ORNL-1649 This document consists of 152 pages. Copy “?éof 2646 coples. Series A, Contract No. We7405~enge24 AIRCRAFT NUCLEAR PRGPULSION PROJECT QUARTERLY PRGGRESS REPORT For Period Ending December 10, 1953 R. C;:Briuht, Director A. J. Miller, Assistant Director W. B. Cottrell, Editor DATE {SSUED SAN 18 1954 OAK RIDGE NATIONAL LABORATORY :Operated by CARBIDE AND CARBON CHEMICALS COMPANY A Divisien of Union Carbide and Carbon Corporation Post Office Box P : Ock Ridge, Tennessee IR 3 4456 03L0OLAI & FO Oz MEZOQAILPPEIPECENPIEIOOLROCTPIREMADMO Qo x 1 T =, . Adamson . Affel . Baldock Barton . Blizafd, . Briant . Bruce . Callihan . Cardwell Cathcart . Center . Charpie . Cisar . Clewett . Clifford . Cottrell . Cochran . Cowen . Culler . Eister . Emlet (Y-12) Ergen . Fraas ||qe er . Fiowe gcl—imIJUE‘U?'CUJ?R‘F‘OOUJFHI§>m.<=EUZJO>'U-nE,,.'.§:J;: . Johnson . Keilholtz . Keim . Kelley ertesz INTERNAL DISTRIBUTION . Blenshlp . Bredig . Grigorief] {consultant) . Grlmesm e seholder # Humes (K-25) %, 64 ”g 65 . 6?‘* 68. %A 69. 70. 71. 72. 73. 74, EgflmmIbUFI'flmfliigflg;Nt_f—055:553?« l-.. ORNL-1649 Progress » M. King. - A. Lane . E. Ldrsgfi: . S. Livingston . N. L};Brn . D Manly . Mann . McDonald . Meem . Miller . Morgan . Murphy . Nessle . Poppendiek . Reyling . Savage . Savolainen . Shipley isman . Smith (consultant) . Snell . Storrs . Susano . Swartout . Taylor . Trice . Van Artsdalen . YonderLage . Warde . Weinberg . White G. E&Whlfman E. P. fllgner (consultant) G. C. Wfilmms J. C. Wllsem C E. Wmfer& 75-84. ANP Library #, 85. B lology Li brar;fi?:_;;,_. iii \ 86-&’6. Laboratory Records Department 94. Reactor Experimental 9]\:\\€_quoratory Records, ORNL R.C. Engineering Library 92. Health Physics L.ibrary ' 95-97. Central Research Library 93. 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S.d8aval Radiological Defense Lab"fi: 1tory 238. eadquarters ’ '~ 239-240. rsity of California Radiation Laboraf’%y, Berkeley 241-242. Hiversity of California Radiation Luboratorg, Livermore 243, A alter Kidde Nuc!ear Laborotorles, inc. % 244-249.5 250-264; Technical Information Service, Oak Ridge, Tennessee , Curtiss-Wright Corp., Wright Aeronautical Division (K. Campbell) Armed Forces Special Weapons Project, Washington vi ORNL-528 ORNL-629 ORNL-768 ORNL-858 ORNL-919 ANP-60 ANP-65 ORNL-1154 ORNL-1170 ORNL.-1227 ORNL-1294 ORNL-1375 ORNL.-1439 ORNL-1515 ORNL-1556 ORNL-160% Reports previously issued in this series are as follows: Period Ending November 30, 1949 Period Ending February 28, 1950 Period Ending May 31, 1950 Period Ending August 31, 1950 Period Ending December 10, 1950 Period Ending March 10, 1951 Period Ending June 10, 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 Merch 10, 1953 Period Ending June 10, 1953 Period Ending September 10, 1953 CONTENTS FOREWORD PART I. REACTOR THEORY AND DESIGN INTRODUCTION AND SUMMARY 1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT The Experimental Reactor Control rod sleeves Reactor control Flow in fuel tubes Fuel tube cleaning Reactor Physics Critical mass Effect of delayed neutrons Xenon poisoning The NaF-ZrF ;-UF ; Fuel Physical properties Fuel production Corrosion of Inconel Pumps Prototype sodium pump test Pump auxiliaries impeller fabrication Acceptance tests Fluid Circuits Fuel system Sodium system Valves Auxiliary systems {nstrumentation Reactor and Fluid Circuit Cleaning Fuel Recovery and Reprocessing Transportation of fuel Rate of dissolution of ARE fuel Molten fuel dissolution EXPERIMENTAL REACTOR ENGINEERING Pumps for High Temperature Liquids Frozen-sodium-sealed pump for sodium Gas-sealed sump pump for in-pile loop test Covil Magnstic-torque-transmitter pump Rotary-Shaft and Yalve-Stem Seals for Fluorides Graphite-packed seal for spiral-grooved shaft Graphite-BeF , packed seals V-ring seal Bronze-wool, graphite, and MoSz-packed frozen seal Packed seals for valve stems High-Temperature Bearing Development Materials compatibility tests Bearing characteristics Heat Exchanger Test Forced-Circulation Corrosion Loop . REFLECTOR-MODERATED REACTOR DESIGN STUDIES Reactor Physics Fuel Composition and Properties Moderator Regions Hydrodynamics of the Fuel Circuit Pump Design Pressure Shell Fuel-to-NaK Heat Exchanger Reactor Controls Shielding Filling and Draining of the Reactor CRITICAL EXPERIMENTS PART Il. MATERIALS RESEARCH INTRODUCTION AND SUMMARY 5. vifi CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS Thermal Analysis of Fluoride Systems NaF-ZrF ,-UF, RbF-LiF-UF, NaF-ThF, LiF-RbiF-BeF, RbF-BeF ,-ZrF Therma! Analysis of Chloride Systems LiCl-UClH, NaCl-UCI, KCI-UCl, RbCl-UCI 24 24 24 24 24 25 25 25 25 26 28 29 31 31 33 33 38 39 4] 42 42 43 43 44 A7 49 49 49 50 S0 51 51 51 51 51 $1 52 60 CsCl-UCl, KCI-LiCl-UCl, RbCI-UCI, CsCI-UCl KCI-LiCl-UCl, NaCI-LiCl-UCI Quenching Experiments with the NaF-ZrF , System Differential Thermoanalysis of the NaF-ZrF , System Filtration Analysis of Fluorides Mixtures with 53 mole % NaF Mixtures with 50 mole % NafF Mixtures with 75 mole % Naf Fundamental Chemistry _ Spectrophofometry of supercooled fused salts EMF measurements in fused salts Physical chemistry of fused salts Production of Purified Fluorides L aboratory-scale production of molten fluorides Purification of smali fluoride samples for phase studies Production of enriched material for in-pile loop experiment Experimental production facilities Reduction of Na,UF , by hydrogen Treatment of molten NaZrF . with strong reducing agents Reduction of Ni!f:2 by hydrogen Preparation of various fluorides Purification and Properties of Hydroxides Purification of hydroxides Reaction of sodium hydroxide with carbonaceous matter CORROSION RESEARCH Fluoride Corrosion in Static and Seesaw Tests Inconel corrosion by fluorides with metal fluoride odditives Corrosion of various metal combinations Corrosion of cermets nconel with oil and trichloroethylene additives Inconel corrosion by fluorides with MoS, additive Screening tests of metallic bearing materials Fluoride Corrosion of Inconel in Thermal Convection Loops Effect of fluoride batch purity Effect of chromium additives Pretreatment of fluoride with inconel Effect of exposure time Effect of temperature 52 - 52 52 52 53 33 55 55 55 55 56 56 -7 58 59 59 59 59 60 60 60 61 62 63 63 63 64 65 65 65 66 67 &7 67 67 &7 71 72 72 72 EHect of surface-to-volume ratio Fluoride with 6.5 mole % UF, The fluoride NGZrF5 Fluoride Corrosion of Nickel and Stainless Steel Loops Liquid Metal Corrosion Mass transfer in liquid lead Static tests of BeD in sodium, lithium, and lead Spinner tests of Inconel and type 405 stainless steel in sodium Static tests of beoring materials in sodium, lithium, and lead Static tests of stainless steels in lithium Static tests of solid fuel elements in sodium ond sodium hydroxide Fundamental Corrosion Research Oxidizing power of hydroxide corrosion products Equilibrium pressure of hydrogen in hydroxide-metal systems Mass transfer of chromium in Inconel-fluoride systems . METALLURGY AND CERAMICS Welding and Brazing Research Brazing of radiator assemblies Brazing of high-conductivity radiator fins “*Electroless’’ preplating of brazing alloys Inert-arc welding of solid fuel elements Mechanical Properties of Incone! Stress-rupture of Inconel in fluoride fuel Tube-burst tests of triaxially stressed tubes Environmental effects on creep of Inconel High-Conductivity Metals for Radiator Fins Diffusion barriers Clod copper Electroplated copper Solid phase bonding Fabrication of Special Materials Extrusion of Inconel-type alloys Rolling of chromium-cobalt alloy Rolling of cobalt Rolling of iron-chromium-nicke! alloy Cold-rolled columbium alloy Tubular Fuel Elements Ceramic Research Glass-type pump seals Ceramic container for fuel High-density graphite Combustion of Sodium Alloys 72 72 73 73 74 74 75 75 76 76 76 80 80 80 82 83 84 84 86 89 89 21 91 91 92 92 93 93 93 93 93 93 94 94 94 24 94 96 96 96 96 726 8. 10. HEAT TRANSFER AND PHYSICAL PROPERTIES RESEARCH Heat Capacity Thermal Conductivity of Solidified Salts Density and Viscosity of Fluorides Vapor Pressures of Fluorides Electrical Conductivity of Fluorides Forced Convection Heat Transfer with NaF-KF-LiF Eutectic Flow in Thermal-Convection Loops Fivid Flow in an Annulus Transient Surface-Boiling Studies Circulating-Fuel Heat Transfer . RADIATION DAMAGE Irradiation of Fuel Capsules Creep Under lrradiation In-Pile Circulating Loops ANALYTICAL STUDIES OF REACTOR MATERIALS Analytical Chemistry of Reactor Materials Oxidation states of iron Oxidation stotes of chromium Determination of UF 4 and "U‘Fd Reducing power of NaZrFs with zirconium addition Dissolution of fluoride mixtures containing zirconium Petrographic Examination of Fluorides Mass Spectrometer Investigations of irradiated Fluoride Fuels Calculation of UF 4 in unirradiated fuels Caleulation of U235 lost from irradiated fuel Determination of U232 burnup Summary of Service Chemical Analyses PART 1il. SHIELDING RESEARCH INTRODUCTION AND SUMMARY 11. BULK SHIELDING REACTOR Spectrum of Gamma Rays Emitted by the BSR Spectrometer arrangement Results Fast-Neutron Leakage Spectra of the BSR 12, LID TANK FACILITY Slant Penetration of Neutrons Through Water Air Duct Tests 97 97 98 98 99 100 101 102 102 102 103 103 104 106 107 107 107 108 108 109 109 109 110 110 11 11 112 115 116 116 116 116 117 120 120 121 xi 13. 14, 15, 16, Removal Cross Sections Survey of Lid Tank Experiments TOWER SHIELDING FACILITY SHIELDING ANALYSIS Visible Light from a Nuclear Power Plant Neutron Reflection Coefficient for Water PART V. APPENDIXES LIST OF REPORTS ISSUED DURING THE QUARTER DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS AT ORNL ORGAMIZATION CHART xii 122 123 125 126 126 126 129 131 139 ANP PROJECT QUARTERLY PROGRESS REPORT FOREWGORD This quarterly progress report of the Aircraft Nuclear Propulsion Project ot 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 into three major parts: |. Reactor Theory and Design, Il. Materials Research, and Iil. Shielding Research. Each part has a separate introduction and summary. 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 portien 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 Reactor 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 included in this ARE section, where convenient. The few exceptions are referenced to the specific section of the report where more detailed infor- mation may be found. The ANP research, in addition to that for the Aircraft Reactor Experiment, falls into 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 three phases of research are covered in Parts |, H, and llI, respectively, of this report. INTRODUCTION AND SUMMARY Assembly of the Aircraft Reactor Experiment {sec. 1) is nearing completion; all but a few of the components have been received and installed. The items still missing include the new control rod sleeves and parts of the fuel and sodium gas-sealed pumps. As the installation progresses, various auxiliary systems, such as the helium, water, and hydraulic off-gas systems, are being subjected to operational tests. Tests of a prototype pump have established the design criteria for the pump cooling and lubricating systems, as well as the operating characteristics of the pump. sleeves will effect a reduction in the structural poison in the core and hence reduce the critical mass. the uranium requirement inside the pressure shell will be well under 40 Ib of U235, Other physics calculations on reactor kinetics reveal that reactor operation and control will not be adversely affected by either xenon poisoning or the loss of delayed neutrons in the circulating fuel. The fuel, which will be obtained by the addition of d concentrate, NazUFé, to a solvent mixture, NaZrF_, has been shown to be reasonably compatible with the Inconel container metal for the temperatures and times required. Production of both the solvent and con- centrate are essentially complete and the impurities in each are well below acceptable levels. Pro- cedures have been established to assure that the fuel and sedium systems in the experiment will be adequately cleaned prior to being filled with the solvent and sodium, respectively. Additional studies of the fuel recovery and processing problem have established dissolution rates for both molten and solid fuel in batches containing 4 kg of U235, With the near-completion of the Aircraft Reactor Experiment the emphasis of the experimental work has shifted to the development of components for general aircraft reactor application and supporting research. Valves, pumps, bearings and other components of high-temperature fluoride and liquid metal systems are being developed for these studies (sec. 2). Satisfactory, short, frozen-sodium shaft seals, with length-to-inside-diameter ratios of 1 to 5, have been developed for sodium pumps. A small gas-seaied pump for an in-pile loop and a canned The new control rod Recent physics calculations indicate that magnetic-torque transmitter for g high-temperature gircraft pump are being developed. Packed seals for fluoride pumps and valves have achieved {imited success under controlled conditions, but none of the seals are sufficiently reliable for service conditions. Equipment for high-temperature bearing tests is being assembled, and a number of potential bearing materials are being screened on the basis of wear. A fluoride-to-sodium heat exchange tube bundle and a high-velocity forced-convection cor- rosion loop are being operated, but results are not yet available on either test. The designs of a family of 50- to 300-megawatt reflector-moderated reactors have been prepared to permit an Air Force evaluation of over-all aircraft performance (sec. 3). The four reactors considered for 50-, 100-, 200-, ond 300-megawatt power out- puts are all of the same general species, having all been extrapolated from the same base., The designs are based upon presumed fuel and were made with insufficient physics information for the establishment of over-all design limitations. In other respects, however, the designs are on a firmer basis, since considerable shielding and heat exchanger data are availableand the hydrodynamics of the fuel circuit and fuel pump have been con- firmed by hydraulic mockups. The gas-sealed centrifugal pump is designed with special baffling above the impeller so that it will cperate con- tinuously with its shaft 80 deg from the vertical, and it will even operate for 1 min when inverted without “‘gassing®® the pump. Other aspects of the aircraft reactor, including moderator cooling, assembly of the reactor, control, and fuel filling and draining, are discussed. The Critical Experiment Facility was used during this quarter to determine the static physics charac- teristics of a mockup of an air-cooled water- moderated reactor for the General Electric Aircraft Nuclear Propulsion Project (sec. 4). At the same time, preparations were made for measurements of o “mockup of a Pratt and Whitney supercritical-water reactor and a Nuclear Development Associates sodium-cooled reactor, as well as the Laboratory’s reflector-moderated reactor. ' ANP QUARTERLY PROGRESS REPORT 1. CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT E. 5 Bettis Jo L.n Mcem ANP Division There were no modifications in the ARE design or critical developments in the ARE program during the quarter, and therefore installation of equipment received maximum attention. The installation is proceeding satisfactorily, ond there was no mojor holdup occasioned by lack of components. For the first time since the project began, it is not neces- sary to explain either the cause for some delay or some change in plan, and this is the most signifi- cant remark that can be made to indicate the status the project attained during this quarter. [t is fairly clear now that completion of the installation of the ARE can be effected in an uneventful manner, One fuel pump sump tank ond one sodium pump sump tank were received, and therefore the instal- lation of piping for the fuel and sodium circuits could proceed in a relatively uninhibited manner, even though the spare fuel and spare sodium tanks have not yet been received. Tests on a prototype pump have established the requirements of the auxiliary systems for the pumps, as well as the pump's operating characteristics and performance. Since all major components of the experiment are now on hand, all crafts have been able to proceed with relatively little interference. Installotion of elecirical heaters ond insulation is progressing rapidly, The heaters are approximately 65% in- stalled, but, since the locations for the remaining heaters are more accessible, the job is 75% com- plete with regard to man-hours. Insulation work is following closely behind heater installation. The most difficult insulating job, that of insulating tanks and lines in the dump tank pit, is 95% fin- ished, and cbout 15% of the other insulation has been installed. Additional daota hove been obtained from calcu- lations of the reactor statics and dynamics. An evaluation of the effects of xenon poisoning und of the loss of delayed neutrons indicated that neither will be of much consequence to reactor operation. Caleulations of the critical mass by two different methods gove values which show that the uranium requirement in the pressure shell will be well be- low the 40 lb permissible from the viewpoint of vranium allocation and ARE technology. A soving in mass of about 4 |Ib was realized by a reduction in the peisoning from the control rod sleeves. The new Inconel-clad control rod sleeves, which will replace the three concentric rod sleeves thai were to be used around each rod, have been fabri- cated. The spoce outside each new sleeve will be filled with beryllium oxide strips. The reactor, which is complete except for installation of the new rod sleeves and the beryllium oxide strips, has been subjected to preliminory cleaning tests. Tests with simulated reactor fuel tubes indicate that fully established turbulent flow will be real- ized at a Reyneolds number of 5000, which is wel! under the ARE design value of 14,000, The ultimate fuel will be obtained by the addi- tion of sufficient fuel concentrate, N02UF6, to the fuel solvent, NaZrF_, to make the reactor critical. The production of the 3300 Ib of solvent was com- pleted some time ago, and conversion of the 115 kg of U233 into Na,UF is nearing completion, Initial tests of the concentrate show that, with the exception of one batch which was rejected, the impurities in the concentrate are gratifyingly low. Density and viscosity data have been obtained for the concentrate, Methods for recovering and reprocessing the fuel after completion of the experiment have received increased attention. It now appears that the used tuel may be transported from the experiment to the reprocessing area in aluminum containers contain- ing 4 kg of U?33 and that about 20 such transfers will be required. While the reprocessing procedures had been established previously, additional dota have been obtained on dissolution rates for the fuel in both liquid and molten states. THE EXPERIMENTAL REACTOR The reactor is completely assembled (Fig. 1.1) except for the Inconel-clad stainless steel control rod sleeves, which have not yet been received, and the beryllium oxide strips that will be ploced around these sleeves, The fuel tubes have been cleaned according to the established cleaning pro- cedures. Additional measurements of flow phe- nomena in glass mockups of the fuel tube indicate that fully developed turbulent flow is realized at Reynolds numbers abeve 5000, that is, well ynder the reactor design value of 14,000. PERICD ENDING DECEMBER 10, 1953 UNCLASSIFIED PHOTO 11991 SODIUM QUTLET LINE CONTROL ROD S T " 3 REGULATING-ROD SLEEVE FISSION C N : Fig. 1.1. Photograph of the Experimental Reactor. ANP QUARTERLY PROGRESS REPORT Control Rod Sleeves In order to reduce the structural poisons in the reactor, it was decided to replace the three con- centric Inconel sleeves around each of the four control rods (one regulating and three safety) with a single Inconel-clad stainless steel sleeve in each hole. Fabrication of the new control rod sleeves has been finished by the International Nickel Company, but they have not yet been de- livered. To further reduce the critical mass of the reactor, the volume outside the rod sleeve was to be packed with beryllium oxide pellets. However, since the beryllium oxide pellets would occupy only about 50% of the volume, it was recently decided to fill this volume with strips from unused beryllium oxide moderator blocks. It is anticipated that approxi- mately 90% of the void volume can be filled with the beryllium oxide strips. Reactor Control A. Lo SOUthern Jo Kn LeSIie ANP Division There has been only one chonge (on addition) to the reactor control equipment. An auxiliary low- speed drive has been designed for the regulating rod. This addition was made to provide a means of manually operating the regulating rod when the reactor is operating at zero power during rod cali- bration. In the event of fission chamber failure or in- adequate Ak in the regulating rod, it may be neces- sary to change these components during the course of the experiment. Consequently, some of the drive mechanism has been relocated to facilitate replacement of both the regulating rod and the fission chambers. No other changes have been made in the reactor controls and this phase of the work is complete insofar as preliminary checking can determine completeness, Flow in Fuel Tubes J. Lang G. M. Winn Reactor Experimental Engineering Division The experimental relationship between the friction factor and the Reynolds modulus obtained in a glass replica of the reactor fuel tube system —-— indicated that the transition region (that region which lies between laminar and fully developed turbulent flow) lies approximately between Reynolds moduli of 2000 and 5000. Recent photographs, Figs. 1.2, 1.3, and 1.4, show the diffusion of dye filaments in the second 180-deg return bend in a fuel tube under three Reynolds number flow con- ditions: 600, 1300, and 2000, respectively. Un- fortunately, there was more than one dye source present in the entrance region of the fuel tube system, and thus several filament trajectories were superimposed. In general, the sharp, distinct filaments shown in Figs. 1.2 and 1.3 are character- istic of laminar flow, whereas the diffuse filaments indicate some turbulence, Fig. 1.4. Fuel Tube Cileaning G. D. Whitman, ANP Division The cleaning procedures for the fuel and sodium systems inside aond outside the reactor are de- scribed in a following subsection (cf., ‘*Reactor and Fluid Circuit Cleaning’’). However, a de- scription of the application of these procedures to the actual cleaning of the reactor fuel tubes follows. The six parallel fuel circuits in the reactor were first rinsed, individually, with tap water. This flushing was accomplished by valving off the out- lets of all but one circuit and passing water at a pressure of 60 psi into the fuel inlet header. Following this pressure flushing, each circuit was flushed with distilled water, and the water was blown out as completely as possible with helium. After the water that could be blown out had been removed, a vacuum system was connected through a dry-ice cold trap to the fuel tubes, and the re- maining water was evaporated from the fuel circuit at room temperature, During the evaporation pro- cess 2 cth of helium was bled into the system inlet to sweep out all moisture. The residual moisture was determined by the Karl Fisher and dew-point methods. The reactor will be complete when the control rod sleeves and the bottom fuel outlet manifold have been installed, and it will then be ready for inclusion in the system. TORNL-1609, ANP Quar. Prog. Rep. Sept. 19, 1953, J. l. Lang and G. M. Winn. PERIOD ENDING DECEMBER 10, 1953 e L UNCLASSIFIED PHOTO 20724 Fig. 1.2. Diffusion of Dye Filaments in a Mockup of a Fuel Tube at a Reynolds Number of 600 (Second 180-deg Bend). ANP QUARTERLY PROGRESS REPORT UNCLASSIFIE PHOTO 20738 0 SR Fig. 1.3. Diffusion of Dye Filaments in a Mockup of a Fuel Tube at a Reynolds Number of 1300 (Second 180-deg Bend). 10 PERIOD ENDING DECEMBER 10, 1953 WA UNCLASSIFIED | PHOTO 20725 Re—-2000 Fig. 1.4, Diffusion of Dye Filaments in a Mockup of a Fuel Tube at a Reynolds Number of 2000 (Second 180-deg Bend). 