE-O xTvPET ™ F. C. Maienschein H. E. Hunderford E. B. Johnson T. A. Love R. W, Peelle Physics Division The Bulk Shielding Facilify (BSF) reactor has been used for determining the power and power distribution of loadings similar to those used in the new Tower Shielding Facility (TSF) reactor. The search for a better fast-neutron spectrometer continues, but no substantial progress has been made. Calculations of gamma-ray dose in a standard divided-shielddesign have been extended to include crew-shield penetration. ' GAMMA-RAY AIR-SCATTERING CALCULATIONS F. C. Maienschein Physics Division Calculations of the dose received from air- scattered gamma rays at the outside edge of the crew compartment of an aircraft divided shield were reported previously,! These calculations have now been extended to determine the dose inside the crew compartment. was used for calculations of the gamma-ray pene- tration through the crew shield, and in this case the ANP-53 crew shield® was replaced by a simple cylinder of about the same dimensions. | A further calculation was made of the penetration of the direct {unscattered) gamma radiation reach- ing the crew position through the reor of the crew shield. The buildup facters used were those reported by Fano.? Results of the air-scattering calculation indicate that although the skew-beam flux at the outside of the crew shield is extremely small compared with 1), L. Meem et al., ANP Quor. Prog. Rep. June 10, 1953, ORNL-1556, p ]09, F. Bly and F. C. Maienschein, A “Calculation of the Gemma Rudiation Reaching the ANP-53 Crew Shield, ORNL CF-53-5-117 (May 23, 1953). 2H. Goldstein, Reactor Handbook, Vol. 1, p 831, Tech- nical Information Service, U, 5. Atomic Energy Com- mlssmn, 1953. Reporf of the Shielding Board for the Aircraft Nucleor Propulsion Program, ANP-53 (Oct. 16, 1950). 41J. Eano, Nucleonics 11, 55 (1953). The NDA method? the radial-beam flux, the dose inside the crew com- partment from the skew-beam flux is comparable to the total radial-beam dose. This result was ex- pected, since the skew-beam gomma rays, having scattered through smaller angles on the averoge than the radial-beam gomma rays, are of higher energy and thus penetrate the crew shield more readily. The total dose inside the crew compartment from air-scattered gamma rays, as calculated by this method, is somewhat lower than that calculated by the ANP-53 method? (Table 11.1); however, the agreement is good if it is considered that for the ANP-53 calculations it was assumed that all the gamma rays were of 3-Mev energy and that all were radially emitted from the reactor shield. The cal- culations would be in poorer agreement if the correction {unknown) for difference in leakage ratios between the ftwo reactors were applied. For the direct beam, the large increase shown by this calculation in comparison with the value obtained by the ANP-53 calculation is coused primarily by the gamma radiation that escapes around the edges of the shadow shield. The gamma-ray spectral medsurements at the BSF show that a large number of low-energy (< 2-Mev) gamma rays emerge from the reactor shield in such o direction that they may strike the crew shield without scattering. Unfortunately, the spectrometer background at such positions was so large that little value can be placed on the absolute magni- tude of the dose. In general, for all results from these calculations, the shapes of the curves ob- tained are considerably more reliable thon the absolute magnitudes. The contributions to the fotal scattered dose of various regions outside the reactor and crew shields are shown in Figs. 11,1 and 11,2, The effect