11 REACTOR PHYSICS W. K. Ergen J. Bengston Cl Bo Mi”s, ANP DiViSion The UNIVAC multigroup calculations and the locally performed two-group calculations gave values which show that the reduction of the poison- ing by the control rod sleeves has brought the ARE uranium requirement in the pressure shell to well below the 40 |b permissible from the viewpoint of uranium allocation and ARE technology., The de- layed neutrons lost from the reactor by fuel circu- lation are predominantly those of long decay constants, and hence a small amount of excess reactivity makes it unnecessary for the reactor to wait for the long-delayed neutrons. Small excess reactivities will bring the reactor to relatively short-time constants (10 sec or so), but when the reactor is at such a time constant it is still sub- stantially below the prompt-critical condition be- cause the short-delayed neutrons are not appreciably reduced by fuel circulation. Speculation as to the xenon poisoning shows that the poisoning effect would be easily observable if the xenon were not lost from the fuel and if the xenon cross section were as large for the ARE as it is for low-tempera- ture reactors. Failure to observe the poisoning might mean that Xe!3® or its parent iodine had evaporated from the fuel, or it might mean that the effective xenon cross section was reduced by the high moderator and neutron temperature. In the latter case, xenon poisoning would still be a prob- lem in high-powered reactors of the ARE type. Critical Mass The multigroup calculations made by using the UNIVAC at New York University and essentially the method previously employed? by the ANP Phys- ics Group have now yielded fairly reliable, though not yet final, data regarding the uranium requirement of the ARE. These data are in agreement with data obtained by the two-group method. They indicate that the elimination of much of the poison origi- nally incorporated into the control rod sleeves has brought the mass requirement to about 15% below the 40 Ib which would be permissible from the viewpoint of phase diagrams, allocation. corrosion, and 2Reporf of the Technical Advisory Board to the Techni- cal Committee of the Aircraft Nuclear Propulsion Program, ANP-52 (Aug. 4, 1950). 35. Glasstone and M. C. Edlund, The Elements of Nuclear Reactor Theory, p. 329 ff., Van Nostrand, New York, 1952, 12 Effect of Delayed Neutrons About one half the delayed neutrons are given off outside the reactor and lost from the chain re- action. These delayed neutrons are predominantly those with long-lived precursors. Most of the short- lived precursors decay before they are swept out of the reactor, and their delayed neutrons are hence not appreciably reduced by the fuel circulation. Since the neutrons with the long delay times are not very effective, it will take only a small excess reactivity to allow the chain reaction to proceed without ‘‘waiting’’ for these neutrons. A small excess reactivity will hence bring the reactor to a relatively short time constant of about 10 seconds. When the reactor is operating at full power, it will probably exhibit small reactivity fluctuations, and this may temporarily bring the reactor from just critical (infinite period) to a period as low as 10 seconds. Hence, it might be difficult to keep the reactor steadily at a period longer than 10 seconds. It should be noted that reactivity fluctuations are not expected at low power operation so that it should be possible to operate the reactor on a 30-sec period during the preliminary, low-power experiments. Because of the only slightly reduced effectiveness of neutrons with short delay times, the amount of additional reactivity required to bring the reactor from a 10-sec period to prompt critical is quite substantial, and the fuel circu- lation of the ARE does not appreciably reduce this safety feature below its value familiar in stationary- fuel reactors. Xenon Poisoning Renewed consideration was given to the problem of xenon poisoning of the ARE. There is reason to believe that the Xe!33, and probably also its parent iodine, will evaporate from the fuel so that the xenon poisoning effect will not occur in a reactor of the circulating-fluoride-fuel type. Confirmation of this would be a major accomplishment of the ARE. To investigate the problem experimentally, the reactor would have to be run for a time some- what shorter than the half life of 135 at full power and then reduced to very low power. |[f the xenon poisoning is appreciable, the reactivity will exhibit the characteristic behavior of first dropping because of xenon formation by 1135 decay and of then increasing because of Xe!35 decay. Calculations performed by standard methods® show that this behavior should be easily detectable if the iodine and xenon do not evaporate from the tuel. If the behavior is not found it could mean that the poison is given off. However, it couid also mean that the xenon poisoning is largely reduced by the high moderator and neutron temperature and that the xenon cross section is rather low for neutrons that are somewhat faster than room-temper- ature thermal neutrons. If the latter explanation were correct, even with reduced cross section, would be a poison for reactors of higher power than the ARE. There is the interesting possibility that the xenon, and maybe the iodine, is released from the fuel at a rate comparable to the rate of radiocactive decay. If so, the build-up and decay of the xenon poisoning could still be observed, but the time constants would be different. A quantitative analysis of the reactivity following a full-power run would disclose this situation. There is also the theoretical possibility that other fission fragments have large resonances at energies above those corresponding to the xenon resononce. in high-temperature reactors such as the ARE, these poisons might conceivably be observable. xenon, absorption neutron THE NaF-ZrF4-UF4 FUEL The reactor fuel will be obtained by the addition of a sufficient amount of the fuel concentrate, Na2UF6, to the fue! solvent, NaZrF_, to make the reactor critical. [t is expected that the final fuel composition will contain approximately 5.5 mole % UF,. In lieu of more precise knowledge of the ultimate fuel extensive material research and physical property measurements have been conducted on the fuel composition with 4 and with 6.5 mole % UF ,, as well as on the carrier and solvent., About 3300 [b of the carrier hasbeen produced and is being held under helium pending its use in the experiment. The batch production of the fuel concentrate is nearing completion. The impurities in all but one of these batches were exceptionally low, averaging (not including the one poor batch) 48 ppm of Fe, 16 ppm of Cr, and 46 ppm of Ni. composition, Physical Properties H. F. Poppendiek Reactor Experimental Engineering Division The densities and viscosities of the fuel solvent, NaZrFs, and of the fuel concentrate, Na UF being measured; these properties of the fuef wnth 6.5 mole % UF, were reported previously.? Pre- PERIOD ENDING DECEMBER 10, 1953 liminary density and viscosity values for Ma, UF are: viscosity, 17.5 cp at 725°C, 9.9 cp at 975°C; density, 5.598 — 0.00119 T g/cm?®, where 660°C < T < 1000°C. Fuel Production G. J. Nessle J. E. Eorgan F. A. Doss Materials Chemistry Division Production of the required 3300 ib of NaZrF now complete, and the material is being held under helium pending its use in the ARE. Processing of the 253 Ib of U235 to produce the fuel concenfmf&, No U¥F g+ 1S nearing completion. e fuel concentrate is being produced by the Y-12 Production Division to permit control of the large quantities of enriched uranium (93.5% U233) involved. Several weeks were spent training operators in the handling of the equipment, as well as in testing the equipment. A few necessary changes and repairs were made at that time. The production equipment and the processing technique are essentially the same as those previously described® for fuel solvent production. The concentrate will be produced in 15 batches of approximately 30 |b each. It has proved possible, after some simple operational changes, to process three batches of material per week. However, three weeks of down-time have been necessary during this period because of analytical difficulties and lack of approval of an additicnal uranium allotment. To date, 10 of the required 15 batches have been processed, and it is estimoted that the production operation will be completed before December 1. As the material was produced by the Y-12 Production Division for use in the ARE, analyses of the material were made by both the Y-12 and the ORNL groups involved. Any differences in these analyses were to be explained and corrected by wusing, if necessary, the triplicate sample originally taken before production operations proceeded. The analytical situation appears quite satisfactory at present, as production control and fissionable material accountability is concerned. Some discrepancies between the two laboratories have occurred in determinations of the uranium content of the UFd; these difficulties have insofar 4{\NP Quar. Prog. Rep. Sept. 10, 1953, ORNL-1609, p. 15, 51bid,, p. 15. 13 ANP QUARTERLY PROGRESS REPORT been shown to be due to foulty sampling of some poorly milled UF,. Agreement between the labo- ratories on uranium content of the Na UF, has been excellent; tentative plans call for accounta- bility transfer of the material on the basis of these analyses. A comparison of the analytical data for eight of the ten batches for which results are available is shown in Table 1.1. It should be noted here that the Y-12 laboratory used spectrographic analysis for determining the metallic impurities, while the ANP laboratory used chemical metheds of analysis. In generol, the agreement is ex- ceptionally good. One batch (AREU-4) must, on the basis of these figures, be reprocessed to lower the impurities to an acceptable level. Corrosion of Incone! G. M. Adamson W. D. Manly Metallurgy Division The comrosion of Inconel by fluoride fuels has been discussed at length in this and preceding quarterly reports. In loop tests in which the fluoride is circulated at 4 gpm by thermal con- vection, the maximum depth of attack is less than 10 mils after 500 hr at 1500°F. The recent cor- rosion dota emphasize the value of minimizing the impurities in the fluoride and, to a lesser extent, TABLE 1.1. COMPARISON OF ANALYTICAL RESULTS FOR SAMPLES OF ARE FUEL CONCENTRATE COMPOSITION ANALYSIS BATCH NO. LABORATORY Weight Fraction (g/g) Impurities (ppm) U Na F Fe Cr Ni B AREUA Y-12* 0.59534 0.1142 0.2881 25 25 35 1 {310423) ANP** 0.59521 0.1225 0.2830 20 20 AREU-2 Y-12 0.59529 0.1158 0.2865 23 11 53 0.2 {310430) ANP 0.59598 80 20 90 AREL-3 Y-12 0.59702 0.1168 0.2879 30 20 50 0.2 (310431) ANP 0.59661 75 20 20 AREU-4 Y-12 0.59470 0.1160 0.2876 150 12 7400 0.2 (310432) ANP 0.59577 220 25 475 AREU-5 Y-12 0.59637 0.1171 0.2883 14 6 21 0.2 (310433) ANP 0.59649 60 20 35 AREU-6 Y-12 0.59548 0.1174 0.2905 21 9 42 0.2 {310434) ANP 0.59582 60 20 45 AREU-7 Y-12 0.59513 35 6 67 0.2 (310435) ANP 0.59583 100 20 65 AREL-8 Y-12 0.59530 32 6 56 0.2 (310437) ANP 0.59544 75 20 50 14 * Fe, Cr, and Ni contents were obtained by spectrographic analysis. * K ANP gnalyses were made by wet chemical means. the value of clean container systems. Additives to the fluoride, such as ZrH_, which would serve to reduce the impurities in the fluoride mixtures and hence the fluoride corrosion, are being evaluated. For a detailed report of the investi- gation of the effect of various parameters on cor- rosion in the fluoride-Inconel system, as well as of the postulated mechanism, see sec. 6, “‘Cor- rosion Research.’’ PUMPS H. W. Savage W. R, Huntley W. G. Cobb A. G. Grindel! ANP Division Developmental work on the gas-sealed pumps and auxiliaries for the ARE has progressed to the point where acceptance tests of the pumps for the ARE and the mockup reliability test of a spare pump and auxiliaries are being performed. Parts for the two fuel and two sodium gas-sealed sump-type pumps are being received; one pump has been completely assembled and two others are almost complete. The pumps for the two systems will be identical except for minor differences in the impeller designs and the volumes of the sump tanks, since the fuel pump sump tank must provide volume for the addition of the fuel concentrate, as well as for the thermal expansion of the fuel. The operating characteristics of these pumps, as well as of the components, and the design of the pump auxiliary systems for oil and dibutylcarbitol distribution are being determined on a prototype sodium pump. The fuel pump auxiliary system design will be the same as that determined for the sodium pump, since the requirements of the latter will be the more severe. Prototype Sodium Pump Test The tests of the gas-sealed sump pump with NGF:«Zr1":“--UF4 (50-46-4 mole %) were reported previously.® This pump is a prototype of those being assembied for the ARE. The termination of these tests was caused by bearing noises which were determined to be the result of slippage of the inner bearing race on the shaft. The bearing race and the shaft were altered to give a 0.0002- to 0.0004-in. interference fit between them attempt to eliminate the slippage. in an Subsequent SANP Quar, Prog. Rep, Sept. 10, 1953, ORNL-1609, p. 21, PERIOD ENDING DECEMBER 10, 1933 operation of the pump with sedium was terminated after more than 300 hr ot 1200°F, because a gasket leak at the parting face of the pump casing allowed sodium to jet against the top flange near the primary gos seal. No further bearing trouble was experienced during the tests, and postrun examination revealed no bearing or shaft wear. it is therefore concluded that altering the inter- ference fit between the bedring and the shaft did successfully eliminate the slippage that occurred in the previous tests. Postrun examination showed that sodium vapor had no deleterious effect on the silver-impregnated graphite seal material. To confirm this, silver- impregnated graphite was immersed in sodium at 300°F for 150 hr in a laboratory test, and again no harmful effects could be observed. The test to determine the reliability of the secondary gas sedl at high system pressure (50 psig) was not completed because of the gasket leak which brought about early termination of pump operation. This test is now being made without sodium in the pump. It was found that gas entrained during pump start- up could be rejected in ]/2 min at a pumping rate of 35 gpm and in 4 min at a pumping rate of 125 gpm. Since there is no apparent demand for higher pumping rates during startup, the degassing charac- teristic appears to be satisfactory. Pump Auxiliaries The guxiliary systems for the ARE pumps are (1) a system for circulating light spindle oil as shaft coolant and seal lubricant and (2) a system tor circulating dibutylcarbitol, which acts as a heat barrier between the sump tank liquid and the gas seal. With the prototype sodium pump operating at 1500°F, heat loads of 11,000 Btu/hr in the dibutylcarbitol from the seal-cavity cooling annulus and 7,000 Btu/hr in the light spindle oil used as shaft and bearing [ubricant were found. This heat was removed by water in external heat exchangers. The tests conducted with the prototype pump to determine heat loads of the shati- and seal-cooling circuits showed the need for redesign of these circuits to eliminate the rotary union in the shaft- cooling circuit at the upper end of the pump shaft. The cooling circuits have been modified, and the cooling and lubricating oil now enters the bearing housing between the upper bearing and the lubricant- to-air face seal. Upon entry, the oil is divided; some of it frickles over the bearings for lubrication 15 and heat removal and the other portion passes into the hollow shaft for heat removal. All the oil collects below the bottom bearing in a cavity out- side the primary face seal. It flows from the cavity by gravity through nine Y-in.-dia connections at the bottom of the bearing T’lousing to the lubricant reservoir for cooling and recirculation. Heat exchanger tests indicate that eight pumps will be required for the pump lubrication (oil) system, thut is, one operating and one spare pump for each of the sodium and fuel pumps. One pump and its spare will suffice for the dibutylcarbitol system for all sodium and fuel pumps. These auxiliary lubrication and coolant pumps have been checked and found to be satisfactory. Impeller Fabrication All the Inconel impeller castings made by the vendor have been rejected because of imperfections such as sand inclusions, cold shots in the cast metal, and porosity. Each of these imperfections could result in pump failure because of structural failure of the impeller vanes or becouse the re- moval of sand inclusions and metal particles in the cold shots would cause intolerable unbalance of the impeller. As a consequence of these re- jections, impellers for ARE pumps are being fabricated from Inconel plate stock. Impellers for the sodium pumps (high flow) will have curved vanes to provide heat and flow characteristics similar to those of the cast impellers. Impellers for the fuel pumps (low flow)} will be drilled, and, although somewhat less efficient than the curved- vane impellers, they will be easier to fabricate. Acceptance Tests Test equipment has been designed and some of it has been placed in operation for running ac- ceptance tests on all ARE pumps. These tests will include (1) water tests to determine head, flow, and efficiency of each pump, (2) short-time, cold, run-in tests of seals at 65 psig, and (3) hot tests at ARE temperature, flow, conditions. The water test equipment is now in operation, and an all-lnconel loop is being fabri- cated for the hot tests. The hot test loop will include an ARE-type gas conirol system and the cooling and lubricating circuits described above. One ARE pump has been assembled and checked with cald water. The performance of the pump in this preliminary test was acceptable. and pressure 16 FLUID CIRCUITS G. D. Whitman ANP Division G. A. Cristy Fngineering and Maintenance Division Except for the installation of the pumps, the external fuel and sodium systems are essentially completed. Installation of heaters and insulation on these systems is about 65% complete, Additional tests of the bellows-seal and the frangible-disk valves have demonstrated the reliability -of each under ARE operating conditions. A flow diagram showing the major fluid circuits and design-point operating conditions is presented in Fig. 1.5. Fuel System One fuel pump casing has been received and installed in the fuel system; the other is expected early in December, Only seven welds remain to be made before the system, including the spare pump, is completed; some operational tests may be undertaken before installation of the spare pump. The fuel fill-and-flush tanks have been thoroughly cleaned by hot-water flushing and dried by prolonged evacuation at slightly elevated (around 200°C). temperatures Sodium System The status of the sodium system is comparable to that of the fuel system. Assembly of the sodium piping is essentially complete, except for 17 welds that are in the vicinity of either the pump casings or of the reactor. Valves The valve tests reported previously indicated that the valves to be used in the ARE would give satisfactory performance if held at a temperature of 1200°F or below. The only additional infor- mation on bellows-seal valves involves operational tests at 1100 to 1275°F after the valve had first been held for some time at 1300°F in contact with the fluoride Ncfl:fiZrF“‘-Uf"4 (50-46-4 mole %). In these latter tests, there was no leakage but some sticking; the operationat 1300°F apparently softened the sealing surfaces. The pressure required to open the stuck valves increased rapidly and then leveled off as the valve-closed time was increased; at no time, however, did the pressure exceed 30 psig. D#G. 22047 TG STACK -_...8_29_. pas - % N 7 SURGE 82A :L ; \_ TANX 855 M zmGaTe SwWitH avin. SATES v Iy o GUreoLs e h ! e ; T T LT PG 858 p30p : 7 @ P —— 1 I} \. - i ol ) :g @A =f . : Par o : Yaoin.LINE } Yo | g 5l reEa x g5z 1o ssa N4 l 5 _______ . i — r ! o 4 la o) L\ AMAAL L - 3 ’ A b 4-in LINE | IOTEST ey | W ' - a \ VYWY » | g r+_RESERVOIR lr‘"D'fl“;"“—*“ beem—dE T 2| S o3 , g @ 3 ! oI cm z A — 2 6-in. GATE p——— [ ‘ | YIR: iyl H ) 851 ot |, | ms3 . BASIN B | | Lo rt! & |l =g i@ 5 WA | i I %5 X D £ % ’e, o 4 & 3-in. VALVES | ] Nak VAPOR “lp 27 by vent 2 e —— : TRAP — & Foy NO.5,8,7, NO. 1 A & ; AND 8 T AN | N : ; P50 b g : : na ) 16-in LINE _@33,(}% ‘ w) N8 i 141-144 145148 1472143 'T:j"wux 201 WA'}ER 511 i VAPOR TRAP SUMP J —[ | i SUPPLY 505 - W) | ND.2 g! | N\ 877 i 875 2 i & = = T T se—{ 2 o LAPL i 509 Pigg [/ P85 Pi8E INEEY 92 0 33 5 - o o ul36 . i Lo Yugz g «jug3 7 /8ol 801 = Y 3 fi P193 o L8 5 o 2 & ~ w 7 u N84, ! PUMP NO 1 PUMP NO. 2 x 2 \35/ l20{68 g FILTER | 19 \ P4t - F32 ECONCMIZER = COOLER (o ™ o 23,0 2 = L v & & 2= {108 E @ ax i n g I3 0= Q: o g Wz ; ; i 7 z 2 =R TR I Frow rueL g z | HaX \ 4?*/ \ | ENRICHMENT 4 gl »iZ% 34 Lt ] 41 FES L 3 & T aEe —" Ppag > ol nl A2/ o 2lE Il - B =z T 7 z TO DRAIN s (He 2 2 @ e 5 ;\:r) EE 937 532 T2 2 {he =z 8 =1 | 5 /o | 2 | | / I / Z < Y 87y v < o (O | BULK- o) 106 " 101 8 - READ - | L { L 70 | | i : 3 [ | | ; | \ £5 - FUEL o = : - ] CARRIER I - —>—x3, 570 : 2 Do il S Loy \ o ! ' jo2 == : ! I o 28 5 DUMP /a TANK NO.3 TANK NO.2 |5 B TANK o 'Papz 56/ P20 B9 /T @ T e68 Pl FILL AND DUMPLINE 125 858 1 { 85 BLEsD Ling LJ e FLOW IN gpm FUEL L RELIEF VALVE REFLECTOR COOLANT —==3~ FREEZE VALVE EI PRESSURE IN psig a7 837 .y, et 70_’ . FCa ; 875 [ B70 gscz | 100 110120001 - = ' L . -3UT 7 ~ . : FROM He SUPPLY TRAILER . HELIUM {s) PUSH-BUTTON CONTROL 862 gz oBn | g, ' A TEMPERATURE IN °F WATER ; H‘L RESERVE MANIFOLD ' I | PRESSURE !N INCHES OF WATER, GAGE COOLANT FLOW CONDITIONS ARE BASED ON THE ASSUMPTION THAY THE COQLANT HAS THEZ FOLLOWING PROPERTIES: CEEOOG VENT SYSTEM v ) * FLOW N ¢fm ! s 12 He BOTTLES £ =240 o/ ¥t _M__ —p— —_ VALVE NORMALLY QPEN VALVE NORMALLY CLOSED THROTTLING VALVE CHECK VALVE X—D—@ REMOTE MANUAL VALVE HREMOTE MANUAL ANGLE VALVE 5 ¥— MANUAL VALVE gfl SOLENOID VALVE Fig. 1.5. Aircraft Reactor Experiment Flow Diagram. ¢ =7 TO 13 cp AT OPERATING CONDITIONS Cp=0.23 Btu/Ib°F APPROXIMATE VOLUME OF MAIN SYSTEM: INTERNAL 0.3 13 INITIAL-EXTERMAL 3.5 f+2 ENRICHING FLUID (.7 £+ mMaX, TOTAL 6.5 t3 MAX, 17 Additional data are also available from tests of the frangible-disk valves operated in the fluoride fuel. The frangible-disk valve was held in contact with molten fuel at a temperature of about 1300°F for 160 hr and was immediately by the actuator at the end of this period. The valve operation was normal in every way. These tests on ARE-type valves in fluoride mixtures have produced con- fidence in the valves that are to be used in the ARE, and no serious valve problems are anticipated in the operation of the experiment. All frangible- disk valves have been completely fabricated, tested, and installed. All the Fulton-Sylphon valves were thoroughly flushed and tested with helium for leak tightness across the valve seat. opened Auxiliary Systems The helium heat exchanger loops and the as- sociated hydraulic systems have been completed and tested in air. These tests have shown that the systems have lower pressure drop than calculated, which will result in lower operational speeds and consequently greater factors of safety (or greater heat transfer capacity). The pit for the secondary off-gas system has been completed, and about 75% of the piping for the system has been installed. The planned program for inspecting and festing of individual systems is being put info operation as rapidly as the instaliations are completed. INSTRUMENTATION R. G. Affel ANP Division The detailing of all instrumentation is essentially complete. System changes have necessitated the TEST TEMPERATURE ordering of several small components which will arrive soon. Approximately two man-weeks of labor will be required before the control room is ready for complete testing. The flux servo system was installed and tested with simulated error signals. Ninety per cent of the needed copper tubing (12,000% ft) is installed. All this tubing is to be checked for leaks. The major jobs re- maining to be done are checking and testing, REACTOR AND FLUID CIRCUIT CLEANING .. A. Mann ANP Division F. F. Blankenship Materials Chemistry Division G. A. Adamson Metallurgy Division G. A. Cristy Engineering and Maintenance Division Corrosion tests have repeatedly demonstrated the deleterious effects of surface contamination of container metals and of impurities in the fluoride mixture on the corrosive attack on the container. Accordingly, both the sodium and the fuel systems will be cleaned. The methods and procedures to be followed in cleaning these systems have been established by a group consisting of representatives from the ANP chemical, metallurgical, engineering, and ARE groups. The complete, final report of this committee has been issued, and a brief outline of the procedures is given below in the order in which they will be used. PURPOSE Reactor Fuel Tubes . Helium pressure test Flush with hot, high-velocity water Flush with distilled water Purge with helium N W Ry - . Dry by evacuation Room Leak test 150 to 180°F Remove foreign material Room Replace dirty water Room Remove water Room Complete removal of water Fuel System Without Reactor Flush components with hot water Rinse with distilled water Helium pressure test Circulate hot water with de!‘ergent7 hobs o N Drain and evacuate 150 to 180°F Remove foreign material Room Replace dirty water Room Leak test 150 to 180°F Remove greuse Room Remove dirty water 7 The detergent to be used is “*Kon Kleen;'" active ingredient, sodium metasilicate, 19 ANP QUARTERLY PROGRESS REPORT TEST TEMPERATURE PURPOSE Sodium System Without Reactor 1. Fiush components with hot water 150 to 180°F Remove fareign moterial 2. Rinse with distilled water Roam Replace dirty water 3. Helium pressure test Roam Leak test 4. Circulate hot water with defergent7 150 to 180°F Remove grease 5. Drain Room Remove dirty water 6. Rinse with distilled water Room Remove all traces of dirt 7. Alternate evacuation and helium purging Up to 600°F Removs maoisture Fuel System with Reactor, Part 18 1. Pressure test Room Leak test 2. Circulate hot water with cle'rergem7 150 to 180°F Remove grease 3. Rinse with tap water Room Rermove cleaning water 4. Rinse with distilled water Reom Remove all contamination 5. Alternate evacuation and helium purging Up to 600°F Remove moisture 6. Vacuum rate of rise Up to 600°F ! eak test 7. Hot gas test: argon in fuel tubes, helium in annulus Up to 600°F l_eak test Sodium System with Reactor 1. Alternate evacuation and helium purging 2. Circulate sodium 600°F Up to 600°F Remove moisture Clean and leck test Fuel System with Reactor, Part 1 1. Heat reactor and fuel piping 1200°F . Evacuate 1200°F 3. Circulate solvent, NGZrFS 1100°F Replace with clean solvent 1100°F FUEL RECOVERY AND REPROCESSING D. E. Ferguson F. N. Browder G. I. Cathers Chemical Technology Division Transportation of Fuel Although the basic outline of an aqueous pro- cessing method to be used for ARE fuel recovery has been established,® the method of transporting the fuel from the recctor to the processing site 8Clecming of the ""Fuel System with Reactor'' is di- vided into two parts because the high-temperature phase is dependent upon preheating the reactor with sodium. D, E. ferguson, G. I, Cathers, and O, K. Tallent, ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p. 102. 20 Above fluoride mixture melting point To fill system Clean and leak test Ready for operation will somewhat offect the detoils of the dissolution procedure. A comparative study indicated that the most satisfactory method would be transportation of the solid fuel in partially filled open-top aluminum A measured amount of the molien fuel would be poured into the cans and allowed to solidify, and the cans would be transported one at a time to the Metal Recovery Building where both can and contents would be dissolved. The use of open-top instead of sealed cans would simplify the remote- filling operation and decrease the dissolution time. None of the valuable material should be lost as o result of fragmentation of the charge if there is sufficient freeboard in the aluminum container. To illustrate the feasibility of canning molten fluorides in metallic aluminum, three 10-kg batches cans. of NGF'ZFF4'UF4 (50-46-4 mole %) were transferred to air-cooled aluminum containers. Each of the cans, two with 0.065-in. walls and one with a 0.25-in. wall, withstood the test in which the fluoride temperature was 600°C. It was discovered that the thin-walled cans should not be subjected to more than 5-in. vacuum or 5-psig pressure, Previously, cans approximately 5 in, in diameter for transporting 6.6 kg of fuel plus flush material, which would contain 0.5 kg of U235 if the mixture wiere homogeneous (or a maximum of 1 kg of U233 if the mixture were not homogeneous), were con- sidered.'® This amount of fuel would dissolve to give about 35 gal of solution. The process appears to be most attractive if 4 kg of uranium per container, which amounts to 53 kg of fuel plus flush (assuming a homogeneous mix- ture), can be transported and charged to the 500-ga! dissolver to make one 265-gal batch of solution. For this size batch, the containers would be about 9 in. in diameter and 18 in. long. This quantity would be more compatible with the size of the Metal Recovery Building equipment, and the number of containers required to transport the 80 to 100 kg of U235 would be reduced from 100 to about 25. Processing of 4 kg of U235 per batch in the existing dissolver is subject to approval by the Criticality Hazards Committee. Another method of dissolving the fuel involves introducing it to the dissolver in the molten state. If this method were used, the fuel could be trans- ported in a special can that would hold about 25 kg of U235, The advantages of this method are the Y pid., p. 103, PERIOD ENDING DECEMBER 10, 1953 rapidity of the dissolution and the decrease in the number of truck trips required for the transportation. The disadvantages are the difficulty of providing shielding suitable for use in the handling of the molten material and the excessive development work required on equipment that would have to be built into the cans, for example, heaters for remeliting the fuel if it solidified before the dissolution was carried out, temperature-measuring and devices for controlling flow. Construction of the dissolver at the ARE site would eliminate the need for constructing transportable shielding and would permit the use of less rugged heaters, but it would require too large a capital expenditure. It would also split the locations of the chemical operations, and additional operators required. instruments, would be Rate of Dissolution of ARE Fuel Based on the results of laboratory dissolution experiments, a dissolution time of 6 to 12 hr is estimated for a 53-kg charge of ARE fuel. For a 6.6-kg batch, @ dissolution time of 3 to 6 hr is estimated. A reliable value for the dissolution time cannot be calculated from the data obtained in these smali-scale experiments, since the dis- solution rate is not constant throughout the dis- solving. With a 1-in.-dia cylindrical charge weighing approximately 50 g, an initial penetration rate of 0.003 in./min was observed over o 4-min period (Table 1.2). Over the next 3 min, the penetration rate increased to (0.006 in./min, based on the initial dimensions. This increased rate is due to uneven penetration and cracking, which exposes TABLE 1.2. DISSOLUTION RATE OF SOLID ARE FUEL Cylindrical samples 1 in. in diameter tested inboiling 4 M nitric acid plus 0.67 M aluminum nitrate solution INITIAL AREA TIME IN AMOUNT CALCULATED OF SAMPLE DISSOL VANT DISSOLVED PENETRATION RATE (in.2) {min) (wt %) (in./min) First immersion 3.5 4 8.7 0.003 Second immersion 3.5 3 23 0.006 First immersion 5.2 5 13 0.005 First immersion* 3.9 2 2.2 0.002 + ) ’ Solution already contained the normal complement of dissolved fuel. 21 ANP QUARTERLY PROGRESS REPORT ln @ second experiment, a rate of 0.005 in./min was observed After the 5-min period, the sample had broken up into several smaller pieces. more surface area. penetration over a 5-min interval. Another fresh sample tested in a solution that already contained a normal complement of dis- solved fuel gave a penetration rate of 0,002 in./min; however, in the final stage of an actual dissolution, the charge would probably have broken up to expose a very large surface area per unit weight. Based on these results, the average penetration rate will probably exceed 0,006 in./min. The sample charges for these experiments were prepared by pouring molten fuel (NaF-ZrF -UF,, 50-48-2 mole %) into a split graphite mo?d and cooling. The pit formed by contraction on cooling was removed by pressing the cylinder against a hot metal plate. The sample was placed in the dis- solvant (boiling4 M nitricacid and 0.67 M aluminum nitrate solution) and then removed and weighed after a measured interval to determine the amount dissolved. Since the scale-up of these laboratory dota is inconclusive, three charges will be poured into aluminum cans, 6 in. in diameter and 9 in. long, and dissolved to estoblish the dissolution time more accurately. Molten Fue! Dissolution No excessive violence was observed in several preliminary tests in which molten ARE fuel, at temperatures of over 550°C, was poured into the aqueous solvent (4 M nitric acid plus 0.67 M aluminum nitrate). On contact with the aqueous phase, the fuel solidified in the form of small irregular particles 1 to 2 mm in diameter and then dissolved at a rate of about 50 g in 10 min, which is comparable with the rate of 50 g in 5 min that had been observed with finely ground fuel. Ap- parently the only precaution required is te make sure that the addition is slow enocugh to prevent excessive overheating and overboiling. However, colloidal material has been present in all dissolver solutions prepared todate from molten fuel, possibly as a result of the instability or hydrolysis of zirconium compounds in agueous solutions at high temperatures., Some modification of the dissolver solution to prevent formation of this material would probably be required if molten fuel dissolution were to be used. 2. EXPERIMENTAL REACTOR ENGINEERING H. W. Savage, ANF Division The completion of the developmental work on the gas-sealed sump pumps for the ARE has made possible a change in emphasis of the experimental work, and increased effort has been devoted to the development of a hydrodynamic type of bearing and a heat exchanger for the reflector-mederated reactor, The developmental work on the frozen-sodium- sealed pump for sodiuvim during this quarter included tests of a frozen-sodium shaft seal with a single ]/2-in.~|ong frozen-sodium sealing region backed up with helium pressure in the sealing cavity., The short seal is being developed in an effort to lower the power consumption of the seal and yet maintain the low leckage obtained with the 5]3’/]6-in.-long seal previously described, In tests, the new short seal dissipated opproximately 3:2 hp, with no power fluctuagtions, ond it was found that the sodium feakage could be made negligible if back-up pres- 22 sure was used to reduce the pressure differential across the seal, Water tests on the gas~sealed sump pump for the in-pile loop showed fluid gassing to be a problem, and, accordingly, chonges are being made in an effort to eliminate this condition, In other tests, the problems associated with de- veloping a conned magnetic-torque-iransmitter for a high-temperature pump for aircraft opplication are being explored, Since the adoption of the gas-sealed sump pump for ARE operation, seal developmental work has been carried on at a reduced rote; however, tests for rotary shaft and valve stem seals have con- tinved with materials which showed promise in earlier tests, These include graphite pocked around a spiral-grooved shaft, graphite and BeF, mixtures, a V-ring type seal, and a bronze-wool, graphite, and MoS, mixture, Although each of these seals has performed well under laboratory controlled con- ditions, none has demonstrated sufficient reliability to warrant adoption in a pump for reactor operation, Also, none of the packing materials tested thus far for valve.stem seals has demonstrated sufficient reliakility to replace the bellows as a valve-stem seal, ‘ A hydrodynamic type of bearing which will operate in liquid metals, fused salts, or other fluids af temperatures of up to 1500°F is being developed. Materials for this application are being screened in a compatibility testing device, and a test progrom is being planned for obtaining the bearing charac- teristics, which are usually represented as friction factor vs. Sommerfeld number. A heat exchanger incorporating some design features proposed for the reflector-moderated re- actor is currently being tested. The fluoride system has operated .at above 1200°F for gver 600 hr, but performance data are not yet available, : A small forced=convection loop is being operated to test corrosion of Inconel in circulating fluoride mixture NaF-ZrF ,-UF , at high fluid velocities and high temperature differences, The fluid velocities obtained in this loop are a factor of 50 or 60 greater than those of thermal-convection loops, but the Reynolds modulus remains in the laminar region at approximately 1700. PUMPS FOR HIGH TEMPERATURE LIQUIDS W. G. Cobb W. R. Huntley A. G, Grindeli W. B. McDonald D. R. Word ANP Division Frozen-Sodium-Sealed Pump for Sodium A frozen-sodivm shaft seal with a ],r:’z-in.-l'ong frozen-sodium region backed up with helium pres- sure in the sealing cavity was tested. This seal differed from the previously tested! short seal in that only one Y%-in.-long sealing region was used and the blanket gas was under pressure, The short seal is being developed in an effort to lower the power consumption of the seal and yet maintain the low leakage obtained with the 5‘3/16-in.-!ong seal previously described.? 'ANP Quar. Prog. Rep. Septf. 10, 1953, ORNL-1609, po 20- : 2ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p. 11. ' 3bid,, p. 17. PERIOD ENDING DECEMBER 10, 1953 For the first 168 hr of operation with the new short seal, the sodium temperature at the pump was 960°F, the suction pressure was 20 psig, and the back-up pressure was 15 psig. The total sodium leakage post the seal during this period was 32 em3 or an average of 4.8 cm?® per day, Approxi- mately %:4 hp was dissipated in the seal and the power trace was smooth, The Buna N rubber O ring used fo seal the back- up gas was destroyed during the application of heat to remove the sodium leakage at the end of 168 hr, and the remainder of the testing was ac- complished without the use of back-up pressure, Operation was continued for a total of 980 hr with a shaft speed of 1800 rpm, a sodium temperature of 960°F, and a suction pressure of 12 psig. A slight flow of helium was maintained across the back of the seal, The increase in the pressure differential across the seal from 5 psig during the first part of the test to approximately 12 psig during the remainder of the test resulted in an increase in sodium leakage past the seal to ap- proximately 20 cm? per day. Erratic power fluctuations caused by sodium oxide migrating to the cold section of the seal became negligible when a standard, cold-irap, bypass filter was installed in the loop. The filter bypasses approximately 0.25 gpm of sodium, Ex- cessive heat loss in the filter (approximately 3 kw) necessitafed its redesign to incorporate o heat economizer designed to waste only 0.3 kw. Transit time through the new filter is approximately 6 minutes, : The results of these tests, together with the previously reported results of tests with longer frozen-sodium seals, have led to the conclusion that a frozen-sodium shaft seal is manageable for a sodium pump, A satisfactory length-to-diameter ratio for such a seal is approximately 1 1o 5, Leakage of sodium past the seal can be made negligible by using back-up pressure to reduce the pressure differential across the seal, Water cooling is an efficient means of freezing the seal, and several thousand hours of operation have demon- stroted that the hazard in' using water with ¢ properly designed seal is frivial, Gas-Secied Sump Pump for In-Pile Loop Tesit A pump is being developed for in-pile loop tests which is similar to the lcboratory-sized gas-sealed pump? but which has been scaled down to reduce 23 ANP QUARTERLY PROGRESS REPORT the fluid holdup in the pump. Water tests of the first pump constructed show that the pump head and flow are within design specifications; however, a fluid gassing problem exists which indicates that some design changes will be necessary, While some gassing of the fluid is evident under all operating conditions, its magnitude increases rapidly as the height of the free fluid surface above the suction is decreased or the shaft speed is increased. This indicates that the difficulty is caused by vortexing of the free fluid surface be- cause of shaft rotation, As an initial aftempt to suppress vortexing, chaonges are being made in the antiswirl baffles in the pump, and the opening through which the rotating shaft enters the suction chamber is being moved to a point well below the free surface of the fluid, Magnetic-Torque-Transmitter Pump High-temperature pumps for aircraft application may include o canned magnetic-torque-transmitter pump with the external shaft driven by suitable means, for example, a gas turbine. The problems encountered in building such a torque transmitter are being investigated. been Reduced-scale tests have conducted for dstermining the expected efficiency of a magnetic torgue transmitter at elevated temperatures, the geometry of such a transmitter, and to what extent slippage occurs as a function of temperature. The first transmitter permanent magnet outer rotor and a squirrel cage (from a tested consisted of a ]/"m-hp motor) inner rotor, A nonmagnetic stainless steel can separated the driving and driven rotors, The entire assembly (excluding the outboard bearings) was heated to above 1100°F, and the slippage rate was appreciable in the tem- perature range of 600 to 1100°F, However, the full zu-hp torque was transmitted from room tempera- ture to 600°F, but the efficiency then dropped off rapidly at higher temperatures, It is yet to be determined whether the drop in efficiency is due to the drop of the flux of the permanent magnef or to lowered resistance of the windings of the rotor. A permanent magnet inner rotor is presently being constructed which will resolve this question, YANP Quar, Prog. Rep. Sept, 10, 1953, ORNL-1609, p. 23. Sfbid., p. 24. SANP Quuar. Prog. Rep. June 10, 1953, ORNL-1556, p. 19. 24 ROTARY-SHAFT AND VALVE-S5TEM SEALS FOR FLUORIDES R. N. Mason P. G, Smith W, C. Tunnell ANP Division Graphite-Packed Seal for Spiral-Grooved Shaft Operation of the spiral-grooved shaft with a powdered-graphite-packed seal? has continued for over 3000 hr with no detectable leakage of fused fluorides. During this periocd of operation, heater failure occurred on two occasions and caused the shaft rotation to stop. When repairs were mude, operation was resumed with no difficulty. Graphite-BeF, Pocked Seal The previously reported® test of a seal packed with @ mixture of 80% graphite powder and 20% BeF, has now operated for over 3700 hours, The leakage for the past 3500 hr was less than 15 in,3. Power fluctuations have occurred, and some of them were accompanied by squecking, which indicated possible metal-to-metal contact between the shaft and the gland, When this pocking material was installed in a tester having a 2]/4-in.-dia shaft, excessive graphite and BeF2 leakage occurred, Investigation dis- closed that the temperature during pretreatment was insufficient to fuse the Ber, and thus the seal contained loosely packed material rather than the solidly fused material desired, For the second test, the pretreatment temperature was raised to 1800°F, solidly, During the dry run of this seal, there wus and the BeF-graphite mixture fused no leakage of powder. The seal region has been operated at a higher temperature (above 975°F) in this test than in previous tests to reduce power fluctuations, There has been some fluoride leckage. V-Ring Seal Operation of the V-ring seal test® was continued, and o heater was installed inside the 21,6-inmadiq rotating shaft, During on extended dry run in which the seal was gradually heated to about 1200° F, tained, The inert gas blanket and seal gas supply was inadvertently shut off for about 20 min during satisfactory sealing of helium was at- operation, and, since some damage to the seal rings occuited, satisfactory sealing could no [onger be obtained, Examination of the rings after disassembly of the seal showed that the outside ring was 75% destroyed and the other rings had either oxidized or worn about 0.0005 to 0.0015 in, on the inside diameter; the shaft was undomaged. The outside ring has been replaced and another dry run will be made in an attempt to seal helivm at above 1000°F before fluorides are introduced. Bronze-Wool, Graphite, and Mo$2~Pocked Frozen Seal The test’ of the bronze-wool, graphite, and MoS - packed frozen seal continued 'until operation wcs terminated at the end of 1198 hr because of leakage of the fluoride mixture NaF-ZrF -UF from a Hange. The leakage rate was less than 1 cm? per day with the fluoride pot at 1175°F ond pressurized to 10 psi, : An odditional test was made with the packing annulus reduced from ¥ to ¥ inch. This test wos unsuccessful and was terminated at the end of 211 hours., The failure was attributed to faulty alignment between the shaft and the packing housing. Packed Seals for Valve Stems A series of tests is in progress to determine whether the packing materials which sealed well in the packing penetration tests® will seal high- temperature fused salts under simulated valve operating conditions. None of the materials tested has demonstrated sufficient reliability to replqce the beliows as a valve-stem seal. The test apparatus consists of a standard valve stem and bonnet placed in a vertical position with the stem down, The pot in which the molten fluo- ride mixture ‘is pressurized against the packing in the bonnet is welded to the top of the bonnet. The stem is rotated at frequent intervals to simulate valve operation, Three tests have been completed in which the valve stem was cycled once every 24 hr, and the pressure on the packing was 5 psi, The packings tested were Asbury graphite 805 ot 1100°F, 90% Asbury graphite 805 plus 10% Bsz at 1100°F, and 50% graphite (from the Y-12 Carbon Shop) plus 50% BeF, ot 1350°F. Each of these tests was terminated because of fluoride leokage. 72ANP Quar, Prog. Rep. Sept. 10, 1953, ORNL- 1609, p. 25 8/bid., p. 28. PERIOD ENDING DECEMBER 10, 1953 HIGH-TEMPERATURE BEARING DEVELOPMENT R. N. Mason P. G. Smith W. C. Tunnell W. K. Stair ANP Division Developmental work is under way on a hydro- dynamic type of bearing which will operate in liquid metals, fused salts, or other fluids at tem- peratures of up to 1500°F. This developmental program will probably be carried out in three phases: (1) test programs to find materials which have low wear and low corrosion rates in the fluids of interest at 1500°F, (2) o test program to establish the bearing characteristics, which are usually represented os friction foctor vs. Sommerfeld number, and (3) the design and test of a bearing for a particular application using the information ob- tained in phases 1 and 2. Effort to date has been confined to materials compatibility fests and to an evaluation of the accuracy which may be expected in carrying out the program for obtaining bearing characteristics, Materials Compatibility Tests Initial screening tests for finding materials that possess satisfactorily low wear and low corrosion rates are being made in apparatus in which a 23/’-m -dia, / in.-thick plote specimen of one ma- fenai rofafes against a Y.in.-dia stationary pin of another or the same material, The end surface of the pin is ground to a Y-in.-radius cylinder to approximate line contact along o radial line of the rotating plate at a mean radius of 1 in, from its center (Fig. 2.1). Contact pressure between the plate and pin is maintained by means of an external coil spring and a sliding shaft, ' A test was run of the equipment made with a Graphitar plate and an Inconel pin. A Hertz stress (estimated stress in the line of contact) of about 10,000 psi, or a 5-lb force load between the plate and pin, was used. With the test specimens sub- merged in acetylene tetrabromide to simulate the fluoride melt, it was found that the 5-lb load was insufficient for breaking through the hydrodynamic film and that the film persisted until a load of about 10 1b was applied to give o Hertz stress of about 15,000 psi. The rate of rotation was 850 rpm, which gave a sliding speed of about 7.4 fps. Having determined the load requirement, a series of correlation tests was conducted with plates of type 416 stainless steel rotating ogainst stationary 25 ANP QUARTERLY PROGRESS REFPORT UNCLASSIFIED DWG 22338 LOADING SPRING —— =3 DRIVE WHEEL | —BALL BEARINGS i[5 . e, SPINDLE SHAFT Fig. 2.1. Bearing Materials Compatibility Tester. pins made of babbit, a die steel, a high-speed tool steel, Stellite 6, Stellite Star J, ond Superoilite. In each case, the stationary pin and the rotating plate were submerged in a regulor-grade, Texaco, Regal “'A'" oil, without additives, that has a viscosity of SUS 40-44 ot 210°F, The results of this series of tests conformed with those obtained earlier at KAPL.? The first test at a high temperature was made with Graphitar rotating against Inconel in Nal- D, 8. Vail, Compatibility of Materials in Liquid Metals, KAPL-589 (Aug. 18, 1951). 