of the shadow shield is shown in Fig. 11.2, which is shaded proportionally to the dose contribution from each region. Furthermore it is shown in that figure 121 ORNL-LR—DWG 974 0.0033 0959 06975 | G.3257 | Lo.0z2a .52 L0.0531 —— . . “ ~SEZ INSERT L - — T i T (WULTIPLY ALL VALUES 8Y 107) UNOERLINED VALUES ARE SKEW BEAM DATA FOR ‘a AND £ AS SHOWN. GEOMETRY APPLES TO RADIAL BEAM DATA ONLY. VALUES ARE N hr 180° REACTOR CREW COMPARTMENT lting from Air Scattering from Individual Cells — Radial and Skew Emission. Fig. 11.1. Gamma-Ray Dose Inside Crew Shield Resu 123 TABLE 11,1, GAMMA-RAY FLUX AND DOSE AT THE CREW COMPARTMENT FLUX OUTSIDE CREW DOSE INSIDE CREW ME THOD COMPARTMENT, &/P COMPARTMENT, D {(photons/sec/ watt) {e/hr/ watt) Radial scattered 4.5 0.36 = 10—9 Skew scattered 0.016 0.19 x 10~7 Direct 15.0 130 % 10~ ANP-53 scattered* ' ' 1.2 x 10~ ANP.53 direct* | 1.2 x 10~ Skyshine** 114 x 10~° *No correction made for reactor leakage. **H, E. Hungerford, The Skyshine Experiments at the Bulk Shielding Facility, ORNL-1611 (to be issed). e ORNL~ LR~ DWG 1024 i oooos W 88;920 B 0 oo -10 " 7777 00001 TO [ LM*_W#J(DOOOO EZZ/A/ 0p00s & a ’ B REACTOR CREW COMPARTMENT VALUES ARE N T (MULTIPLY ALL VALUES BY T Fig. 11.2, Gamma-Ray Dose Inside Crew Shield from Air Scattering. 125 ANP QGUARTERLY PROGRESS REPORT that air-scattering from regions far from the aircraft makes an important contribution to the dose in the crew compartment. In all cases, only the penetra- tion through the side wall of the crew shield haos been given, since the front- and rear-wall contri- butions were negligible. Complete calculations were carried out for the front wall for both radial and skew emission in order to prove that these contributions were actually small. The dependence of the scattered dose (for all angles) upon the prescatter energy is shown in Fig. 11.3, and the dependence on the postscatter energy in Fig. 11.4. It is interesting to note that the average energy for the skew-emission dose is higher than that for the radial-emission dose. The scattered dose is presented as o function of the source angle in Fig. 11.5 and as o function of ORN[MG 2G4 )6 r hr -watt a g~ SCATTERED GAMMA-RAY DOSE ( w 0 20 40 6.0 8.0 00 120 £, PRESGATTER GAMMA-RAY ENERGY [Mev) Fig. 11.3, Scottered Gomma-Ray Dose as a Function of Prescatter Energy. 126 the receiver angle in Fig. 11.6. The distribution around the reactor shield (Fig. 11.5) clearly shows the peak in the forward direction which is due to the relatively favorable geometry for radiation starting forward, After a drop of a factor of about 300 out to 55 deg, the dose contribution rises rapidly of the edge of the shadow shield and again falls off because of geometry. It thus appears obvious that the shadow shield should extend to larger angles. It should, of course, be thinner with increasing values of a. It also appears that the ANP-53 crew shield was considercbly overshielded in the rear. This was recognized at the time it was designed, ORNL--LR—DWG BES - RADIAL EMISSION D oeeme- GKEW BEAM | hr ) [ SCATTERED GAMMA-RAY DOSE ( 3 & o 1072 | 0~ L 0 1.0 2.0 3.0 4.0 £, POSTSCATTER GAMMA-RAY ENERGY (Mev) Fig. 11.4. Scottered Gamma-Ray Dose as a Function of Postscatter Energy. wr PERIOD ENDING JUNE 10, 1954 100° an 80° : OANL-LA-DWG 866 S0 -RADIAL EMISSION < LT /,/.\ N 170° SCATTERED GAMMA=RAY DOSE (M} Fig. 11.5. Scattered Gamma-Ray Dose as a Function of Source Angle. ORNL~LR-DWG 86T o SKEW BEAM * RADIAL EMISSION \\\\‘:\ T ;.\\\‘\ VL 1 \ y Wy § \ h A \ b vy VA } W ‘&\ ~13 10——14 10—14 10—-13 10—-12 10— R r . SCATTERED GAMMA~RAY DOSE (hr‘m) Fig. 11.6. Scottered Gamma-Ray Dose as a Function of Receiver Angle. 