26 ZrF UF , ot 1200°F for 2 hours. The Graphitar appeared to be scratched lightly and it was warped, while the Inconel specimen appeared to be burred and built up at the rubbing surface. The following materials are being screened in seesaw corrosion tests: B, ZrC + e, Mg0Q + Ni, BeO + Ni, BeQO, hot-pressed Al O/, high-density graphite, TiC + Ni, TiC + Co, GC, B,C, WC + Co, and WC + Ni. Plates and pins of most of the above materials either have been ordered or will be fabricated. Combinations of materials which possess both satisfactorily low wear and low corrosion rates based on the qualitative results of the screening tests will subsequently be tested in equipment which simulates a conventional journal and bearing. Bearing Characteristics An evoluation has been made of the accuracy of the data which will be obtained in the test program for determining bearing characteristics. Two ex- perimental approaches for obtaining the data have been reviewed. First, the test may be simply a load-life experimental study to establish reliability of a single bearing design in fluoride melts, Information gained from this approach might or might not be useful in predicting the behavior of other bearings; certainly, the information could be used only with limited assuyrance., On the other hand, the experimental study may be a performance test in which an effort is made to establish pa- rameters which will describe the bearing charac- teristics in such a way that they may be employed to determine design configurations other than the one tested, The information to be obtained by using the second approach is usually presenied as a plot of fy vs. S, where {, is the bearing friction factor (dimensionless) and §_ is the Sommerfeld variable (dimensionless): uN [ 1 ’ 0 TR\ and , (2) fo = 1i=) where i = absolute viscosity of lubricating fluid, Ibf-sec/i n.2, = journal speed, rps, unit bearing load, psi, = journal radius, in., = radial ¢learance, in., = coefficient of bearing friction. —_— 0 1"’32 i The significance of these relationships indesign work will depend on their accuracy, which can be defined in terms of the accuracy fo be associated with each of the variables. The fractional pre- cision of § .will depend upon the fractional pre- cision of y, P, N, r, and ¢, and it may be shown that as \ as \ 2 2 o 2 (3) Isn = E P+ N Py 85” ’ asn ‘ st P2 e p? aP or T as ¥ " . r? , where P, P+ . represent the probable errors n of § , pu ... . Taking the partial derivative: of §, with respect to g, N, ... and substituting in Eq. 3 gives ‘ \2 2 SN, s, 2 e [) + - — S i # N N 2 S ? S n 2 n 2 + - Py, + 4 - P; P 2 2 2 s P P P s, p N P/ PERIOD ENDING DECEMBER 10, 1953 As an example, suppose that it is desired to evaluate § with a precision of 210%. If it is assumed, for the first approximation, that each of the quantities y, N, P, r, and ¢ may be measured with the some precision, then ] = N7 (0.1) = 0.0302, and P_/x is the fractional precision of y, N, P, r, or ¢, The variables N, P, and r can be measured to even greater precision than that required for the example given above, but it is doubtful whether i or ¢ can be measured to even this fractional precision, In this case, it may be assumed that N, P, and r are measured without error, and Pp/,u and P_/c are then found to have a fractional pre- cision of : 5 (C'l) :'EII: - X It is possible to measure ¢ with this degree of precision, but p is likely to involve a much larger error; thus, if it were necessary to evaluate S, to +10%, the measuring technique for p would have to be much refined, ' Equation 5 may be used to indicate the resulting precision of S for the case in which the probable errors in y, N, P, r, and ¢ are known, As on ex- ample, a journal bearing in a fluoride mixture under the following typical operating conditions will be considered: speed, 1800 rpm; unit load, 20 psi; journal radius, 1 in.; fluoride temperature, 1200°F; radial clearance between bearing and sleeve, 0.0015 inch. At 1200°F, the viscosity of the fluo- ride mixture NaF-ZrfF -UF, is reported to be 11.8 cp plus or minus approximately 30%, or p=11.8x 145 x 1077 = 1.72 » 107¢(lb;~sec/in.?) + 30%. The expected fractional precisions for the variables are then: Px/x (Px/x)z I 0.3 0.09 N 0.0025 0.0000063 P 0.025 0.00063 r 0.001 0.000001 c 0.067 0.00449 27 ANP QUARTERLY PROGRESS REPORT and PS = ’\/(0.09) + (0.0000063) + (0.00063) + (0.000004) + (0.01796) 7 = 40,1086 = 0.3296 or 32.96% Gnd o 2 S 1.72 x 107% x 30 ] no 2 0.0015 = 1,146 £ 33% If /, (bearing friction factor) is considered in a similar manner, For the assumed operating conditions stated 7,2 above, the estimated values of the guantities f, = ; measured for evaluating f, are: 7, 1in.; R = 6 in.; RWe W, 180 Ib; ¢, 0.0015 in.; ¥, 0.5 Ib. The expected where frqcfiona{precisions for the variables are then r = radius of journal, in., P /x (Px/x)2 = frictional force of torque arm length, R, Ibf, F 0.20 0.04 R = torque arm length, in., R 0.10 0.01 W = total bearing load, Ib, , 0:00] 0.000001 c = rc:diql 'cieamnce between bearing and W 0.1 0.01 sleeve, in. 0.067 0.0049 The expression for obtaining the fractional pre- ‘ cision can then be written as and P /e — - ""f*— = \[(0.04) 4+ (0.000004) + (0.01) + (0.01) + (0.0049) = V0.0649 = 0,255 or 25.5% b and T e and the derivations represent ideal conditions, 0.5 x 12 o = Y80 % 6 » 0.0015 0.308 + 25.5% . Thus it appears that the precision of the ex- perimental effort on bearing characteristics is limited by the precision of the data available for the viscosity of the fluoride and the ability to accurately measure the radial clearance and fric- tional force. If consideration is given to the fact that the Ap/AT for the fluoride is quite large ond that the lubricant film temperature is extremely difficult to accurately measure, the precision of the experimental results will be even lower. Be- couse errors of this type are usually cumulative 28 reality may result in errors in the relationships as high as 75 or 100%, which would indicate that refinements in measuring all critical parameters may be needed. HEAT EXCHANGER TEST R. E. MacPherson H. J. Stumpf ANP Division A heat exchanger some design features proposed for the reflector-moderated re- actor is currently being tested. This exchanger, which is enclosed in a 5-in. tube furnace for pre- heating, has the molten fluoride mixture NaF-ZrF (53-47 mole %) circulating in the she!! ond high- incorporating SURGE TANK EM PUMP PERIOD ENDING DECEMBER 10, 1953 UNCLASSIFIED DWG. 22339 Fig. 2.2. Heat Exchanger Test Apparatus. temperature sodium circulating through the tube bundies (Fig. 2.2}. A sump-type pump continuously circulates the fluoride mixture; the sodium flow is maintained by an electromagnetic pump. The sodium is resistance heated, and it enters the tubes in one half the exchanger to heat the fluo- rides. It is then cooled in a sodium-to-air radiator {100 kw, interrupted-fin type) from which it enters the tubes in the other half of the heat exchanger to cool the fluorides and then returns to the sodium heating coil. Thus both the fluoride mixture and the sodium are alternately cooled and heated. The heat exchanger is designed to transfer 1.5 mega- wafts, but performance to date has been limited by an inability to supply sufficient power to the resistance heater for the sodium. The fluoride system has been in operation at above 1200°F for over 600 hr, and the completed bifluid system has been operating at 1200 to 1400°F for 300 hours. Performance data are not yet available. FORCED.CIRCULATIOM CORROSION LOOP D. F. Salmon ANP Division A smail forced-circulation corrosion loop was operated with NaF~ZrF4-UF4 to extend the cor- rosion data being obtained from low-velocity thermal convection loops. The aim is to eventually obtain a fluid velocity of 5 fps or greoter with o temper- ature difference of 200°F. Circulation is obtained with an Eastern Industries, E-100, centrifugal pump that was redesigned for sump-type operation and fabricated of lInconel. . A gas seal is main- tained in a water-cooled, Teflon-packed gland on 29 ANP QUARTERLY PROGRESS REPORT the shaft. The loop was fabricated of 0.1-in.-OD, 0.025-in.-wall Inconel tubing., Heoting is accom- plished by passing an electric current through 17 ft of the tubing, and cooling is by natural convection to air from 9 ft of tubing. The tubing is coiled to reduce the space requirement. The fluid velocities obtained in this loop are a factor of 50 or 60 greater than those of thermal convection loops, but the Reynolds modulus re- mains in the lominar region at approximately 1700. The loop operated approximately 200 hr with o gradual slowing down of flow, which indicated possible plugging. A summary of the results is presented graphically in Fig. 2.3, which shows velocity and temperature difference as a function of elapsed time. The speed of the pump, which was checked periodically, did not vary appreciably. The velocity of the fluoride mixture was calcu- lated, rather than measured, from heat balance data on the loop heating section, nated becouse of the indicated plugging ond is being examined metallurgically. A second loop is The loop was termi- being construcied. UNCLASSIFIED DWG. 22340 3 — | | MAXIMUM TEMPERATURE, 1500°F O] | T WM’O\\ VELOCITY | 420 ! o1 ‘-0."\,"“0\ 2 | e O . 400 o o —— 380 E (&) 8 ¢ oo 360 — bes 1 _________ i / ] q A — gl o E— AT S \ AT ——1 340 \/ //‘MMM“A ;#;//A —1 320 0 | L 300 0 12 24 36 48 60 72 84 96 108 120 132 144 OPERATING TIME (hr) Fig. 2.3. Velocity and Temperature Difference as a Function of Operating Time for NaF-ZrFA-UFd in an Incone! Forced-Circulation Corrosion Loop. 30 PERICD ENDING DECEMBER 10, 1953 3. REFLECTOR.MODERATED REACTOR DESIGN STUDIES R. W. Bussard A. P. ; Fraas ANP Division In May 1953, the U, 5. Air Force requested that a series of design studies be prepared to outline the characteristics of 100-, 300-, oand 600-megowatt circulating-fuel reactors. Unfortunctely, the re- actor physics work essential for establishing the over-all design limitations has lagged far behind schedule, and mony materials questions are not yet solved. However, even though detailed designs cannot be prepared at this time, it has been recog- nized that design information is urgently needed, and therefore the preliminary layouts presented in this report were prepared to give the best estimate of preliminary designs that can be made now. While interest was expressed in reactor power outputs of as much as 600 megawatts, a careful examination of the basic {imiting factors indicated that it would not be prudent to extrapolate the information currently ovailable beyond a reactor size of 300 megawatts. The design for the 600- megawatt reactor was therefore dropped, ond studies for 50- and 200-megowatt reactors, in addition to the 100- ond 300-megawatt reactor designs re- quested, were prepared. | The more important parameters for the deSigns are presented in Tables 3.1 and 3.2. Layouts for 50-, 100-, and 300-megawatt reactors are presented in Figs. 3.1, 3.2, and 3.3, respectively. Detail drawings for the main heot exchanger, pumps, and core shells have been prepared for the 50-megawatt reactor. The designs for the 50- and 100-megawatt reactors appear to be quite conservative so that it seems probable that their ratings could be increased to 75 and 150 megawatts, respectively., Parameters such as flow rates and temperature drops would, of course, increase accordingly. An explanation and qualification of the more important features of the principal components and the considerations that determined them has been prepared to supplement the drawings. The static physics, moderator cooling, pumps, hydrodynamics, heat exchangers, pressure shell, control, and fuel addition and drainage systems are each dlscussed, in turn. An effort has been made to include in these discussions an indication of the status of the design ot this time aond of the experimental work planned to provide badly needed design data. A famthrxtty with information published previously is presumed. 172 REACTOR PHYSICS A wide variety of reactor types and geometries has been considered for circulating fluoride fuels. Of these, the most promising ot this time appears to be a reflector-moderated reactor with a central island. Sodium-cooled beryllium seems to be the best choice as the moderating material. However, the picture is very incomplete becouse calculations for only about 25 reflector-moderated reactors had been completed until recently because of lack of computing facilities. Calculations for approxi- mately 100 two-region reactors have just been completed, and the results ore being analyzed; calculations for about 200 three- to nine-region reactors have been started.. By April 1954, this work should be far enough along to give a more complete picture of the effects of reactor geometry and materiols of construction on parameters such as critical mass and power distribution for every type of circulating-fuel reactor that has appeared promising. A particular effort will be made ta find an arrangement that will give the most rugged, reliable, and simple reactor construction consistent with limitations imposed by corrosion and gommo- heating effects. Five critical assemblies have been run to date on reflector-moderated reactors. All these have been rather rough approximations to what currently appears to be a promisingfull-scale aircraft reactor, The wedkest element in all of these approximations has been the fuel region composition. In the first critical assembly, the fuel region was simulated with cans’of sodium interspersed with disks of metallic urenium, while in the subsequent assem- blies, cans of powdered NaF-ZrF, -UQ, were employed. - The first assembly suffered from severe inhomogeneity and self-shielding of the fuel, and the subsequent assemblies hod the shortcoming A P, Froas, C. B. Mills, and A, D. Callihan, ANP Quar. Prog. Rep. March 10, 1953, ORNL-1515, p. 41, 25, P. Fraas, ANP Quar, Prog. Rep. Sept. 10, 1953, ORNL-1609, p. 31. 31 ANP QUARTERLY PROGRESS REPORT TABLE 3.1. PRINCIPAL DIMENSIONS OF A SERIES OF REFLECTOR-MODERATED CIRCULATING-FUEL REACTORS Power, megawatts 50 109 200 Core diamater, in. 18 18 20 Power density in fuel, kw/cm> 1.35 2.7 3.9 Pressure she!l outside diemeter, in. 48.5 50.6 56.4 Fuel System Fuel volume in core, 3 1.3 1.3 1.8 Core inlet outside diameter, in. 10 10 11 Core inlet inside diameter, in. 7 7 7.7 Core inlet araa, in.2 40 40 49 Fuel volume in inlet and outlet ducts, £13 0.4 0.4 0.5 FFusl volume in heat exchanger, £ 1.25 2.5 5 Fuel volume in pump and plenum, fr3 0.3 0.3 0.5 Total fuel volume cireulating, £43 3.25 4.5 7.8 Fuel expansion tank volume, f13 (8% of system volume) 0.26 0.36 0.62 Fuel Pumps Fuel pump impeller diameter, in. 5.75 7 8.5 Fuel pump impeller inlet diameter, in. 3.5 4,5 5.5 Fuel pump impeller discharge height, in. 1.1 1.5 1.8 Fuel pump shaft center line to center line spacing, in. 20 21 22.5 Plenum chamber width, in. 14.5 15 15.5 Plenum and volute chamber fength, in. 30 31 33 Plenum and volute chamber height, in. 2.0 2.0 2.4 Impeller rpm 2700 2700 2500 Estimated impeller weight, b 8 12 17 Impeller shoft diameter, in. 1.5 1.75 2 Impeller overhang, in. 12 13 14 Critical speed, rpm 6000 6000 5200 Sodium Pump Na pump impeller diometer, in. 3.4 4,1 5.0 Na pump impeller inlet diameter, in, 2.4 2.9 3.5 Na pump impeller discharge height, in. 0.75 0.9 1.1 Na expansion tank volume, f13 (10% of system volume) 0.08 6.09 Na in Be passages, 3 0.43 0.47 Na in pressure shell, £ 0.15 0.15 Na in pump and heat exchanger, £43 0.20 0.25 Fuel-to-NaK Heat Exchanger Heat exchanger thickness, in. 1.7 2.75 4,65 Heat exchanger inside diameter, in. 42 42 44 Heat exchanger outside diameter, in, 45,4 47.5 53.3 Heat exchanger volume, £ 6 10 20 Angle between tubes and equatorial plane, deg 27 27 27 Number of tubes 2304 3744 6600 Tube diameter, in. 0.1875 0.1875 0.1875 Tube spacing, in. 0.2097 0.2097 0.2119 Number of tube bundles 12 12 12 Tuke arrangement in each bundle 8x 24 13 x 24 22x 25 32 300 23 3.9 62.0 2.7 12.8 67 0.7 7.5 1.0 11.9 0.95 10 6.75 3.2 27 17.5 37 3.0 2300 24 2.25 15 5000 5.9 4,2 1.2 5.9 47 58.8 30 27 9072 0.1875 0.2094 12 28 x 27 PERIOD ENDING DECEMBER 10, 1953 TABLE 3.1. (continued) Moderator Region Volume of Be plus fuel, fr3 Velume of Be only, £43 Mo. of coolant holes in reflector No. of coolant holes in island Na coolant tube inside diameter, in, Na coolant tube wall thickness, in. Mo pressure shell annulus thickness, in. that the atomic density in the fuel region was only about 33% of that of liquid fluoride fuel., These shortcomings will be largely eliminated in a new critical experiment that is being planned. Sheet Teflon and 0.002-in.-thick uranium foil will be stacked in laminge to simulate the fluoride fuel, both from the standpoint of atomic density and of neutron-scattéring and absorption cross sections. Another wedk point in the critical assemblies that have been run has been that the geometric approximation left much to be desired. It is planned that the sheet Teflon and the wranium foil will be cut fo appropriate lengths to permit the construction of a configuration essentially the some as that envisioned for the full-scale reactor. FUEL COMPOSITION AND PROPERTIES It was assumed in the design work that the physical properties of the fluoride fuel at 1500°F would be similar to those of Naf-KF-LiF-UF, (10.9-43,5-44.5-1.1 mole %), which are, opproxi- mately, the following: heat capacity, 0.45 Btu/1b°F; thermal conductivity, 2.6 Btu/hrft:°F; viscosity, 3 cp; density 2.0 g/cm®. Heat transfer considera tions show thot such a fuel would be much better than the sodium-zirconium fuel that has been pre- pored for the ARE. If such a fuel were not de- veloped, it would be necessary to increase the heat exchanger size given in the accompanying designs by a foctor of 2 or 3, This would mean an increase in shield weight of about 6% and o serious increase in heat exchanger header complexity, There are two possibie ways of obtaining a satisfactory fuel. The first is to use the above- mentioned fluoride, or some modification of it, in a container fabricated of an alloy that has no chro- miuvm. This should eliminate the serious short- coming of this particular fluoride mixture, that is, its tendency to form low-conductivity films of a highly insoluble potassium-chromium-flyoride mix- 22,4 22.4 25.8 31,5 21.1 21,1 24.0 28.8 208 208 554 554 86 86 210 210 0.155 0.187 0.187 0.218 0.016 0,016 0.016 0.016 0.125 0.125 0.187 0,200 ture on surfaces in Inconel systems. Initial tests show Hastelloy B (80% Ni—20% Mo) to be a possi- ble container material for this fluoride, and it will be tested in the coming months, A second way to obtain a fuel with satisfactory physical properties is to test various melts that have compositions that seem promising, and such o program is now under way, Melts of seven dif- ferent compositions containing lithium, sodium, rubidium, and beryllium are being prepared, and their physical properties are being determined. The best of these melts will be selected for cor- rosion test work, MODERATOR REGIONS Heat tronsfer analyses indicate that a sodium- cooled beryllium moderator. is superior from the heat transfer standpoint to any other arrangement considered thus far.? Static and seesaw Inconel capsule tests show virtually no signs of corrosion or mass transfer at 1500°F for beryllium specimens immersed in sodium. However, it is felt that mass transfer in a larger-scale test loop operating with a substantial temperature differential and with the beryllium presenting the hottest surface in the system {(as in a full-scale reactor) is very likely to be serious. A test loop to investigate this possibility has been designed, and it is scheduled to operate about February 1954. transfer prove to be serious, canned beryllium will be tested. . It is expected that the results of the multigroup reactor calculations will indicate what further moderator designs should be prepared and analyzed from the heat transfer standpoint to investigate the possibilities of exploiting lead and bismuth as moderator coclants. lead and bismuth are Should mass 3R. W. Bussard et al., The Moderator Cooling System for the Reflector-Moderated Reactor, ORNL.-1517 (to be published). 33 ANP QUARTERLY PROGRESS REPORT interesting because of their high densities. Some metallurgical work is currently under woy in on attempt to find a system that is satisfactory with respect to corrosion and mass transfer. Some structural problems exist in designing the suppoit of the reflector and the island. At first glance, it would appear that thess assemblies hang from the flat decks that span the pump region. This would not be the case when the fluoride system had been filled, however, because the be- ryllium density (1.84 g/cm®) would be somewhat less than the expected fuel density (about 2.0 a/cm®) and hence the moderator region would tend to float in the fuel. With the moderator region TABLE 3.2. HEAT TRANSFER SYSTEM CHARACTERISTICS FOR A SERIES OF REFLECTOR-MODERATED CIRCULATING-FUEL REACTORS REACTUR POWER, megawotts Fuelsto-Nak Heat Exchanger and Related Systems Fuel temperature drop, °F NoK temperature rise, °F Fuel AP in heat exchanger, psi NaK AP in heat exchanger, psi Fue! flew rate, |b/sec NaK flow rate, tb/sec Fuel flow rate, cfs NaK flow rate, cfs Fuel velocity in heat exchanger, fps Fuel flow Reynolds number in heat exchanger NaK velocity in heat axchanger, fps Over-all hect transfer coefficient, Btu/hr 2O F Fue!-NaK temperoture difference, °F Sndiumsto-Nak Heot Exchanger and Relatad Systems Na temperature drop in heat exchanger, °F NaK temperature rise in heat exchanger, °F Na AP in heat exchanger, psi NaK AP in heat exchanger; psi Power generoted in islond, kw Power generated in reflectar, kw Power generated in pressure shell, kw Na flow rate in reflector, lb/sec Na flow rate in island and pressure shell, lb/sec Total Na flow rate, |b/sec Ma temperature rise in pressure shell, °F Na AP in pressure shell, psi Na temperature rise in island, °F Na AP in island, psi Na temperature rise in reflector, °F Ma AP in reflector, psi Me-Mak temperature difference, °F Shield Cooling System Power ganerated in 6-in. lead layer, kw Power generated in 24-in. HZO layer, low 34 50 100 200 300 400 400 400 400 400 400 400 400 35 51 61 75 40 58 69 85 263 527 1,053 1,580 474 948 1,896 2,844 2.1 4,2 8.4 12.6 10.5 21.0 42.0 63.0 8.0 9.9 1.2 12,2 4,600 5,700 6,700 7,000 36.4 44.9 50.9 55.4 3,150 3,500 3,700 3,850 95 100 110 115 160 150 150 150 100 150 150 150 7 7 7 7 7 7 7 7 500 1,000 2,000 3,000 1,700 3,400 7,500 11,200 190 350 500 620 53 72 154 234 22 28 51 76 75 100 205 310 28 39 30 26 4 6 6 6 72 ni 120 124 32 21 12 12 100 150 150 15C 36 27 18 18 A3 110 210 300 350 <3 <6 <12 <18 _ g ‘ R U o ”\ g i v % A £ Fog 'y / ! f/) Na~TO-NoK i/ { HEAT EXCHANGER - / } il HEADER FOR Na TO / REFLECTOR . ~ REFLECTOR Neo Be ISLAND (% OUTER CORE SHELL -1 INNER CORE SHELL -7 \\ QUTLET HEADER s_\. Fig. 3.1. 50-Meguwati Reflector-Moderated Reactor. %A i AT PERIOD ENDING DECEMBER 10, 1953 p L a” W A g " e — FOR PUMP ASSEMEBLY SEE FIG. 3.4 —-8'% LavER -~ INCONEL JACKET RN Be REFLECTOR ----—-MODERATOR - COOLING TuBE A~ COOLING TURE 1 CONNECTOR --—TUBE BUNDLE SPACERS LAYER 10 8™ LAYER - PRESSURE SHELL LINER 35 ANP QUARTERLY PROGRESS REPORY "/ / HEADER FCR Na TO QUTER CORE SHELL ‘ : \ ' REFLECTOR Na |, OUTLET HEADER- I oM T E R w 3 IT S o £ lgé‘ S = o /7 REFLECTOR - ] * . INNER CORE SHELL ~ l»fi ‘ I A DWG. 22342 — FOR PUMP 25SEMBLY ¥~ SEE FIG 3.4 = |--UFPER PUMP DECK _-LOWER PUMP DECK LA .- - B'C LAYER Al AN W A~ INCONEL JACKET \ - COOLING TUBE CONNECGTER /T JACKET £OR B0 ; LAYER g0 | AYER T PRESSURE SHELL LINER T Na PASSAGE - Fig. 3.2, 100-Megawait Reflector-Moderated Reactor. 36 PERIOD ENDING DECEMBER 10, 1953 LWG 22343 T FOR PUMP ASSEMALY SEE FIG. 3.4 /U PPER PUMP DECK ~LOWER PUMP DECK ~REFLECTOR SHELL Na-TO- NoK 0 B LAYER HEAT EXCHANGER™ TUBE. - v SINCONEL JACKET HEADER FOR Na TO REFLECTOR .-Be REFLECTOR CONTROL ) ROD PASSAGE - TUBE BUNDLE SPACERS MODERATGR COOLING TUBE CONNECTOR Be ISLAND-— TUBE BUNDOLE INNER ~JACKET FOR CORE SHELL~— i B"Y LAYER OUTER - u'0 L avER WA\ CORE SHELL-. \ T PRESSURE SHELL LINER “Na PASSAGE REFLECTOR Na OUTLET HEADER- - PRESSURE SHELL NoK OUTLET Fig. 3.3. 300-Megawatt Reflector-Moderated Reactor, 37 ANF QUARTERLY PROGRESS REPORT buoyed up by the fuel in on almost ideal ‘‘shock mounting,”’ occelerations that would occur under crash londing conditions (in which vertical ac- celerations of as much as 10 g should be antici- pated) would yield a net upward force on the reflector assembly that would be about equal to its weight, In the designs illustrated, the re- flector assemblies are attached to the lower deck of the pump region by welds to the walls of the sodium return passages. The sodium pump und heat exchanger subos- semblies are positioned around the top edge of the upper pump deck. The pipe from the sodium pump discharge makes a slip fit into the reflector sodium inlet tube. The leakage through this slip fit into the sodium return passage simply re- circulates with no penalty other than o small increase in the required pump capacity. The two sodium pump and heat exchanger assemblies could be welded in place by passing a bead around the lower edge at the outer periphery and up and across the top where they adjoin the plenum chamber between the two fuel pumps. The reactor designs illustrated in Figs. 3.1, 3.2, and 3.3 presume that canning of the beryllium will be required but that trace leaks in the Inconel- to-can connections can be tolerated. Therefore the tubes passing through the rifle-drilled holes in the reflector are designed to be driven into tapered bores in the fittings shown at the equator, and the outer ends of these same tubes are to be rolled into their respective header sheets at the top and botiom. The tube connecting fittings at the egquator also serve as dowels to locate the two beryllium hemispheres relative to each other. HYDRODYNAMICS OF THE FUEL CIRCUIT Hydrodynamic considerations impose a number of resiraints on the geometry of the fuel passages in the reactor fuel circuit. These restraints are not sharply defined, but they are sufficiently rigid that they limit the fuel circuit proportions for a well-balanced design to a considerably narrower range than would at first be expected. The pres- sure drop through the hecot exchanger goes up very rapidly with fuel flow rate —~ so much so that there is little to be gained by going to fuel velocities which correspond to pressure drops in the heat exchanger of greater than 80 psi. This is particu- larly true if on effort is made to keep the stresses in the core shells and heat exchanger tube walls 38 to 500 psi or less, If the rest of the fuel system is examined carefully, it is evident that it should involve a pressure drop very much less than that ocross the heat exchanger in a well-balanced design. Thus, since o bleed-off-air turbine-pump- drive arrangement seems to be at once the lightest, simplest, and most flexible and since such a pump drive system should be designed to require less than 3% air bleed-off from the compressors, there seems to be o strong incentive to limit the pressure drop through the pump inlet and discharge passages and the core inlet end outlet ducts to not more than 15 psi. If it is assumed that the reactor core diameter is determined by a compromise between shielding and other considerations and that the islond diometer will be determined by fission density distribution considerations, it becomes evident that these two basic dimensions for the fuel circuit moy be token as o base for determining the proportions for the rest of the system. The inlet and outlet duct passage arrangements should be as small as possi- ble to simplify the shielding problem and to reduce shield weight. However, it is of paramount im- portance that boundary-layer thickening and flow separation be avoided in the diffuser region at the inlet to the core. This, in turn, imposes a lower limit on the flow passage area that may be used ot the core inlet. An asymmetric design might be employed in that this flow limitation would not apply to the outlet duct, particularly since a diffuser giving a reasonably high effi- ciency can probably be provided between that duct and the inlet to the heat exchanger. It is quite evident from this stondpoint alone that further shielding tests should be made to investigate the effect of fuel duct size on over-all shield weight. While the problem is not clearly defined, it does appear from flow test work, as well as from analy- ses of the adverse pressure gradients along the diffuser walls in the core inlet region, that the outer diameter of the core inlet duct should be roughly 60% of the core diameter, while the inner diameter of that duct should be between 60 and 70% of the outer diameter, The plenum chamber between the volute dis- charges and the core inlet ducts should provide for smooth fransition and uniform velocity disiribution at the inlet to the core. One promising way of accomplishing this would be to provide a plenum chamber having a height somewhat greater than the thickness of the core inlet duct and a diameter at least half again as large as that of the inlet duct. The pump volutes could then be designed to give a fluid velocity entering the plenum chamber about equal to that in the core inlet duct, This would mean that for high power density reactors, for example, a 200-megawatt reactor with an 18-in.-dia core, the di scharge annulus of a mixed flow impeller should be made as wide as possible so that it would approach the height of the plenum chamber. For low power density reactors, for example, a 50-megawatt reactor with an 18-in.-dia core, the disparity in these dimensions is inherently large. A careful examination of the magnitude of the eddy losses to be expected of the pump inlet, through the impeller, in the pump volute, and in the plenum chamber leads to the conclusion that the pump should be designed for on inlet velocity of not much more than about 30 fps, that is, «a velocity giving a dynamic head of about 12 psi. On the basis of these considerations, it is possible to specify dimensions for the principal fuel flow possages for a wide range of reactor power outputs. While the dimensions are given in Table 3.2, it should be noted that there is some latitude in the selection of these dimensions. Furthermore, it is clear that passages of larger size would require substantially heavier shields, while passages of substantially smaller size would be likely to give some sort of difficulty from the hydrodynamic standpoint. Some preliminary flow tests were carried out a year ago,‘s- and the results were encouraging. Additional flow tests ore being made to determine whether the proportions indicated in Table 3.2 lead to a satisfactory configuration for a reactor de- signed for an output of 50 to 100 megawatts. PUMP DESIGN In examining the over-all system requirements, it becomes evident that it would be very desirable if the fuel pumps could perform several functions in addition to that of pumping the fuel. It seems likely that most of the xenon and possibly some of the other fission-product poisons might be removed from the fluoride mixture if helium could be bubbled through the fuel while it was being thoroughly agitated. Furthermore, it would be advantageous 4R. E. Ball, Investigation of the Fluid Flow Pattern in a Model of the *Fireball"’ Reactor, ORNL Y-F15-11 (Sept. 4, 1952). . ' PERIOD ENDING DECEMBER 10, 1953 to have the pumps serve as mixing chambers in which high-uranium-content fuel could be added 1o the main fuel stream to enrich the mixture and compensate for burnup. Finally, it is essentidl that on expansion tank be provided sc that en- trainment. of bubbles would not prove fo be a problem in maneuvers or in “bumpy’’ flight. The pump arrangement shown in Fig. 3.4 waos worked out to provide for all the points mentioned above. The pump design embodies o varigtion of an idea tested three years ago.® The back of the impeller is used to centrifuge out the gas bubbles. The design also uses the overhung impeller, the bearing arrangement, and the face-type gas seal of the previous pump, along with several features that have proved to be successful in sump pump test work during the past year, The principles of opera- tion are quite simple and have been demonstrated with a plastic model. The fuel in the expansion tank above the impeller tends to swirl in the same direction as the impeller (but at a lower speed) so that it forms a vortex with o roughly conical free surface. A relatively large amount of Fluid leaks from the high-pressure region at the impeller outlet through the labyrinth seal ot the periphery of the impeller into the swirl chamber immediately above the impeller. Since this swirl chamber is connected through a bypass line to the impeller iniet, the leakage is simply recirculated at no cost other than an increase in pumping power. Some fluid lecks upward through the clearance between the top lip of the swirl chamber and the outside of the centrifuge cup on the back face of the impeller, but this leckage rate is small because there is little pressure differential across this gap. The leakage that does occur is made up by bleed flow from small holes in the periphery of the centrifuge cup on the back of the impeller. Thus there is a small, continuous flow from the fluid vortex in the expansion tank into the ceatrifuge cup and thence into the swirl chamber, The fluid region close to the surface of the vortex in the expansion tank is highly turbulent ond full of bubbles so that it should be effective in degassing the fuel, In initial tests of o water pump with the above- mentioned features, the pump continued to pump when it was inverted for as much as a minute before the liquid supply in the centrifuge cup and Pump Reworked to Test Special Features for Operation with Liquid Metals, ORNL Y-F15-6 (Feb, 1, 1951}, 39 ANP QUARTERLY PROGRESS REPORT UNCLASSIFIED DWG E-16919A SECTION A-A PUMP HOUSING -—— 4 11— - OlL SUPPLY PASSAGE L3 -~ Ol BATH LOWER BEARING ANC SEAL SUPPORT‘n\i s IMPELLER SHAFT LOWER BEARING -~ LA OiL BATH-—---- FACE-TYPE GAS SEAL.- ——-¥§ A= - PUMP WELL CASING 4™ COQLING CIL PASSAGE ——NATURAL URANIUM LA o T ¥ TUBE ——-ri FUEL PELLET SUPPL , “CSLINGER Ik EXPANSION TANK CENTRIFUGE CUP LABYRINTH SEAle SWIRL CHAMBER - —m [# =7 PUMP INL Fig. 3.4, Gas-Sealed Aircraft Pump. 40 swirl chamber was depleted by fluid ledkage from the swirl chamber inte the expansion tank. The slinger ring on the impeller shaft at the top of the expansion tank acts to throw the fiuid away from the shaft when the pump is inverted and thus pre- vent flow into the clearance around the shaft between the slinger ond the face-type gas seal. A number of other special features have been included in the pump design to adapt it to the full- scale reactor shield, The pump has been designed so that it con be removed or installed as a sub- assembly with the impeller, shaft, seals, and bearings in a single compact unit. As can be seen from Figs. 3.1 and 3.4, this assembly fits into the bore of a cylindrical casing welded to the top of the reactor pressure shell, A 3-in. fayer of uranium just above a'l’z-in. layer of B'? around the lower part of the impeller shaft is at the same level as the reactor gomma shield just outside the pressure shell. The steel seal and bearing support above the loyer of uranium gives additional gomma shield- ing, while the oil-filled cavity between the bearings acts as neutron shielding. An air-driven turbine and geor box could be mounted on the flange at the upper end of the pump casing. Since the outer surface of this casing was designed to be about 8 in. inside the outer surface of the shield, there need be no protuberances from the shield surface in the final installation. The long, slender, “vill supply’’ tube for fuel enrichment is suffi- ciently smail thot it should not permit much gamma edkage. The large amount of impeller shaft overhang inherent in the design ofthe pump could be avoided if a journai bearing could be operated with the molten fluoride mixture acting as the lubricant. The pump impeller and volute could be designed so that the radial load on such a bearing would be very small, perhaps not more than 25 psi. Since the viscosity of the fluoride mixture is greater than that of water, the arrangement should be completely satisfactory after the pump was storted. Pickup and scuffing of the bearing or shaft during startup or shutdown might prove to be problems. A test rig for investigating the compatibility of different materials in sodium at high temperature has been built and is being used to evaluate the compati- bility of various material combinations in fluorides (cf., sec. 2). Preliminary results are encouraging. PERIOD ENDING DECEMBER 10, 1953 PRESSURE SHELL As much as 1/‘2% of the reactor output may go into gamma heating of the pressure shell. The provi- sion for the removal of this heat presents several design choices, none of which is entirely satis- factory. If no provision is made for cooling the pressure shell, it will be much hotter than the fluoride mixture that contacts its inner surface. Unfortunately, this would give the highest pressure shell temperatures in precisely the zone in which the stresses are the highest, that is, in the ligo- ments between the NaK ducts in the lower header region. The pressure shell might be cooled with the coolant system that serves the lead gamma shield, but this would make temperature control difficult over a wide range of power outputs because of the inherently large temperature differential. Furthermore, it would make necessary the dumping of large amounts of heat at a low temperature level through the shield coolantradiator — a costly matter in an airplane. Cooling of the pressure shell with sodium from the island moderator cooling system has the disadvantage of complicating the pressure shell construction, but it has many advantages. It provides a simple way of returning the sodium that has passed through the island cocling passages to the sodium-to-NaK heat exchanger. It not only serves to cool the pressure shell in the critically stressed fuel discharge end of the reactor so that the creep strength of the Inconel there becomes mote than odequate, but it makes it possible to maintain the entire pressure shell at an essentially vniform temperature. This should markedly reduce both thermal distortion and clearance buildups resulting from differential thermal expansion be- tween the pressure shell, heat exchanger tube bundles, and the reflector. The consfruction envisioned is shown in Figs. 3.1, 3.2, and 3.3. The inner liner assembly con- sists of a %-in.-thick Inconel shell, a %-in. hot- pressed B'? layer, and an immer 0.062-in.-thick Inconel can. Inconel is also used to cover the interior of the pressure shell between the two end headers. Closely spaced grooves, 1@ in. deep and 2 in. wide, milled into the outer surface of the liner provide omple flow possage areo for the sodium to flow upward. Four transfer tubes carry the sodium from the top of the pressure shell to the outer face of the pressure shell ot the sodium-to- NaK heat ‘exchanger inlets. The hot-pressed 4] ANP QUARTERLY PROGRESS REPORTY blocks are diamond-shaped, with 60-deg angles at their vertexes, and they have rabbeted edges. This design permits the covering of a spherical surface with a single block size and shape, The construction in the vicinity of the end header regions was deliberately varied in Figs. 3.1, 3.2, and 3.3 to show three different ways in which the detail design may be handled. Another design would involve the use of a pressure shell with forged annular protuberances to receive the headers, while yet another possibility would be a shell sufficiently thick (obout 2.5 in.) to permit machining annular grooves for the headers. This latter arrangement would have the advantage of a more regular ex- terior surface on the pressure shell. boron FUEL-TO-NoK HEAT EXCHAMGER The heat exchanger envisioned for the reflector- moderoted reactors is similar to that described previously.® The major unknowns in the fuel-to- NaK heat exchanger are concerned with fabrication. It is presumed that the tube-to-header connections wouldbe heliarc welded by hand, as in the 210-tube six-tube-bundle fluoride-to-NaK heat exchanger presently being tested. It is expected that a large amount of welding research and component testing will be required to obtain a satisfactory solution to this tube-to-header welding problem. The tube bundles are separated by long cans that are triangular in cross section and parallel the tubes. The cans would be filled with hot-pressed B'0 to assist in inhibiting the low-energy neutron flux in the heat exchanger region. REACTOR CONTROLS One of the salient features of the circulating-fue! reactor is its inkerent stability. It appears that no fast-moving control rods will be required to cope with fast transients. No large amounts of shim control are required to compensate for burnup, be- cause a wuranium-rich flueride can be added at intervals during the course of operation. The only important control needed is a means of varying the mean operating temperature. |ln examining over-oll power plont considerations, it appearad that provi- sion for effecting a temperature change of 200°F should be included in the design (that is, about 6A. P. Fraas and M. E. LaVerne, Heat Exchanger Design Charts, ORMNL-1330 (Dec. 7, 1952). 42 3% Ak/R). |t appears from critical experiments that this could be provided by « 3/-m.-OD tube filled with B9 and placed to give a Q?-lnn stroke from the pump deck down through the center of the island. While the rod ond its actuating mechenism are not shown in designs illustrated, a 0.50-in.-ID tube has been provided through the center of the island to receive the rod. In the design contem- plated, the rod wculd be machined from a thick- walled tube to give three equally spaced ]/1 “in.- high ridges along its entire length. These ridges would serve tc center the rod in the passage pro- vided and thus give uniform flow of sodium coolant around the rod periphery. The ridges would also engage a grooved guide at the upper end to prevent rotation. A short length of s/s-in,,-'IO thread ot the top of the rod would be engaged by an internally threaded tube enclosed in g casing extending up through the shield. The tube would be rotated by a drive at the top end through a bellows szal. This seal would be brazed to the center of o bar with three spherical bearings, one at each end ond one in the center. This bar would be inclined ot on angle of about 15 deg to the axis of the control rod drive tube. The center bearing would be fixed relative to the rod casing, while a crank attached to the upper end of the control rod drive tube would be attached to the lower spherical bearing. As the rod drive tube rotated, the axis of the drive bor would generate o cone, with the centers of the spherical beorings at its upper and The drive actuator would be an electric motor and gear box connected to the upper spherical becring. The closed end of the bellows at the center bearing would describe a wobble-plate type of motion. lower ends generating circles. The fue! enrichment system thot appears most attractive is based on the use of cold-prassed disks of No,UF . about ]/ in. in diameter and / in. high. These dlsks could be fed to the tops of the fue! addition tubes in the pumps by meens of arelatively simple “’pill vending'’ machine. Several promising detail designs of such machines have been pre- pared from which it appears that this phase of the development should not prove to be a serious problem. However, it may be difficult to prevent the fue! inthe expansion tank from entering, freezing, and blocking the fuel enrichment tube during in- advertent inverted flight., The pump design has been worked out so that this should not happen, but the effectiveness of this feature of the swirl expansion tank design has not yet been checked. SHIELDING A substantial amount of fundamental work on the shielding problems of the reflector-moderoted reactor has been covered in other reports,?+7.8 Three important factors remain undetermined: namely, the energy spectrumof the short half-fived fission-product gammas, the effect of variations in endduct and in fuel pump andheader tank geometry, andthe possibilities inthe use of unusual materiols such as lithiumhydride. Tests have been carefully planned to provide the essential design data neces- sary for determining the first two of these factors during the coming months, and a long-range program is contemplated for investigating unysual materials. it is hoped that a shielding experiment that is to be carried out in conjunction with a critical experi- ment will yield data thot will validate the use of multigroup, multiregion calculations as an aid 'in shield analysis. ' ' Analyses of the effects of the neutron-to-gamma dose ratio and of varying proportions of gamma shielding on shield weight are being made. The strength of lead and bismuth alloys at temperatures of 250 to 350°F, shield design for jet fuels in place of water, shield heat removal, and the dose from the shield after shutdown are also subjects currently being investigated, FILLING AND DRAINING OF THE REACTOR After 100 hr or more of full-power operation, the fuel charge will require roughly 10 in. of lead shielding to bring the rodiation from it down to a level of 1 r/hr at 5 feet. This will mean that a tank to contdin the roughly 4 3 of fuel required for a 50-megawatt reactor will require about 10 tons 7 Report of the 1953 Summer Shielding Session, ORNL.- 1575 {to be published). 8k, H, Abernathy et af., Lid Tank Shielding Tests of the ReflectorsModerated Reactor, ORNL-1616 (to be published). PERIOD ENDING DECEMBER 10, 1953 of lead shielding. This could be mounted on © heavily constructed dolly or on an elevator de- signed to raise it up from a pit under the runway. A cooling system to remove the afterheat from the fuel will also be required. Preliminary estimates indicate that the aofterheat can be removed most readily by circulating a liquid metal through a coiled tube in the fluoride drain tank and then through a radiator to dump the heat to air forced through the radiator by a blower. Such a system should be reasonably light, compact, and mobile. The filling and draining operation could be carried out by carefully positioning the airplone over the tank ofter removing a vertical plug, perhaps 5 in. in diameter, from the bottom of the reactor shield. The shielded tank assembly with a pipe extending vertically upward from it could be raised so that the pipe would project into the hole in the shield. A coupling at the upper end of the pipe would engage o corresponding coupling at the bottom of the reactor pressure shell. The space between the drain valve at the bottom of the reactor and the valve ot the top of the pipe from the drain tank could then be evacuated, the drain tank valve and thereactor drain valve opened, and the fuel drained. A blast of helium directed through appropriate fittings could be used to blow out the fuel droplets between the valves. The valves could then be closed, the tank disconnected, lowered, and dropped into its pit or removed to a suitable location. . The reverse procedure could be followed in the filling operation. Helium pressure on the ligquid in the shielded tank would serve to force the fuel up into the reactor, | The biggest problem in the design of a system of this sort is the detail design of the valves and the remotely operated coupling to get the high degree of leak tightness and exceptional reli- ability required. Several rough preliminary layouts have been sketched and work is proceeding in an effort to develop valves and couplings that could be used for routine filling and draining operations in current ORNL expetimental laboratory work. 43 4. CRITICAL EXPERIMENTS A, D. Callihan D. V. P. Williams H. Lynn Physics Division J. W. Noaks, Pratt and Whitney Aircraft Division The Critical Experiment Facility at the Oak Ridge National Laboratory may be used to de- termine the static physics characteristics of a wide variety of reactor mockups. During the past quarter; measurements were completed on a mockup of an air-cooled, water-moderated reactor (AC-1) for the General Electric Aircraft Nuclear Propulsion Project, and preparations were made for measurements on a mockup of a Pratt and Whitney supercritical-water reactor, a Nuclear Development Associates sodium-cooled reactor, and the ORNL's reflector-mederated reactor (cf., sec. 3, '‘The Reflector-Moderated Reactor’’). The air-cycle reactor (AC.1) mockup was an approximated cylinder 27 in, in length, 30 in. in diameter, surrounded by a 7]/2-ft-thick beryllium reflector. The fuel elements and the uranium dis- tribution were simulated by properly spaced urenium disks sandwiched between mild steel which repre- sented the structural materials of the reactor. Each fuel element was surrounded by a 1-in.