127 ANP QUARTERLY PROGRESS REPORT but the excess shielding was supposed to com- pensate somewhat for the presence of ducts in the rear hemisphere of the reactor shield. The distribution of the dose inside the crew shield as a function of the angle around the crew- comperiment midpoint (Fig. 11.6) shows a larger contribution from scattered gamma rays from the rear of the crew-shield side wall, as would be expected becouse the crew compartment was taken to be a long cylinder with walls of uniform thick- ness. lhe variations observed in this case are not nearly so large as those for the source angle. The direct dose as a function of the energy of the gamma rays emerging from the reactor shield is shown in Fig. 11.7. The dependence of the dose upon the energy is determined by the large angle at which the major contribution to the direct dose occurs, as shown in Fig. 11.8 in which the direct dose as a function of the source angle a is pre- sented. Figure 11.8 again shows the extreme im- portance of the radiation escaping arcund the edges of the shadow shield. The detailed calculations of the gamma radiationreaching the crew compartment will be published in o separate report,® THERMAL-NEUTRON-FLUX PERTURBATION BY GOLD FOILS IN WATER E. B. Johnson Physics Division It is well known that when a detector that is not infinitely thin is intfroduced into a neutron flux, the flux is perturbed. The magnitude of the perturbation depends upon the absorption cross section of the detector and its physical dimensions and upon the properties of the medium into which it is introduced. [t is customary at the BSF to use the flux of the ORNL Standard Graphile Pile® as the reference flux for the calibration of foils and counters. However, when g foil is used for thermal-neutron-flux measure- ments in water, the flux perturbation caused by the presence of the foil is different since the slowing- down properties of water differ from those of graph- ite. Klema and Ritchie’ arrived at a semiempirical SE. C. Maienschein, F. Bly, ond T. A. Love, Sources and Attenuation of Gamma Radiation from a Divided Aircruft Shield, ORNL-1714 (to be published). 6¢. D. Klema, R. H. Ritchie, and G. McCamman, Recalibration of the X-10 Standard Graphite Pile, ORNL-1398 (Nov. 11, 1952). 128 correction for this effect for the 25-sq cm 5mil- thick indium foils normally used for flux measure- ments in water, The presence of these foils results in a 22% depression of the flux from its unperturbed value, Frequently it would have been desirable to use gold foils for flux determinations in water, but no value for the flux depression has been available, Therefore an experiment has been completed ot the BSF that was designed for measuring this effect, Thermal-neutron-flux measurements were made with both the indium and the gold (1-sq cm, 5mil- thick) foils ot five positions along the north-south center line of the reactor, The perturbation factor of 1.22 was opplied, os usval, to the indium-foil data, but no such correction was made on the gold- foil data, The results are shown in Fig, 11.9. If it is assumed that a probable error of +2% is associ- ated with each measurement, there is no discernible difference between the thermal-neutron-flux meas- urements mode with indium foils using the perturba- tion correction and those with gold foils using no correction, While it is recognized that the ! . a T presence of any obsorbing material must result in a local depression of the flux, it seems that this effect is within the probable errors of measurement for small gold foils in a water medium, It should be pointed out that such would not necessarily be the cose if gold foils of larger dimensions were used, These results substantiate those obtained by P. M. Uthe® from a similar experiment performed in the water tank at the thermal column of the ORNL Graphite Reactor, REACTOR POWER CALIBRATION TECHNIQUES?® E. B. Johnson G. M. Estabrook R. G, Cochran Physics Division Correlation of shielding data from the TSF with those from the BSF requires that the power de- terminaiions for the two reactors be made by com- parable methods, Power determinations at the BSF have been based on thermal-neutron-flux measure- ments obtained by inserting foils between the 7E. D. Klema ond R. H. Ritchie, Preliminary Results on the Determination of Thermal Neutron Flux in Water, ORNL CF-51-4-103 (Apr. 24, 1951). 8p. M. Uthe, private communication. &, B. Johnson et al., Povrer Calibration Techniques for BSR, ORNL.-1723 (to be published). PERIOD ENDING JUNE 10, 1954 ORNL~LR-DWG 1684 10_? L ) hr-watt -« Mev e GAMMA~RAY DOSE ( 1 4 6 8 10 12 14 £, SOURCE ENERGY (Mev) Fig. 11.7. Direct Gamma-Roy Dose as o Function of Sour?:e Energy. 129 PROGRESS REPORT ANP QUARTERLY (ORNL-LR-DWG 417 ook O INDIUM FOILS X GOLD FOWLS vy Swatt {neutrons semP -sec-watt) THERMAL NEUTRON FLUX, 10 [_fi,.,_..“‘__.,.»ufi_*J__.,_.___.._"..:”. - e - 30 a0 50 50 70 80 90 DISTANCE ALONG REACTQR CENTER LINE FROM NORTH FACE OF ACTIVE [-ATT[CE {cm) Fig. 11.9. Comparison of Indivm- and Gold-Foil Measurements in Water Surrounding BSF Reactor,’ plates of the fuel elements in the reactor lattice, ' ° However, the Tower Shielding fuel elements have lead-shielded tops which make foil insertion im- possible, and, in addition, the reactor control chambers are located directly over the lattice, It has been planned to make the flux measure- ments in the TSF reactor by inserting cobalt wires downthe center of the fuelelements. However, this technique will yield only a total flux rather than 105 |, Meem and E. B. Johnson, Deferminotion of the Fower of the Shield Testing Reactor — I. Neutron Flux Measurements in the Water-Reflected Reoctor, ORMNL- 1027 (Aug. 13, 1951); E. B. Johnson and J. L, Meem, Determination of the Power of the Bulk Shielding Re- actor - {l. Neutron Flux Medasurements in Several Beryllivm Oxide-Reflected Critical Assemblies, ORNL - 1438 (May 7, 1953). PERIOD ENDING JUNE 10, 1954 the thermal flux needed for power-calibration caleu- lations, Therefore, a method for obtaining reactor power directly from total-flux measurements was needed. ln the method used at the BSF for power calibration, the total and epicadmium activities of detectors placed throughout the reactor lattice are obtained. The thermal-neutron flux obtained by this methed is used to calculate the reactor power, since the fission rote ond therefore the power level of an enriched-fuel depend primarily on the thermal-nevtron flux, . The relation between the thermal-neutron flux as measured with foils and the power produced as indicated by strategically placed thermocouples has been very carefully determined.'' It was found that the power of the reactor ‘reactor can be expressed as (1) = P = 4,264 x m‘“cgm , where P is the power in watts and G is the mass of U233 in the volume over which 5”7 is the average thermal flux. The constant contains the fission cross section, energy release per fission, fast- fission factor, and conversion factors., Various methods hove been investigated at the BSF for simplifying ond adapting the currently used power-calibration procedure for use with the TSF reactor. The reactor power was calculated for each of three loadings by using both total flux and ther- mal flux as measured