-thick layer of Plexiglas to simulate the woter moderator. The originally designed assembly was an octagonal array of 37 fuel channels. This system, however, could not be made critical with the 7.5-in.-thick berylliumreflector until the number of fuel channels had been increased to 43 ond five channels had been loaded with twice the uranium prescribed, The critical loading was then 29.8 kg of U235 with 98.9 kg of steel. Flux and power distributions have been determined throughout the core and re- flector. The data will be reported in detail in a topical report to be issved by ORNL when the experiment is complete ond are discussed in General Electric’'s Engineering Progress Reports. - . P . - : ! - ; ; » - - . : ‘ x i ; ¢ INTRODUCTION AND SUMMARY The research on high-temperature liquids for reactors has been primarily concerned with the determination of phase diagrams of fluoride and chloride systems with and without fissionable material, although a substantial effort has been devoted to the production and purification of halides and hydroxides (sec. 5). Detailed study of the NaF-ZrF4—UF system’ has continued be- cause of its:intendej application in the Aircroft Reactor Experiment, wherein Na,UF will be added to Ndszs to produce a suitable fuel. Techniques such as quenching, petrographic and x-ray analysis, differential thermal analysis, and high-temperature filtration, as well as the customary thermal analyses, have been applied to the binary and ternary systems NaF-ZrF4 and NaF-ZrF -UF . Spectrophotometric measurements of fused salts containing UF . show only two absorption-spectrum patterns in the NaF-ZrF[UFé system. Some physical property data substantiate the concept that the fused salts are essentially totally ionized. A number of fluoride samples were purified in the laboratory for various tests while the facilities for the routine production of these fluoride mixtures are being constructed. In an examination of various phases of fluoride purification techniques, it was found that thereductionrate of N02UF5 by hydrogen increases with temperature. The reducing power of NaH when added to NaZrF:5 is being studied, as well as the kinetics of the H, reduction of Nitt in NaZrF_. Chemical studies of hydroxides were limited to the purification of NaOH ond 5r(OH)., and the determination of the carbonaceous content of NaOH at 700°C. The recent corrosion studies have been devoted almost entirely to the effects of various parameters on the corrosion of Inconel by fluorides, although some work with hydroxides and liquid metals was continued (sec. 6}. Studies of the corrosion of inconel by the tluoride fuel NQFS-Zer-UFd (50-46-4 mole %) have substantiated earlier conclusions that the total attack in 500 hr is independent of temperature from 1500 to 1650°F and that the attack rate is time dependent. The value of having purified fluoride melts in clean container systems has been investigated not only with regard to the metallic fluoride impurities in the melt, the concen- tration of which may be directly correlated with corrosive attack, but also with regard to contami- nation of the container surface, which has a less deleterious effect. Pretreatments of the fuel with chromium and Inconel; as well as the precirculdtion of fuel, have each been effective in reducing attack, and an observed correlation betweendepth of attack and surface-to-volume ratio of the Inconel and fluoride also indicates that the impurities in the fluorides are the major factor in corrosion. The attack by fluorides on Inconel at 1500°F apparently increases in intensity, although not in maximum depth, as the UF content of the meltis increased from O to 6.5 mole %. Static fests of various combinations of metals in fiuorides ot 1500°F have shown that one of the metals is attacked to a greater extent than the other, even though the heavily attacked metal may be relatively unattacked when alone in the fluoride. The fluoride fuel NaF-ZrF4-UF4 (50-46-4 mole %) has also been circulated in nickel and type 430 stainless steel loops for 500 hr at 1500°F; no plugging occutred, but some mass transfer was observed in the stain- less steel loop. A number of metal specimens were tested in circulating lead in quartz therma! con- vection loops. Of the metals tested, oxidized types 347 and 304 stainless steel and molybdenum- nickel alloy produced the least mass transfer. Other liquid metal studies included tests of inconel, stainless steels, and potential bearing materials in lithium, sodium, and lead. The more fundamental corrosion studies included ' the determination of the pressure of and an equilibrium hydregen in hydroxide metal attempt to determine the chemical equilibria for the postulated fluoride corrosion which results in the selective leaching of chromium from Inconel. systems The fabrication of high-gerformance radidgtors and high-conductivity radiator fins represents the major undertaking of the metallurgical research program, which also included stress-rupture tests of Inconel, the forming of special alloys, the fabrication of solid fuel elements, and the develop- ment of ceramic materials (sec. 7). Although Inconel-clad copper fins are inadequate because of diffusion of nickel into the copper, acopper- aluminum bronze fin and types 310 or 346 stainless steel clad copper fins have been fabricated .and satisfactorily brazed to Inconel or stainless steel tubing with a number of special brozing alloys. The technique of “*backing-up’’ the heliarc-welded 47 ANP QUARTERLY PROGRESS REPORT radiator joints with a brazing alloy produces leak-tight radiators. The stress-rupture various combinations of sound, life of Inconel temperature, stress, and environment is being measured; the tests at 3500 psi and 1500°F show that, as far as rupture life is concerned, air is the most beneficial environment and hydroger the most deletericus. unider Stress-rupture data are also being obtained for lnconel tubes and for fine- and coarse- grained Incone! sheet. The limited effort on solid fuel elements has been confined to the drawing of tubular fuel elements and the plug welding of steel-clad fuel plates, both of which were effected with some success. Incone! tubing for experimental equipment is being drawn satis- factorily with reductions as high as 36%. The rolling of various metals and ulloys was effected: columbium sheet after an 87.5% reduction at 1100°F showed complete recrystallization. The physical properties of several fluorides and stainless metal alloys have been measured at temperatures of up to 1000°C, and the heat transfer characteristics of reactor coolants are being studied in various systems (sec. 8). The viscosity and density of the ARE fuel concentrate, Na UF , and the vapor pressure of the fuel solvent, NchFs, have been measured. All these values are in line with those anticipated for these materials. Other heat capacity, thermal conductivity, density, and vapor pressure data for fluoride compositions of interest have been obtained. Measurements of the velocity in a thermal convection loop indicate that the radial gradients cause turbulence in the liquid flow at Reynolds numbers above 100; the laminar tlow solutions for flow in this regime are thus in serious error. Forced-convection heat transfer data with the NafF-KF-LiF eutectic in nickel and in lnconel tubes have been confirmed; the heat transfer in Inconel is less than that in temperature nickel because of film formation. Preliminary data are available on a forced-flow heat transfer experi- ment in the laminar flow regime, and apparatus is 48 being prepared for measurements of fluid flow in an annulus and for studies of surface-boiling phenomena. The radiation damage program included studies of the stability of fluoride mixtures in Inconel and determinations of the creep of metals under irradi- ation, as well as the construction of the in-pile circulating loop (sec. 9). Although [nconel capsules containing fluorides show more tendency toward intergranular corrosion under irradiation than they do in out+of-pile control tests, there is no definite trend in corrosion behavior that can be correlated with either irradiation time (from 53 to 810 hr) or irradiation level (from 230 to 8000 watts/cm3). There were no in-pile creep tests made in the LITR or in the MTR during this quarter, because more reliable apparatus for creep measurements in both these irradiation facilities was being developed and constiucted. The in-pile circulating loop, types of which will be used for forced-convection corrosion tests in the LITR aond in the MTR, is being fabricated and assembled. The analytical studies of reactor materials in- cluded chemical, spectrometric, and petrographic anclyses of fuel composition and corrosion products (sec. 10). The chemical analyses were primarily concerned with the oxidation states of the con- stituents and the metallic corrosion products in the fluoride mixtures. Techniques were established for determining the amounts of FeF_, Cer, and UF_ in fluoride fue! mixtures, as wefil as the reducing power of NaZrF_. None of a number of solvents tested were as effective for dissolving the proposed fluoride fuels from the various walls of metal containers as the nitric and boric acid mixtures whichare currently used. Petrographic examination of a numbker of irradiated and wun- irradiated fuel samples did not reveal any dif- ferences in the two species. A technique has been developed by which the mass spectrometer may provide a measurement of the uranium burnup (due to irradiation), as well as a quantitative analysis of the uranium in the fuel. PERIOD ENDING DECEMBER 10, 1953 5. CHEMISTRY OF HIGH-TEMPERATURE LIQUIDS W. R. Grimes Materials Chemistry Division Exploratory investigations of phase equilibria in fused salt systems by the simple thermal onalysis techniqgue have been extended to a number of chloride systems containing UCI, and UCl, as the fissionable components, as well as to examination of several fluoride systems with ond without uranium that are of possible interest as fuels, fuel solvents, or coolants. In addition, the thermal analysis technique has been applied to a number of previously studied systems in attempts to verify some doubtful values in various regions or to amplify data for regions in which previous experi- mentation was limited. ' During the past several months, other techniques useful for more refined studies of systems pre- viously shown to be valuable have been developed and perfected. These techniques, especially quenching, with petrographic and x-ray examination of the solids, differential thermal analysis, and high-temperature filtration, have been applied, in general, to the binary and ternary systems NaF-ZrF, and NaF-ZrF -UF,. While it is not yet possible to interpret even the binary system uniquely, it appears that combining the data ob- tained by these techniques and the thermal analysis data will permit rapid progress to be made., Spectrophotometric measurements of the absorp- tion spectra of UF, and UF, in fused salts were made. Only two absorption spectrum patterns were found for glasses from the NaF-ZrF -UF, system, Glasses were obtained with compositions of between 16.65 and 50 mole % UF, in the ZrF UF binary system. Additiondl electrolyses of various solutions of metal chlorides in potassium chloride were made. Erratic results obtained with chromium anodes have been ascribed tentatively to oxidation of the chromium. When a thin plate of nickel was used to protect zirconium electrodes from oxidation, reproducible emf’'s were observed. Measurements of the density and electricadl conductivity of the fused salt system LiCi-KCl are indicative of considerable persisting in these high-temperature fluids; these and other measurements substantiate the concept thot fused salts are essentially totally ionized. shorf-range order Facilities for the production of a variety of fluoride compositions are being prepared and laboratory~scale production is to be discontinued, except for the preparation of small enriched- uranium-bearing specimens for radiation stability tests. A number of small fluoride samples were purified for phase equilibria studies, plans were made for producing the enriched material for the in-pile loop experiments, ond various simple structural metal fluorides were prepared for corro- sion studies, In a study of the reduction of Na, UF by hydrogen it was found that the rate of reduction increased as a function of temperature. A further check on the possibility of retaining reducing species in NaZrF. melts was attempted by treating NalZrF, with NoH, and the kinetics of the reduction of Nit* in NaZrF, by H, at 800°C was studied be- couse of the |mportance of this reduction in fluoride puritication procedures. Studies of the reaction of sodium hydroxide with carbonaceous matter suggest that it would not be possible to maintain sodium hydroxide free from carbonate at high temperatures in metals which contain oppreciable quontities of carbon, THERMAL ANALYSIS OF FLUORIDE SYSTEMS C. J. Borton L. M. Bratcher J. Truitt Materials Chemistry Division Additional data that are believed to be more reliable than some obtained sarly in the phase study program were obtained on a number of binary and ternary fluoride systems. Revised diograms will be published in topical reports when the studies are completed. Although some data for compositions in the NaF.-ZrF -UF, system were obtained by the therma! analysis technique, it is expected that the major emphasis will be placed upon other techniques for study of this system. MaF.ZrF -UF, The probable existence of a ternary eutectic NGF-ZrF4~UF4 (63.5-19.0-17.5 mole %} with a melting point of 800°C was mentioned in the pre- vious report. Closer study of this region has 49 ANP QUARTERLY PROGRESS REPORT shown that this composition is probably near the bottom of a very narrow valley extending down toward the chszé-NaZrF eutectic. Additional data between the 40 to 50 mole % NaF lines at several uranium concentration levels essentially confirmed the location of the isotherms in this region that are shown in the published diagram for the system.! Data for the 75% NaF mixtures are shown in Fig. 5.1. These data indicate solid solution formation between Na,ZrF, and Na,UF, over a wide range of compositions, and they have been confirmed by other methods (cf., subsection on ‘‘Filtration Analysis of the NaF-ZrF -UF, System”’). Thermal data obtained from sealed capsules of ternary mixtures containing more than 75% NaF have, in general, indicated melting points that were lower than had been anticipated for this region and lower than had been indicated by the few runs made previously, Further study of this system by other metheds is planned. 1. M. Bratcher, R. E. Traber, Jr., and C, J. Barton, ANP Quar. Prog. Rep. Sept. 10, 1952, ORNL-1375, Fig. 41, p. 78. RbF-LiF-UF, Data for the system RbF-LiF-UF , were published in a previous report.?2 Additional data obtained during this quarter confirmed the existence of two low-melting-point regions in the ternary system, one close to the RbF-LiF binary eutectic and the other in the vicinity of the LiF-UF, eutectic. These data indicate that revision of the isotherms shown in the published diogram is needed. [t appears to be rather difficult to obtain reproducible thermal data in this system, possibly because of the oxygen introduced into the mixtures by the hygroscopic alkali fluorides. NaF-ThF Thermal analysis of the system NaF-ThF, supplemented by petrographic study of a number of slowly cooled compesitions, continued during this quarter, It seems to be a difficult matter to prepare 2J. P, Blakely et al,, ANP Quar. Prog. Rep. Sept. 10, 1951, ORNL.-1154, Fig. 13.3, p. 159. DWG. 22251 ‘ | 800 |- N SO D SN OUR e nPA7 // )/ @ 750 b L ey g L-’/ w / T > o < 700 - )/ ,,,,,,,,,,,,,,,,,, L ! o i = o 650 & B0 |reeeeerem L — 550 1 N [ L NagUF, 10 20 30 40 50 60 70 80 90 NayZr F, NazZr F; (mole %) Fig. 5.1. The System N03UF7-N03ZrF7. 50 compounds in this system in the pure, well-crystal- lized condition needed for petrogrophic identifi- cation. Twocrystalline phases have been observed, and one of them is believed to be Na,ThF , the identity of the other is still uncertain, Further study will be required before a satisfactory equi- librium diagram for the system can be prepared. LiF-RbF-BeF, Thermal data were obtained in the system LiF- RbF-BeF ., for 50 compositions with up to a BeF, concentrotion of 60 mole %. The lowest melting point observed with these mixtures was 345°C for material containing (in mole %) 37% LiF, 8% RbF, and 55% BeF,. The composition of this material is close to that of the LiF-BeF, binary eutectic reported to melt at 356°C.% It appears unlikely that compositions with melting points substantially lower than 350°C exist in this system, except possibly in the high Bef, region which is not amenable to thermal analysis; materials with such high BeF, concentration would probably be of fittle interest in the reactor program because of their probable high viscosity, These results are similarto those reported previously for the KF-LiF- Be2 system.? RbF-BeF 2" er'z‘1 The system RbF-BeF -ZrF, was investigated with ZrF . concentrations of up to 50 mole % and with Bef, concentrations of up to 60 mole %, Halts in the cooling curves at temperatures near 300°C were observed for several compositions in this system. |t is likely that the lowest thermal effects observed can be attributed to a solid-phase transition, since the RbF-BeF, and RbF-ZrF, binary systems showed thermal effects in the 300 to 350°C temperature range which were almost certainly due to such transition. No liguidus temperatures significontly lower than 380°C (the minimum melting peint in the RbF-BeF, system) were observed with the compositions tested. THERMAL ANALYSIS OF CHLORIDE SYSTEMS C. J. Barton S. A. Boyer Materials Chemistry Division The UCI; used for most of the phase studies of chloride systems made this quarter was prepared 3D. M. Roy, R. Roy, and E. F, Osborn, J. Am. Ceramic Soc. 33, 85 (1950). PERIOD ENDING DECEMBER 10, 1953 by the Y-12 Chemical Division by hydrogenation of UCl,. The melting point (840°C) obtained for this material agrees well with the melting point of 842 + 5°C reported by Kraus.® Uranivm and chloride analyses of the as-received material checked with the theoretical values very closely; however, slight hydrolysis of the material occurs on heating, even when considerable precaution is taken to avoid exposure. LiCl-UCI, Thermal data obtdined with ¢ number of composi- tions in the system LiCl-UCl; are shown in Fig. 5.2, The only eutectic in the system contains approximately 25 mole % UCl, and melts at 495 £ 5°C, This is apparently the lowest melting alkali chloride~UCI, binary eutectic, The thermal data gave no indication of compound formation in this system, but, since the solid phases have not yet been studied, the absence of incongruently melting compounds is not completely demonstrated. NaCI-UCl, An equilibrium diagram for the system NaCl-UCI, was reported by Kraus,® who stated that there were no compounds formed by these components. Thermal data obtoined with nine mixtures containing 33 to 85 mole % UCl, are in reasonable agreement with Kraus' data. However, the absence of the eutectic break in the cooling curves for the mixtures con- taining 65 mole % UCI, or more seems to indicate the possibility of compound formation in the high UCI; region. A thermal effect at 415 £ 5°C through- out the region studied here, which was not reported by Kraus, could be due to « solid transition of a compound in the system. KCl-ucl, Krous also published an equilibrium diagram for the system KC]-UCI3.5 Thermal analysis of the system in this laboratory with mixtures containing 12 to 65 mole % UCI3 has essentially verified Kraus' data. The maximum difference in liquidus temperatures in the two studies was about 15°C, 4., M. Bratcher and C. 1. Barton, ANP Quar. Prog. Rep. March 10, 1953, ORNL-1515, p. 113. S¢. A Kraus, Phuse Dioyrams of S5oeme Complex Salts of Uranium with Halides of the Alkali and Alkaline Earth Metals, M-251 (July 1, 1943). 51 ANP QUARTERLY PROGRESS REPORT 900 (—- ‘ 700 fromveeee o 2. PRI o o E » 2 800 Ll 0. = wl — 500 400 fo e e 300 term— e i ‘ LiCl 1O 20 30 40 50 60 70 80 90 Ucty UCl3 (mole %) Fig. 5.2. The System LiCl-UCi.. RLCI-UCI Preliminary data indicate that there are two eutectics in the system RbCI-UCl;, one ot about 15 mole % UCI,; that melts at 605 £ 5°C and another lying between 40 and 50 mole % UCI; that melts at 515 £ 5°C, The compound Rb,UCI, melts congruently at about 730°C, while Rb,UCI appears to melt incongruently at approximately 565°C, CsCl-ucl, Only a few thermal measurements have been made on the system CsCl-UCI,, but there is indication that there are probably only two eutectics in this system. One eutectic has a high CsCl content and melts at 580°C, and the other contuins approxi- mately 47 mole % UCI; and melts at about 540°C. KCi-LiCl-ucl, The few studies that have been made on the system KCI-LiCl-UCl, indicate that the addition of about 4 mole % UCI; to the KCI-LiCl eutectic ; lowers the melting point from 355 to 345°C, but 52 further additions of UCI, result in a sharp increase in melting point that is probably due to formation of the high-melting-point K2UC|5 complex. Study of this system will be resumed at a later date. RbCI-UCI,, Preliminary data for the system RbCI-UCI, indi- cate that Rb,UCI, melts congruently at about 630°C and thot the eutectic in the 40 to 50 mole % region melts at approximately 340°C. CsCl-ucl, Preliminary data for the system CsCl-UCI, were presented in the previous report.® A tentative equilibrium diagram for this system, based on the available thermal data, is shown in Fig. 5.3, The compound Cs,UCI,, with a melting point of 657 t 5°C, is believed to be the only congruently melting compound in this system. There is some evidence for a compound such as CsU,Cly or ®R. J. Sheil, 5. A. Boyer, and C. J. Barton, ANP Quar, Prog. Rep. Sept. 10, 1953, ORNL.-1609, p. 59. PERIOD ENDING DECEMBER 10, 1953 800 700 600 |— 500 TEMPERATURE {°C) 400 300 oo s o R ~. L.SZUC!E 200 IR DWG 22253 o ] ______ SRRSO SR S N U Oy T O — & Cs{l 10 20 30 40 50 €0 70 80 90 UCi, Ucl, {(mole %) Fig. 5.3. The System CsCI-UCI . CsU,Cl,, that would melt incongruently at about 382°C, which is only a few degrees above the melting point of the eutectic containing approxi- mately 58 mole % UCI . KCl-Licl-ucl, Preliminary data reported’ for the system KCI- LiCI-UCI, indicated a eutectic of unknown compo- sition that melts at approximately 275°C, Further investigation of the system during this quarter confirmed the existence of the low-melting-point eutectic, which appears fo contain (in mole %) approximately 45% KCl, 20% LiCl, and 35% UC|54. However, a more complete study of the system will be required for determining the eutectic composition with more certainty. 