with small cobalt foils, Loading 208 {Fig. 11.10) had a slab reflector of beryllium oxide on all four sides, whereas loadings 22A (Fig. 11.11) and 24B (Fig. 11.12) were com- pletely water reflected. The resulting dota indi- cate that there is a constant ratio (within 1%) be- tween the power calculated from thermal flux and that calculated from total flux. This indicates that if the constant in Eq. 1 is changed to 3.86 < 107 Y the average flux in the equation can be considered to be the total flux as measured with cobalt de- tectors. Flux measurements made by using cobalt wires have been obtained for the BSF reactor but the calculations have not yet been completed. 115, L. Meem, L. B, Holland, and G. McCammon, Determination of the Power of the Bulk Shielding Re- actor — |ll. Measurement of the Energy Releused per Fission, ORNL-1537 (Feb, 15, 1954). 131 ANP GUARTERLY PROGRESS REPORT CORNL-LR-ODWG {418 REACTOR GRID PLATE > am = ) wy w < — wl =3 . o o W@%WU DD - )08)-)8)-, \¥_/// OOOOOC - OOOOO P | CONTROL ROD — T e \ \ PARTIAL FUEL ASSEMBLY Fig. 11.11. BSF Reactor Loading 22A, HOOOHOG | OOOOOO, 133 TN T TN T TN LT T T N Ty )\1/\\!/ \'J\H/m j)) =Ly LT N N Py 14 15 7w X ")lfi A A K m TN \J Q) N ) A\ Woo>oo< OO0 PERIOD ENDING JUNE 10, 1954 ks LEAKAGE-FLUX CHANGES DUE TO ORNI~LR-DWG 1424 3»(10‘*r , ‘CONTROL-ROD SETTINGS | J. D. Flynn R. G. Cochran Physics Division An experiment was performed ot the BSF to de- termine the effect of the leakage flux of changes in setting of the regulating rod and the safety rods in order to obtain better correlation between suc- cessive shielding measurements. At positions close to the reactor foce, some effect would be expected; however, at large distances from the reactor, there should be no effect. The data (Fig. 11.13) indicate that there was no appreciable ef- ,,,,,,,,,,,,,,, B fect on the thermal flux ot distances of 40 to 80 ’ M. " In another experiment, the effect of changing the quantity of U23% in o reactor loading was investi- gated. A 110-g element was replaced with a 140-g element on the side of the reactor opposite to the side at which the measurement was made, and it was found that the leakage flux was insensitive to the quantity of uranium, 2 Nvyy / watt {neutrons/cm ~sec-watt) I o 3 THERMAL MEUTRON FLUX, o a0 DISTANCE FROM NORTH FACE OF REACTOR (cm) Fig. 11.13. Comporison of Effects of Various Control-Rod Settings on Leakage Flux of BSF Reactor. 135 i AT A 08 A 8 A L A T T S S e s s T ANP QUARTERLY PROGRESS REPORT 12. TOWER SHIELDING FACILITY C. E. Clifford T.V. Blosser D. L, Gillilond! .. B. Holland J. L. Hull F. N. Watson Physics Division Construction at the Tower Shielding Facility (TSF) was completed May 1 (Fig, 12.1), and the facility is now operating on a schedule of four days per week. Critical experiments with the reactor were started early in March, and installation of the detector tank and experimental instrumenta- tion was completely by the end of March. Since that time, experimental operation of the facility has been continued intermittently to permit final adjustments and modifications of all equipment, EXPERIMENTAL PROGRAM Preliminary measurements have been taken for the differential shielding experiments in a small water tank (5 x 5 x 5 ft) which is known as the detector tank, For these measurements the reactor was in the 47-ton hemispherical water tank. The relaxation length for the air-scattered neutron radi- ation entering the sides and front of the detector tank is 5.3 cm, which is to be compared to the 4,5-cm relaxation length obtained in the BSF Skyshine experiments? and the 5-cm relaxation fengthused in the 1953 Summer Shielding Session.3 Neutron- and gamma-dose measurements in the reactor tank have been made with the 3-in. fission counter, the 900-cc ion chamber, and the three- chamber BF; counter. These measurements are in qualitative agreement with those taken at the BSF with a similar reactor loading.? The first of a series of measurements for deter- mining the background from ground-scattered radiation has been completed with the reoctor and 10n loan from General Electric Company. 2H. E. Hungerford, The Skyshine Experiments at the Bulk Shielding Facility, ORNL-1611 (to be issued). 3 Report of the 1953 Summer Shielding Session, ORNL- 1575 (to be issued). 4R. G. Cochron, J. L. Meem, T. E. Cole, and E. B. Johnson, Reactivity Measurements with the Bulk Shield- ing Reactor, ORNL-1682 (to be issued). 136 detector tanks 4 ft above the ground, The separa- tion distances of the tanks were 35, 65, and 100 ft. Since the power of the reactor is not absclutely established, the dota from this experiment will be reported first in terms of radiation intensity relative to a nominal power, which is expected to be within 30% of the octual power. OPERATION OF THE REACTOR The reactor was loaded to criticality on March 12. This loading consisted of 26 fuel elements, the minimum number required for criticality, Since that time, ancther loading (30 fuel elements ina 5 by 6 array) has been checked by a critical experi- ment and is now being used for the preliminary measuyrements, Both these loadings were run earlier in the BSF reactor with TSF reactor fuel elements, and the power distribution of the 5 by 6 element array was determined with BSF reactor elements, The twe TSF reactor loadings are com- parcble in shim-rod and regulating-red positions to those used at the BSF, During the experiments the reactor was run at power levels of 1 to 10 watts at the 200-ft eleve tion and at approximately 4 kw at just above the poo! level, The instruments and controls performed satisfactorily under both conditions. Use has been made of the remote visual-indicator system to observe operation and to position the reactor and the crew compartment. The ease with which it is possible to position the large loads by remote control with this system has exceeded expectations. A system of slack-line limit switches has been installed on the large hoists to prevent damage to the hoist gear mechanism such as that experienced during the run-in period. The operation of the switches is satisfoctory, but, to odd assurance, the cables will be kept taut by additional counter- weights that will be odded to the floating sheave blocks. el s S SRR ’%{\;}')“ s e e NI nY e o o Ry b S ] e 5 73:; 2 i S e R o0 SR e '.'&"A 3 3 S Fig. 12.1. Completed Tower Shielding Facility. i 2 B v:~;;:§.‘:f:-‘a;@¢s_ S S i T et G T Qgi::‘:fu; 7 o T J\" Y PS6L ‘Ol INNr ONIANI AOIY3d Part IV APPENDIX REPORT ND. CF-54-3-65 CF-54-4-218 CF-54-5-51 CF.54-2-185 CF-54-4-6 CF-54-4-53 CF-54-4-221 CF-54-5-1 CF-54-5-196 CF-54-3-37 CF-54-3-194 CF-54-4-195 ORNL-1688 ORNL-1716 CF.54-4-47 CF-54-5-40 CF-54-5-47 CF-54-5-189 CF-54-6-6 CF.54-315 CF-54-3-193 13. LIST OF REPORTS ISSUED DURING THE QUARTER TITLE OF REPORT }. Aircraft Reactor Experiment ARE Design Data Supplement ARE Instrumentation List Analysis of Critical Experiments Il. Reflector-Moderated Reactor Effects of Reactor Design Coenditions on Aircraft Gross Weight The Kinetics of the Circulating-Fuel Reactor Reflector-Moderaféd Critical Assembly Experimental Pro- gram Radiation Damoge Elastomers, Lubricants, Fabrics, -and Plastics for Use in Nuclear-Powered Aircraft The Xenon Problem in the ART Stresses in Heat- Temperature Gradient and