7R, J. Sheil and C. J. Barton, ANP Quor, Prog. Rep. June 10, 1953, ORNL-1556, p. 42. NaCl-LiCl-UCH , The very complete investigation of the system NaCl-LiCl-UCl, made during this quarter failed to show a ternary eutectic with ¢ melting point significantly lower than that of the lowest melting NaCl-UCi, binary eutectic (375°C). Since this investigofion covered the region where low melting compositions are believed most likely to occur, no further work with this system is contemplated ot the present. [t is interesting to note that when small amounts of UCI, (2 to 10 mole %) were added to the NaCI-LiCl binary eutectic, the resulting ternary mixtures had lower melting points than those of the binary mixtures. The smallest addi- tion of UCI, (1 mole %) to the KCI-LiCl eutectic resulted in a higher melting point. This difference in behavior probably can be atiributed to the higher melting point of the K,UCI, complex as compared with that of the Na,UCl, complex, 53 ANP QUARTERLY PROGRESS REPORT QUENCHING EXPERIMENTS WITH THE NoF-ZrF , SYSTEM R. E. Thoma J. Truitt G. D. White Materials Chemistry Division Barton Moore C. J. R. E. Recent improvements in previously described® techniques ond equipment for quenching small samples of fused fluorides have greatly improved The technique of determining equilibrium conditions in solid phases by quenching o composition to glass ond then heating to the temperature of interest to grow crystalline phases in the glass gives promise of solving some of the more difficult problems in the NaF-ZrF, system. Only preliminary results with this technique have been obtained. It is opparent that very occurate temperature control is necessary to establish phase relation- ships in the 40 to 60 mole % ZrF, region of the NaF-ZrF , binary system in which most of the quenching work has been done. The improved equipment in use at present controls and measures temperatures to t0.5°C. The two furnaces now in operation make possible the quenching of about 40 samples a day. the reproducibility of results. Qil has proved to be as effective a quenching medium as mercury and it is much more convenient becouse it eliminates the necessity of having to attach it to the capsules. The formation of crystals during quenching has been kept to a minimum through the use of small samples (about 3 mg) and by severe pressing of the portion of the capsule that contains the sample to decrease the sample thickness. However, quench growth is still some- times troublesome in the NaF-ZrF, system. An- other problem that is not yet completely sclved is the presence in most gquench samples of small amounts of an unidentified phase believed to be a complex sodium zirconium oxyfluoride. However, the practice of hydrofluorination of small samples (cf., subsection on ‘‘Production of Purified Fluo- rides’’) and the handling of the finely ground samples in vacuum dry boxes have been of value in keeping these phases at a minimum. Quenching of specimens containing 33 to 46 mole % ZrF, reveals a cubic crystalline phase, Because the phase is isotropic, petrographic 8R. L. Moore and C, J. Barton, ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL-1609, p. 59. 54 identification is difficult; x-ray diffraction tech- niques will be required to establish the liquidus temperatures, The cubic matericl is believed to be a form of Na,ZrF , but the possibility that it is Na,Zr,F | has not yet been eliminated. Material containing4émole % ZrF , shows NaZiF ¢ os the primory phase. The liquidus temperature is 512 10 515°C, and the sclidus temperature is 510°C, Devitrification at 506°C of glass of this composi- tion gave NaZrF . and a small amount of the cubic material. Material containing 48 mole % ZrF, shows as its primary phase on easily crystollized speciss believed to be o high-temperature form of Na,Zr F o (designated R3). In the mixture containing 30 mole % Zrf the primary phaose is R3; the liquidus temperature is near 518°C; and the solidus temperature is slightly above 510°C. Slow cooling of mixtures with this composition invariably gives a phase referred to as NoZrFS and No3Zr4F]9. However, devitrifi- cation of the glass of 506°C provided only crystal- line NaZrF, with some glass residue. It now appears that solid solutions of Na,ZrF, in NaZrF exist which show optical properties that differ only slightly from those of NaZrF .. The devitrifi- cation experiments seem to indicate that NaZrF, is the stable solid phase at this concentration, Material containing 53.8 mele % ZrF, shows R3 as the primary phase; the liquidus temperature has not been established, but it is above 517°C; the solidus temperature is about 511°C, BothNaZr F o and NaZrF . ore present below the solidus tempera- ture. Mixtures contoining 57.2 mole % ZrF | correspond to the formula Na,Zr F .. A glass devitrificotion experiment produced an almost pure crystalline phase with this composition. Although this single experiment apparently confirms the formula, further experimentation at this and neighboring composi- tions will be required to establish beyond doubt the composition of the compound. DIFFEREMTIAL THERMOANALYSIS OF THE NaF-Zr?4 SYSTEM R. A. Bolomey Materials Chemistry Division The equipment used for differential thermo- analysis was modified during this quarter by re- placing the capsules that had thermocounles welded to the side with small copsules that have center thermocouple wells, With the new capsules, much better agreement is obtained between heating and cooling curves, and, for most samples, the break for the liguidus transition is sharper and better defined. Heating and cooling rates of 1°C per minute can now be employed satisfactorily for most samples. Date have been obtained with these capsules for mixtures in the system NaF-ZrF , that were hydrofluorinated before use. A break at about 512°C for samples containing 57 to 60 mole % ZrF4 suggests that the composition of the phase which precipitates from 50 mole % material is probably above 57 mole %. It is possible, however, that the existence of the 512°C break in thisregion is due to o solid transition; this point will be established by combining the data obtained from differenticl analysis with that obtained by using quenching techniques. Data are being collected in the 40 to 50 mole % ZrF, range, and, when ex- tended to higher ZrF concen_frafions, a complete phase diagram will be presem‘ed, FILTRATION ANALYSIS OF FLUORIDES C. J. Barton R, J. Sheil Materials Chemistry Division The high-temperature filtration method of study- ing phase equilibriac mentioned in the previous was applied mainly to NaF-ZrF - 4~UF, Two afiempfs to mixfures were unsuccessful be- repor‘t9 compositions during this quarter. filyer UF“--,Z:’F4 cauyse the high vapor pressure of Z¢F , resulted in separation by sublimation rather than liquid-solid separation. in the NGF-ZYF4-UF4 system, the filtration method provided useful information about an important fuel composition in addition to data for fundamental studies of phase relation- ships in this system. Mixtures with 53 mole % NaF The melting point of o mixture containing (in mole %) 53.5% NaF, 40.0% ZrF4, and 6.5% UF‘1 has been reported as 545°C., However, in g previous filtration at 560°C of a 7 mole % UF, mixture, analysis of the residue and of the filtrate some segregation of o high-uranium- content material. This experiment has been re- peated carefully with a mixture containing {in mole indicated 9R. J. Sheil and C. J. Barton, ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL-~1609, p. 61, PERIOD ENDING DECEMBER 10, 1953 %) 53.0% NaF, 40.0% ZrF,, and 7.0% UF,, which was filtered at 604, 575, crnd 560°C. At the two higher temperatures, there was no detectable dif- ference between the filtrate and the residue, when examined either petrographically or by x-ray dif- fraction. At 560°C, more than 99% of the charge material passed through the filter, but examination of the residue indicated that it was a higher ura- nivm solid solution than the filtrate. There was not enough residue obtained in this experiment for a chemical analysis, and therefore the significance of the apparent phase separation is not clear. In one filtration carried out at 525°C with the same composition, 83 wt % of the material filtered. Both filtrate and residue were found to be solid solu- tions of the type No(U)ZrFs, with the residue hav- ing the hlgher uranium content. It is considered significant that no trace of free UF, was detected in the solid phases. Mixtures with 50 mole % NaF Thermal data for NaF-ZrF -UF , mixtures'® con- taining 50 mole % NaF seemed to indicate that NaUF o end NaZrF, form solid solutions. Petro- graphic and x-ray diffraction studies have shown that solid solutions do occur in this system but that UF, is probobly the primary phase that sepa- rates from the mixture containing 50% NaF, 25% UF,, and 25% ZrF . Filtration experiments ‘with 50% NaF mixtures contammg 10, 12.5, 25, 31, and 37.5 mole % UF, showed that UF, separated in significant omounts from the 12,5, 25, and 31% UF, compositions, with the maximum separation occurting at 25%. Trace amounts of UF, were found in several of the other solid phases, Because of the UF, separation, chemical analysis of the filtrate and residue did not give a relioble indica- tion of the liquidus ond solidus temperatures for most of the 50% NaF join. Mixtures with 75 mole % NaF The thermal data for the 75% NuaF compositions in the NaF-ZrF «UF , system showed evidence of more or less complete solid solution formation between Na,UF, ond Na 3Z¢F .. Since Na UF melts mcongruenfly, it was expected that the sohd solutions would break down near the uranium end of the series. However, prelimincry data indicate ]OL. M. Bratcher and C. J, Barton, ANP Quar. Prog. Rep, Dec, 10, 1952, ORNL.-1439, Frg 10.2, p. 108. 55 ANP QUARTERLY PROGRESS REPORT that o comparatively small amount of Zrf, (1.5 mole %) stabilizes the Na,UF_ structure, Filtrations with mixtures containing 8%, 12%, and ]62/3 mole % UF, (equivalent to 25, 50, and 75 mole % Na,UF,) gave evidence of complete solid solution formation within this range, The x-ray diffraction pattern for all six samples from these filtrations showed only the NusZrF7 ond Na,UF, lines with the maximums shifted to varying extents. Petrographic examination, which is capable of detecting trace aomounts of crystalline phases not shown by x-ray diffraction, showed the presence in some samples of other phases, such das NGQZrFé and NaZrFs, that contained dissolved uranium and, in one case, a small amount of UO,,. An unidentified yellow-brown phase, which ap- to be characteristic of NaF-ZrF -UF, mixtures containing more than 75 mole % NalF, was reported to be present in both the filtrate and the residue from the filiration of the 8]/3 and 127/2 mole % UF , mixtures. Since this phase did not appear in hydrofluorinated mixtures, it seems probable that it is an oxygen-containing complex, The dato shown in Table 5.1 indicate that, within the range of compositions and temperatures covered in these experiments, the system NaF-ZrF -UF, is a true binary system, since the phase separation resulted in mixtures of very nearly the same NaF content as the starting material. The liquidus temperctures determined by the filiration method agreed wel! with the thermal data shown in Fig. 5.1 (cf., subsection on ‘‘Thermal Analysis of Fluoride Systems''). Since cooling curves for these compositions failed to indicate solidus temperatures, the data in Table 5.1 give the first indication spread between liquidus and solidus lines for this system. Further filtrations with compositions approaching that of pears of the temperature the pure compounds are planned for the near future in an attempt to determine whether limits of solid solution exist in this system, FUNDAMENTAL CHEMISTRY Spectrophotometry of Supercooled Fysed Salts H. A. Friedman Materials Chemistry Division Measurements of the absarption spectra for UF andUF ; in quenched fluoride melts with a Beckman DU spectrophotometer and the procedure described by Friedman and Hill'! have been continued. More efficient quenching was obtained by flattening the sealed samples in a hydraulic press to give larger surface areas and thinner melts. With glasses from the NaF-ZrF ,-UF, system, only two absorption spectrum patterns have been found. One of these superficially resembles that of crystalline UF ; the other more closely resembles crystalline Na,UF .. The Na,UF type of pattern was found only in alocalized region near the Na,UF . composition. Glasses of the ZrF -UF system were prepared with compositions in the range from 30 to 70 mole % UF, in 10 mole % intervals. Qutside this region, no glasses were obtained; within this region, Beer's law was closely followed by the UF , type of pattern. Samples along the No ,UF .+Na,ZrF join that contained less than 60% Na,UF . were not studied because of quench growth; the Na,UF , type of pattern wos found in samples that contained from 60 to 100 mole % NGZUFG. Nine mixtures were prepared in which the ZrF, content was held at 41 mole %, but the NaF-to-UF , ratios were adjusted to cover the 114, A. Friedman ond D. G. Hill, ANP Quar, Prog. Rep. Sept. 10, 1953, ORNL-1609, p. 62. TABLE 5.1. LIQUIDUS-SOLIDUS EQUILIBRIA OBTAINED BY FILTRATION INITIAL FILTRATION FRACTION N ANALYSIS OF FRACTIONS (mole %) COMPOSITION N OF CHARGE ) . (mole %) TEMP(E‘Z‘;‘TU“E IN EILTRATE Filirate Residus - o - T NayUF, Na, ZeF, (wt %) NaF ZrF 4 UF, NaF ZE, UF, 75 25 695 63 74.9 5.1 20.0 74.9 8.4 16.7 50 50 742 29 75.0 10.5 14.5 75.1 13.8 11.1 25 75 786 60 74.9 17.4 7.8 75.6 20.4 3.8 56 region from 6 to 59 mole % UF,; all nine mixtures formed glosses and exhibited the UF, type of absorption in conformity with Beer's law. The NaUF .-NaZrF . system is currently being investis gated, : Glosses were obtained with compositions of between 16.67 and 50 mole % UF, in the 2.'1'!’74-UI:3 binary system. At higher concentrations, crystal- lization of UF; could not be avoided. No shift in pattern occurred in the glass region to indicate o change in the nature of the UF, species in the liquid melt ot concentrations below 50 mole % PERIOD ENDING DECEMBER 10, 1953 EMF Measurements in Fused Salts L. E. Topol Materials Chemistry Division The techniques described previously have. been used for additional electrolyses of various solu- tions of metal chlorides in potassium chloride. Containers of morganite, grophite anodes, nd platinum or nickel cathodes are used, and the electrolyses are carried out af a temperature of 850°C, Potassium chloride solutions containing 1 wt % of NiCl, (theoretical E% = 0.82 volt) gave decomposition potentials at 1.20 to 1.26 volts, UF,. The color of the glasses was rust-olive.. A while solutions containing 10 wt % NiCl, yielded typical spectrum is shown in Fig. 5.4 values in the range 1.0 to 1.06 volts. Solutions L .400 [ OWG. 22254 0.300 e e e 2 0.200 |- OPTiCAL DENSITY PER mg OF UF, Q.00 500 500 - 700 800 200 1000 CWAVELENGTH (mu) Fig. 5.4. Absorption Spectrum of the ZrF ~UF , System Containing 33.3 mole % UF .. 57 ANP QUARTERLY PROGRESS REPORT containing 1 to 2 and 10 wt % of FeCl, (theoretical £% = 1.11 volts) yielded decomposition potentials at 1.50 to 1.55 and 1.20 volts, respectively, A mixture containing 2 wt % of NiCI2 and 2 wt % of FeCl, in KCI yielded two distinct brecks in the I-£ curve at 1.18 and 1.56 volts, as shown in Fig. 5.5, UNCLASSIFIED DwG. 22255 0.50 [ 0.40 e - e B— W 0.30 Los : - e e 1 (amp) 0.20 | ~ - e i © INCREASING POTENTIALS A DECREASING POTENTIALS NICKEL CATHODE GRAPHITE ANODE 2.0 30 4.0 E (volts) Fig. 5.5. Electrolysis at 850°C of Mixture Con- taining 2 wt % of NiCi, ond 2 wt % of FeCl, in KCI, in KCi in similar experiments have yielded decomposition potentials at about 1.15 to 1,30 and 1.50 to 1.75 volts, but reproducibility of the results has been rather poor. Electrolyses of KCl with chromium anodes have shown erratic results that are tentatively ascribed Attempts to electrolyze Crf to oxidation of the chromium, even though attempts Rather erratic changes of slope at about 1.2 and 1.9 volts are observed, are made to maintain an inert atmosphere. When a thin plate of nickel was used to protect zirconium electrodes from oxidation, reproducible emf's at 1.06 to 1.12 volts were observed in electrolyses of KCI, Since nickel is not dissolved in KCI ot potentials below 1.8 wvolts, the emf’s 58 obtained are considered to be characteristic of the zirconium. Based on the value for zirconium and the decomposition potential for KCi (3.1 volts), the potential for the reaction ZrCl“m-m‘%h Zr + 2Cl2 appears to be 2.0 volts, When a hydrogen atmos- phere is used over the zirconium electrode system, the decomposition potential rises to 1.12 to 1,18 volts. Whether this may be due to formation of hydrides of zirconium is not certain, Physical Chemistry of Fused Salts E. R, Yan Artsdalen Chemistry Division insight into the fundamental physical chemistry of fused salt systems may be realized from the physical properties of these systems. Accordingly, the depression of the freezing point of sodivm nitrate by a wide variety of solutes was determined, and measurements were made of the density and electrical conductivity of the LiCI-KClI system and the heat capacity of Cdl,. The molten NaCl-KCI system is also being investigated in the hope that it will yield not only practical data, but will also contribute to the understanding of short- Some range order and the nature of electrical conductivity of fused salt systems. This work is reviewed brietly below ond is reported in detail in the Chemistry Division semiannual report. ' 2 Detailed, highly reproducible measurements of the density and the electrical conductivity of the fused salt system LiCl-KCl were completed for the entire composition field over wide temperature ranges. An exact linewr relation between molar velume and composition at any specified tempercture indicates that no complex formation occurs in this molten system. On a molar basis, lithium chloride is a considerably better elecirical conducter than po- It was therefore significant to discover that substitution of lithium chloride for some of the potassium chloride actually lowered the equivalent conductance of the solution below that of pure potassium chloride. This phenomenon was interpreted to be indicative of considerable tassium chloride. short-range order persisting in these high-tempera- ture fluids. A further tentative explanation of this ]2A., S. Dworkin, D. J. Sasmor, and £. R. Yan Artsdalen, Chem, Div. Semionn. Prog. Rep, June 20, 1953, ORNL- 1587, p. 19-25, behavior was based on the consideration of the relative sizes of the ions. Substitution of small Li* ions for some of the much larger K¥ ions will shrink the short-range chloride semilattice of the fluid ond make transport by K* appreciably more difficult. As more and more Li% is substituted for K*, the conductivity will pass through a minimum and again increase until, finally, pure LiCl is present. ' The depression of the freezing point of sodium nitrate was investigated with a wide variety of solutes. The results significantly substantiate the concept that fused salts are essentially totally ionized and behave as nearly ideal electrolytes, even at relatively high concentrations of the solute. Thus, calcium, strontium, barium, and lead chlorides, for example, are completely dissociated, as is LaCl;. FeCl; seems to form two particles only, apparently Fe(:l; and Ci™. Chromate, di- chromate, and bromate ions are stable in liquid sodium nitrate, but the metaperiodate ion, 107, decomposes to yield oxygen and the iodate ion. The heat capacity of high-purity crystalline cadmium iodide has been measured from 15°K to room temperature. It remains necessary to de- in the range well below 15°K because of the observed high heat capacity at this tempera- ture. The heat capacity shows approximately @ l.4-power temperature dependence at the lowest measured range and, therefore, does not yet fit either the usual Debye T3 law or the TZ taw ob- served here for boron nitride. At 298.16°K, the following thermodynamic values have been found for cadmium iodide: ¢, =19,12cal/mole, 5°=38.5,, (F° ~ H3)/T =23.19, (H® — HS)/T = 1531 The $° value has an uncertainty of about 0.5 because of the+high heat capacity value at 15°K and 'its attendant inherent difficulty of extrapolation to 0°K. This uncertainty is, of course, reflected in the two derived quantities, fermine ¢ PRODUCTION OF PURlFI‘ED FLUORIDES F. F. Blankenship G. J. Nessle Materials Chemistry Division L.aboratory-Scale Production of Molten Fluorides F. F. Blankenship F. P, Boody C. M. Blood . M. Watson Materials Chemistry Division Twenty-eight batches fotaii’ng 55 kg of molten fluorides and including one batch containing 15 g PERIOD ENDING DECEMBER 10, 1953 of enriched uranium were prepared and.distributed to various laboratories for testing, When the ex- perimental production facilities have been expanded (cf., subsection on “‘Experimental Production Facilities’), production of materials in this lab- oratory will be terminated, except for the prepara- tion of small enriched-uranium-bearing specimens for radiation stability tests. Purification of Small Fluoride Samples for Phese Studies F. F. Blankenship C. M. Blood F. P. Boody Materials Chemistry Division Studies of phase equilibria in the NaF-ZrF4 and l’\lc!F-Zrl‘""l—UF4 systems by the quenching technique have been hompered by contamination of the materials with traces of water which caused hy- drolysis of the ftetravalent halides during the heating period ond consequent contamination of the quenched melt with oxide or oxyfluoride phases. Materials for this and similar phase equilibrium studies that are essentially free from water and oxygenated compounds are now routinely prepared by amedification of the hydrofivorination-hydrogena- tion pretreatment method used for larger specimens. In this process a number (up to 28) of smali {(~25 g) samples are placed in nickel or platinum crucibles on a nickel rack inside a 4-in.-diameter nickel reactor for processing. A typical processing cycle of 2 hr of treatment with HF at 800°C fol- lowed by stripping with H, at 600°C for 18 hr has been shown to produce material satisfactory for these purposes. To date, a total of six batches comprising 67 samples has been prepared., It appears, however, that handling of the specimens in a vacuum dry box may be required if freedom from oxygenated phases is to be maintained, Production of Enriched Moterial for In-Pile Loop Experiment G. J. Nessle Materials Chemistry Division Plans for preparation of the enriched fuel for o miniature in-pile loop facility and for charging of the material into it have been prepared in accordance with the necessary technical and accountability considerations. |t is planned that this loop will be charged with a uranium-free mixture for prelimi- nary testing before the enriched concentrate is added for final testing. 59 ANP QUARTERLY PROGRESS REPORT The processing equipment for this production operation has been fobricated, When production of the ARE enriched concentrate has been com- pleted, the new process equipment will be installed in Building 9212 in the area now occupied by the ARE facilities. The enriched fluorides for the in-pile loop will be transferred to the X-10 site where the loop will be filled, Experimental Production Facilities J. P. Blakely C. R. Croft Materials Chemistry Division D, E. McCarty R. Reid R, G. Wiley ANP Division A total of 329.2 kg of mixed fluorides was processed ond dispensed to various requestors during this quarter. Included in this total were 14 batches of approximately 2.5 kg each, 3 batches of approximately 10 kg each, and 10 batches of approximately 27 kg each. Although the quontity of fluorides processed decreased, the variety of demand increased sharply. This tendency is becoming more pronounced, and it appears that production of large quantities of a few materials will be of considerably less im- portance; versatility and adaptability of equipment will be primary requirements, Consequently, an installation is being planned for producing o variety of fluoride compositions with a minimum loss of time and equipment. Reduction of Na UF , by Hydrogen . F. Blankenship F. P. Boody Materials Chemistry Division The Na,UF, which had previously been hydro- fluorinated and stripped with hydrogen so that the structural content should have been reduced to less than 30 ppm was bubbled with hydrogen gas in the original nickel! reactor for metal considerable periods of time at 692, 785, and 882°C. It was found that the rate of reduction of the uranium according to the equation ] — UF, + 4H, = UF; + nF increased with temperature, as shown in Table 5.2. When the data given in Table 5.2 are plotted on a semilogarithmic scale as reaction rate vs. (1/T) x 104 the straight line shown in Fig. 5.6 is obtained, From the slope of this curve, the acti- vation energy of the reduction can be calculated to be 33,700 + 2600 cal/mole. Treatment of Molten NuZa‘FS with Strong Reducing Agents . . Blankenship C. M. Biood G. M. Watson Materials Chemistry Division It was previously reported!