Thermal Generating Bodjies Iti. Experimental Engineering 50-Mw Design — Canned Be Crosgs Sections Discussions and Results of the Dielectric Heating Test . Sodium Plumbing from a Molten Flueride 3Salt in a Double-Tube Turbulent Heat Transfer Mixture to Sodium-Potassium Alloy Meat Exchanger IV;. Chemistry Daia on the Ber-NuF System Effects of Fission Products in Performance of a Reactor Using Fluorides as Selvent for Fuel Measurement of the Stability of Lithium Hydride Analysis of Salt Mixtures Fused Salt Compositions ¥. Metallurgy Metallographic Examination of Forced Circulation Loop No. 2 High Flow Velocity and High Tempefufure Gradient Loops » E - ®I ®O ¥ B I orE» 00 AUTHOR(s) B, Cottrell G. Affel B. Mills - J. M. K. Stumpf Wilner Ergen . Scott L. Greenstreet . Stumpf . Wilner J M L. A Meem . Field . Mills . Mitls . Southern . Cottrell . Mann . Sglmon . Orbar R. F. Newton E. Orban F. F. Blankenship C. J. Barton G. M. Adamson W. D. Manly DATE ISSUED 3-2-54 4-7-54 5-12-54 5-21-54 . 5-5-54 4-8-54 4-15-54 5-3-54 5-21-54 3-9-54 3-8-54 4-27-54 8-14-53 To be issued 4-27-54 5.7.54 5-26-54 5-25-54 6-2-54 3-3-54 3-18-54 141 ANP QUARTERLY PROGRESS REPORT REFPORT NO. CF-54-4-162 CF-54-4-224 CF-54.5-88 CF-54-5-91 ORNL-1667 CF-54-4-205 CF-54-5-159 CF-54-5-160 ORNL-1701 ORNL-1702 CF-54-4-96 ORNL-1712 CF-54-5-41 CF-54-6-4 CF-54-5-201 CF-54-5-219 ORNL.-1682 142 TITLLE OF REPCRT Status of Coelumbium, Beryllium, and Lithium Report of Literature Survey of Beryllium Heat Exchanger Fabrication Free Energies of Formation of Oxides and Fluorides Alkali-Meta! Nickel Oxides Containing Trivalent Nickel Vl. Heat Tronsfer and Physical Properties A Feasibility Study of Flow Visualization Using a Phos- phorescent Particle Method Effect of Oil Contomination on the Boiling Heat Transfer Characteristics of Heat Exchangers and Solid Fuel-Plate Reacters Heat Caopacities of Composition No. 12, No. 44, and of l(:st_.rl:‘5 Forced Convection Heat Transfer Between Parallel Plates and in Annuli with Volume Heat Sources within the Fluids A Summary of Density Measurements on Molten Fluoride Mixtures and a Correlation Useful for Predicting Den- sities of Fluaride Mixtures of Known Composition AUTHOR(s) W, D, Manly J. W. Woods S. T. Vil. Anolytical Studies of Reactor Maoterials Data on Aircraft Fuel Samples The Optical Chloride Compounds Properties of Some Inerganic Fluoride and VItl. Radiation Damage Metallographic Analysis of Pratt and Whitney Capsules 1-4, 1-6, and 1-7 Release of Xenon from Fluoride Fuels: Proposal for an Experimental Program 1X. Shielding Secondaery Gamma Ray Study Stant Penetration of Neutrons Through Water Reactivity Measurements with the Bulk Shielding Reactor T O ©Owr - woOm T v U - I ©=x D OO9 . E. Hoffman . Patriarca M. Cisar . Dyar Borie, Jr. . Smith D. Paolmer M. Winn W. Rosenthal L. Milles . Powers . Blalock . Poppendiek . Palmer l. Cohen N. Jones R. Baldock > . N. McVay . D. White . J. Feldman . E. Richt M. T. Rekinson A et . K. Trubey . T. Chapman . G. Cockran al. DATE ISSUED 4-6-54 4-13-54 5-17-54 5-5-54 2-26-54 4-30-54 5-19-54 5-20-54 5-11-54 5-14-54 4-14-54 5-5-54 5-5-54 6-2-54 5-26-54 5-28-54 To be issued REPORT NO. ORNL-1714 ORNL 1723 ORNL-1686 ORNL-1692 PERIOD ENDING JUNE 10, 1954 TITLE OF REPORT Sources and Attenuation of Gamma Radiation from a Di- vided Aircraft Shield Power Calibration Techniques for BSR X. Miscellaneous Quantities and Reactions of Solid Surfaces Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending March 10, 1954 AUTHOR(s) DATE ISSUED F. C. Maienschein To be issued F. Bly T. A. Love E. B. Johaseon To be issuved ef al. E. P. Carter 4-7-54 A. W. Savolainen 4:15-54 143