® that zirconium bars suspended in molten NaZrF ; contained ot 800°C in graphite-lined reactors and stirred by bubbling argon undergo weight losses far in excess of the amount required to reduce structural metal ions. This observation was further confirmed by addi- tional experiments thot involved the treatment of NaZrF. in grophite-lined nickel ot 800°C with zirconium chips; the unrecovered zirconium metal amounted to ot least 20 g per kilogram of fused salt, However, it has since been observed that little if any loss of weight occurred when zirconium metal chips were exposed in a parallel experiment without the graphite liner. The difference might be due to the formation of zirconium carbide, but the available data are insufficient to show it. The possibility that a portion of the zirconium enters the melt as the trifluoride or as disscolved zirconium metal has not been ruled out, IsC, M. Bilood et al., ANP Quar, Prog. Rep. Sept. 10, 1953, ORNL-1609, p. 69. TABLE 5.2 RATE OF REDUCTION OF Na,UF . WITH H, AS A FUNCTION OF TEMPERATURE REACTION RATE AS MOLES OF HF PER LITER OF H, BUBBLED TEMPERATURE 1 % 104 oc °K TCK) 692 965 10.36 785 1058 9,45 882 1155 8.65 8.8 x 107° 3.7 x 107% 1.58 x 10~¢ 60 ~E— —3 DWG. 22256 REACTION RATE as moles of HF per titer of H2 8.0 a5 9.0 g5 0.0 105 1.0 Fig. 5.6. Reaction Rate vs. Reciprocal Temper- ature According to the Equation UF , + 3’2 H, = UF, + HF. The effect of uranium metal on NaZrF, was studied by adding 5.5 wt % of uranium chips to the material in o graphite-lined reactor. Unlike the previous experiments, the mixture was treated with hydrogen for 2 hr at 800°C, and the material was subsequently filtered ot 700°C. The filtrate started freezing at about 540°C and contained five phases typically found in fuels which are reduced to the extent that orange-red UF 4:2ZrF ; and purple UF, oppear. The uranium content of the filtrate was 5.00 wt %, and opparently about 90% of the vranium reacted. [f it is considered that the dis- solved uranium is present as UF; and that all other reduced phases are insoluble, a reducing power of 210 meq per kilogram of melt should have resulted; if all reduced phases remained in solution, PERIOD ENDING DECEMBER 10, 1953 the theoretical reducing power would be 840 meg/kgq. Chemica! analysis by H, evolution showed 550 meq of reducing power per kilogram of filtrate, which indicated the presence of soluble reduced phases in addition to the UF,. Less than 20 ppm of nickel and chromium ond 85 ppm of iron were found, The unfiltered residue in the reactor con- tained black opaque material which liberated gas readily when treated with dilute acids. X-ray analysis of the residue revealed the presence of ZrH, UF,, and uranium metal, along with occluded melt. Thus it appears that the uranium did not react quite completely and that some zirconium tetrafluoride was reduced to the metal by the urco- nium. Since ZrH, is decomposed to the extent of about 25% ot 800°C under 1 atm of Hz,]4 most of the zirconium appears as ZrH. A further check on the possibility of retaining reducing species such as ZrS ZrF3, ZrC, ZrH, and ZrH, in NaZrF. melts wos attempted by treating NaZrF . with NaH (2 moles per kilogram of melt). The mixture was sparged with hydrogen in a graphite liner for 24 hr ot 800°C and then filtered. The reactor heel contained a dark-black residue which was tentatively identified os ZrH mixed with the melt. Chemical analysis of the white filtrate showed 106.5 meq of reducing power per kilogram or about 5% of that expected if all re- duced species were soluble. Efforts to learn more about the nature of this apparent reducing power will be continued, | | Reduction of NiF, by Hydrogen F. F. Blankenship C. M. Blood G. M. Watson Materials Chemistry Division Considerable attention has been given to the kinetics of the reduction of Ni*% in NaZrF . by H, at 800°C because of the importance of this reduc- tion in fluoride purification procedures. The use of graphite linersin the nickel reactors for handling molten salts is a great convenience in loading and unloading and in avoiding excessive contamination with NiF, during the HF treatment. Accordingly, for most of the kinetic studies, graphite-iined reactors have been used. Early suspicions that graphite might reduce Ni*" were allayed by the large, positive, standard free energy estimated for 14M. N. A. Hall, Trans. Foraday Soc., 41, 306 (1945). 61 ANP QUARTERLY PROGRESS REPORT the reduction with carbon, 2NiF,(solid) + C(solid) > 2Ni%(solid) + CF (gas) AF{ g0k = 118 keal , and by the failure to find o noticecble difference in short experiments without graphite liners. |In carrying out more detailed measurements over a period of days, the graphite liners have been found to play an important but incompletely understood role in the reduction of Nit", As a result, the apparent complexity of the reduction kinetics is now explainable from o qualitative standpoint, but 6 new set of experiments without the graphite liner must supplant the many experiments made with the use of the grophite liner before guanti- tative data can be assumed, In a series of triols which provide evidence for the effect of grophite, a weighed portion of nickel fluoride was mixed with previously purified NaZrf which was then pretreated at 800°C in a graghite- lined reactor by bubbling inert gas through it at an approximate rate of 400 cm?/min ond, finally, treated with hydrogen at the same approximate bubble rate. The varying concentrations of hydrogen fluoride in the effluent gases were continuously determined by direct titration of standard potassium hydroxide samples to the phenolphtalein end point as the gases bubbled through. In Table 5.3 are listed the initial concentro- tions of nickel fluoride used, the volume of inert gas passed during pretreatment, the volume of hydrogen passed, and the percentage of hydrogen fluoride recovered during the hydrogen treatment. From the results given in Table 5.3, it appears that the recovery of hydrogen fluoride decreases as the length of pretreatment increases; that is, it is apparent that a fraction of the nickel flueride added is reduced by some agent other than hydrogen and thot this fraction increases with increasing tength of pretreatment. Since it appeared that the graphite liner was in some fashion effectively reducing the NiF,, one experiment that essentially duplicated the lost test shown in Table 53 was conducted without the graphite liner in the nickel vessel. Copious evolution of hydrogen fluoride occuired. The re- covery measured amounted to 72.6% at the end of passage of 60 liters of hydrogen. Kinetically, the reaction appears complex in the presence of graphite, as evidenced by the voriable over-all order with respect to the apparent nickel fluoride concentro- tions. Preparation of Various Fluorides 8. J. Sturm Materials Chemistry Division The methods of preparation and some of the physical properties of numercus complex fluorides that resulted from the interaction of alkali with structural metal ftluorides have been previously re- ported. 19 During this quarter, the simple structural metal fluorides have been used increasingly in corrosion studies, and a major partion of the avail- able time has been devoted to preparation of these The compositions of these fluorides were determined by chemical analyses, and some supplementary x-ray and petrographic determinations were made for further characterization. Several batches of FeF, and FeF, have been prepared by hydroflucrination of the anhydrous chlorides. Emphasis was placed on the preparation of relatively small batches of these two fluorides materials. ORML-1439, p. 123; G. J. Messle and H. W. Savsge, ANP Quar. Prog. Rep. Morch 10, 1953, ORNL-1515, p. 115, 8. J. Sturm and L. G. Overholser, ANP Quoi. Prog. Rep. June 10, 1953, ORNL-1556, p. 48, TABLE 5.3. RESULTS OF TESTS OF THE REDUCTION OF Nin BY HYDROGEN INITIAL VOLUME OF GAS PASSED HF RECOVERED CONCENTRATICN ot P DURING HYDROGEN OF Nif, elium .reweofment Hydrogen Pretreatment TREATMENT (meg/kg) (liters) (liters) (%) 15.7 44 69 82.5 29.2 132 119 81.6 228.0 1285 60 9.0 62 of the highest possible purity. Several pounds of anhydrous Nif, were prepared by hydrofluorination of the hydrated N!F Small quantities of anhydrous NiCl, were S|m|lurly prepared from hydrated N|C| tor specnul uses. Previous work had indicated that the precipitate resulting from the addition of NH HF, toan aqueous solution of CdCl, could not be quantitatively con- vertedto CdF , by hydrofluorination. Further studies have shown that the solids resulting from the addi- tion of NHHF, to aqueous solutions of either CdSO, or Cd(C H,0 )2 give the same x-ray pattern as thclt found for the material prepared from CdCl,; this pattern differs from the accepted pattern for CdF, and apparently indicates complex or double salt formation. |t was observed that the interaction of HF with CdCO, in an aquecus medium gives a precipitate of CdF, that has the x-ray properties accepted for CdF ... In an effort to prepare some of the flyorides of molybdenum, the powdered metal wos brominated to produce a material consisting. mainly of MoBr,. A first attempt to hydrofluorinate this material was incomplete; components of the mixture have not yet been identified. PURIFICATION AND PROPERTIES OF HYDROXIDES Purification of Hydroxides E. E. Ketchen L. G. Overholser Materials Chemistry Division The work with hydroxides during this quarter has been confined to the purification of NaOH and Sr(OH)., according to the methods previously des- <::r|bed:Zl Earlier studies indicated that Na,CO, could be separated from NaOH by filtering the 50 wt % aqueous solution of NaOH through a fine sintered-glass filter instead of decanting off the NoOH solution, as had been done. Further work has shown that the filtration method produces o material with a slightly lower Na,CO, content and without a significant increase m s:hcu content. 16D, R. Cuneo ef al., ANP Quar, Prog. Rep. Mareh 10, 1953, ORNL-1227, p. 104; E. E. Ketchen and L. G. Overholser, ANP Quor. Prog. Rep. June 10, 1952, ORNL-1294, p. 89. : PERIOD ENDING DECEMBER 10, 1953 Seven batches of NaOH containing approximately 400 g each have been prepared by filtration and de- hydration during this quarter. The material obtained contained from 0.02 t0 0.10 wt % of Na,CO, andless than 0.1 wt % of H,0. The Sr(OH)2 purified during this quarter con- tains considerably less SrCO, thon the material previously. prepared; the improvement resulted from the replacement of a faulty filter., For the 12 batches produced, the SarCO3 contents range from 0.02 to 0.06 wt %, and the total olkalinities, as determined by titration, range from 99.8 to 100.0%. Reaction of Sodium Hydroxide with Carbonaceous Matter E. E. Ketchen L. G. Overholser Materials Chemistry Division A few experiments were run to check observations made elsewhere which suggested that the No,CO content increased when NaOH was heated at 700°C ina sealed nickel container. If the increase occurs, it may be due to the presence in the hydroxide of carbonaceous matter capable of oxidationto NaéCOs. Samples of commercial NaOH {about 1.5 wt % Na,CO;) showed no significant heating ot 700°C for 24 hours. analytical change aofter However, the results were not sufficiently precise to detect very small chonges in this range of carbonate concentration. When samples of purified NaOH (one purified by filtration and one by de- contation) for which occurate COJ ~analyses could be obtained were heated in nickel for 24 hr at 700°C, slight increases in carbonate content were observed., Both purified NaOH samples contained 0.02 wt % Na,CO, prior to heating. After heating, the NaOH purified by decantation contained 0.03 wt % Na,CO, and that purified by filtration con- tained 0.06 wt %. The increose observed could be due to oxidation of the carbon present in the nickel or the NaOH. Additional tubes were filled with NaOH to which had been added 0.2 wt % graphite. After these tubes were heated at 700°C for 24 hr, preliminary onalyses indicated a marked increase in carbonate content, This suggests that carbon is oxidized by NaOH ot 700°C; if this is the case, NaOH could not be maintained carbonate free at high temperatures in metals which contain appreci- able percentages of carbon, 63 ANP QUARTERLY PROGRESS REPORT 6. CORROSION RESEARCH W, D. Manly Metallurgy Division W. R, Grimes F. Kertesz Materials Chemistry Division H. W. Savage ANP Division The stotic and the seesaw corrosion testing facilities have been used to study the corrosion of Inconel, cermets, and some spzacial alloys in the fluoride fuel NaF-ZrF -UF, (50-46-4 mole %), The effects of improper degreasing of metals and of contaminants in the fluorides were studied. Several tests of bearing materials and hard-facing alloys have been completed in which the molten fluorides were used as the corroding agent. The materials tested are of considerable interest because of their prospective uses as high-temperature bearings and valve facings. Various cermets have also been tested in the fluorides to determine their fitness for these same applications. Static tests in graphite capsules of Inconel specimens exposed to fluoride fuel to which FeF3, FeF,, or Nil'”'2 additions were made gave anomalous results that were presumably due to the reduction of the added metal flucrides by the graphite, Static tests of itwo dissimilor metals indicated that one metal is usvally prefer- entially attacked, Fluoride corrosion tests were made in [nconel thermal convection loops to supplement and extend the data obtained in the above-mentionsd and previous static tests. Although the thermal con- vection loops normally operate with a hot-leg temperature of 1500°F, the temperatures were varied to determine the effects of both time and temperature. Additional tests have substantiated the earlier conclusion that attack by NaF-ZrF4-UF (50-46-4 mole %)} on Inconel is independent 01 temperature, at least in the range of 1500 to 1650°F, and dependent on exposure time. The purity of the fluoride mixture was shown to be important when the attack depth was increased by a factor of 3 in a loop operated with a low-purity batch of fuel. Reductions in attack, in comparison with that found in the standard loop, that is, a loop operated for 500 hr at 1500°F, were achieved by (1) precirculating the fluorides, (2) pretreating the fluorides with chromium metal flakes, and (3) holding the fluorides with in contact Inconel 64 turnings before circulation in the loop. While pretreatment of the fluorides with metallic chromium was beneficial, a chromium plate on the pipe wall was not, Increasing the UF, content in the flucride mixture from 4 to 4.5 mole % caused an incregse in mass transfer in the loops and possibly a slight increase in corrosion. The maximum depth of attack by the purified ARE fuel solvent, NaZrFS, was comparable to that of the purified fuel mixture (with 4 mole % UF ), although the intensity of the NGZst attack was much less and the same mass transfer wos noted. Additional studies were performed in an attempt to establish the chemical equilibria of the postulated corrosion reaction which results in the selective leaching of chromium from Inconel. Preliminary values have been obtained for the equilibrium con- stants involved, but the data are notyet sufficiently reliable, Nickel and type 430 stainless steel thermal convection loops were also operated with the fluoride NcJF«ZrF"'-UF4 {50-46-4 mole %). There was no measurable mass transfer in the nicke! loop, and the hot-leg surface had a polished finish. The hot leg of the type 430 stainless steel loop showed a small amount of smooth, even removal, with no subsurface voids or intergranular attack; metal crystals were deposited in the cold leg. A number of tests, both static and dynamic, were performed with the liquid metals lithium, sodium, and lead. Tests have been completed on beryllium oxide and a number of bearing materials and cermets. The investigation of the mass transfer and corrosion of metals and alloys in quartz thermal convection loops containing liquid lead have continved. The metals tested during this quarter included o molybdenum-nickel (25% Me--75% Ni) alloy, a chromium-iron-silicon (14% Cr—-84% Fe-2% Si) alloy, and types 304 and 347 stainless steel specimens which were oxidized prior to contact with the lead, The most significant result from this study was the marked improvement in the resistance to mass transfer of these metais be- cause of the oxide film. Some static tests on the corrosion and mass transfer characteristics of various aolloys in contact with lithium were com- pleted. ‘ Static tests on a number of solid fuel elements were performed in sodium and sodium hydroxide. While there was no evidence of attack in any of the sodivm tests, all the fuel elements were severely attacked in the hydroxide. In connection with a study of hydroxide containment, the equilibrium pressures of hydrogen over a number of hydroxide- metal systems are being determined, FLUORIDE CORROSION IN STATIC AND SEESAW TESTS E. E. Hoffman L. R. Trotter J. E. Pope D. C. Vreeland Metallurgy Division H. J. Buftram R. E. Meadows No V. Smifh Materials Chemistry Division Both the static capsule and the seesaw tests provide relatively cheap and simple means of investigating the many parameters which affect fiuoride corrosion. The seesaw tests in which the capsule contdining the fluoride is rocked in a furnace were operated at 4 cps with the hot end of the capsule at 815°C and the cold end at 700°C. The static tests in which the fluoride mixture is sealed in a capsule were vsually conducted for 100 hr at 816°C. In order to permit testing of material which could not be readily fabricated into capsules or to permit variation of the surface-to- volume ratio of fluoride to metal, a number of tests were performed in which the fluoride and metal specimens were contained in graphite capsules. All capsules are filled in a helium dry box, and therefore oxidation or hydrolysis of the contents of the capsules is improbable, Inconel Corrosion by Fluorides with Metal Fluoride Additives In general, corrosion of Inconel specimens by several fluoride preparations in graphite capsules was similar to that obtained when inconel copsules were used in static tests. Some carburization of the specimens was obtained, but no other special effects were observed. | : PERIOD ENDING DECEMBER 10, 1953 However, when fluorides to which had been added considerable quantities of FeFS, Fer', or NiF , were tested in graphite capsules containing inconel specimens, rather surprising results were obtained, In each case, tests at 800°C showed extensive deposition of iron or nickel, usually in films on the graphite capsule and on the speci- men. When Fef_ was the additive (in amounts up to 5 wt %), the corrosion of the Inconei specimen, as observed by metallographic examination and by analysis of the final melt for chromium, was much iess than expected. In similar tests with FeF_, corrosion of the Inconel was virtually undetectdb?e by either technique. |n this case, the precipitated film of metal appeared to be protective. When NiF_ was the added material, however, no protection of the specimen by the film resulted, in these experiments, exiremely heavy subsurface void formation to a depth of 9 mils wes observed. There is evidence that corrosion by fluoride mix- tures with NiF added is more severe at 600 than at 800°C. These experiments suggest that graphite or some unknown contaminant in the graphite is responsible for reduction of the added metal fluoride. While reduction of FeF_, for example, by graphite should have an extremelzy iow equilibrium constant, it is possible that the volatile reaction product (CF“) escapes from the system and permits the reaction to proceed, It is also possible that there is sufficient oxygen or water adsorbed in or on the graphite so that reduction of oxide by carbon (o reaction with an appreciable equilibrium constant at these temperatures) is responsible for these effects. Further study of this and similar reactions is in progress., ' ' Corrosion of Yarious Metal Combinntions A series of experiments in which various metals, either alone or in combination, were tested in molten fluorides in graphite capsules have vielded the following results. (1) Nickel alone shows no attack. (2} Chromium-plated lncone! shows heavy subsurface void formation ‘at all temperatures above 600°C; the voids were 3 to 4 mils deep at 700°C and above. (3) Chromium metal yielded high values for chromium content of the melt; it is possible, however, that slight surface oxidation of the chromium pellets used was responsible for this dissolved material, (4} When Incornel and chromium specimens were exposed fogether, the chromium 65 ANP QUARTERLY PROGRESS REPORT test piece was nearly completely dissolved, while the Incone! specimen, which was covered with a thin metal layer, was almost completely unattacked. (5) However, when Inconel and nickel were exposed together, the nickel appecred unaffected, while the Incone! was badly attacked; at 700°C, for example, the heavy subsurface void formation reached a depth of 3 to 4 mils. Corrosion of Cermets Several cermets which were considered for bearings and hard-facing applications were tested for resistance to fluoride corrosion. Three of these ceramicemetal materials were subjected to static tests. The materials consisted of heavy metal carbides bonded by nickel and were cbtaired from the Firth Sterling Company. Table 6.1 presents results obtained when the cermets were exposed to two fluoride mixtures in graphite capsules. The results indicate fair resistance to the two fluoride mixtures under the conditions of the $est, and there wos no apparent change in the appecrance of the specimens after testing. Another set of specimens was subjected to see- saw tests in Incone!l capsules for 100 hr, with the specimens restricted to the hot zone of the Inconel tube. The hot-zone temperature was 816°C, and the cold-zone temperature was 730°C. Results of these tests are presented in Table 6,2. Some of these specimens contained chromium as a binding TABLE 6.1. RESISTANCE OF CERMETS TO TWO FLUORIDE MIXTURES Static tests at 800°C for 100 hr in graphite containers CERMET FLUORIDE COMPOSITION WEIGHT CHANGE (wt %) (mole %) (%) TiC-CryCo-Ni (42.9-7.1-50.0)* NaF-ZrF, (50-50) .12 NaF-ZrF ~UF, (53-43-4) +1.27 TiC-Mo,C-Ni (72-5-23) NaF-Z:F ; (50-50) +0.047 NaF-ZrF ,-UF, (53-43-4) ~0.005 CryC-Ni (89-11) NaF-ZrF (50-50) 40.152 NaF-ZeF UF, (53-43-4) *0.153 * Firth Sterling 27. TABLE 6.2. RESULTS OF SEESAW TESTS OF VARIOUS CERMET MATERIALS TESTED IN N(fl"’-ZrF“--UF4 (50-46-4 mole %) MATERIAL COMPOSITION (wt %) METALLOGRAPHIC NOTES Metamic L. T-1 (not heat treated) Metamic LT-1 (heat treated) Firth Sterling 27 Kenametal 151A CrAl, 0, (77-23) Cr-A1,0, (7723) TiC2 with Mi-Cr binder* TiC2 with 20% Ni Specimen completely penetrated Specimen complztely penstrated Attack zone 2 to 5 mils deep, 3 mils in most ploces Slightly affected zone 1 tc 2 mils deep *C#., Table 6.1. 66 agent, and, os expected, these materials were attacked by the fluorides. Metamic L.T-1, in hoth the heat-treated and unheat-treated conditions, was attacked severely. While the attack on the Firth Sterling 27 and on the Kenometa! 151A in the seesaw tesfs was only modérofe, the increased severity of attock in the seesaw tests compared with that in the static tests is evidenced by the test data on Firth Sterling 27. Inconel with Dil and Trichloroethylene Additives Several thermal convection loops had more than average depths of attack which were thought to be due to improper degreasing of the loops prior to filling. Therefore, a static test was run in which six drops of oil and six drops of trichlorcethylene were added to an Inconel capsule containing an inconel specimen and the solution was allowed to evaporate in an attempt to duplicate improper de- greasing. This capsule was then loaded with the fluoride mixture NoF would ensure adequate weld buildup and could be removed by finish machining. 90 A series of specimens of the design shown in Fig. 7.7 were welded with variations in arc current, arc time, arc distance, and welding sequence, Since precise control of these variables was re- quired, cone-arc equipment was used throughout the study. The welds were examined metallograph- ically for penetration and flaws. E xamination of test specimens indicated that the most favorable weld geometry could be achieved by two separate inert-arc welding operations, The projection was melted info a ‘‘rivet’’ by an initial choice of arc current, orc time, and arc distance. The arc distance was then reduced, and on addi. tional application of current was used to form the the finished weld, the filler metal being provided by the rivet. The rivet formed during the first operation by using the optimum welding conditions is shown in Fig. 7.8, which is a section through the weld joint corresponding to section A-A in Fig. 7.7. The plug weld formed in the second operation is shown in Fig. 7.9. As may be noted, the penetration of the type 304 stainless steel-clad sheet was complete and reasonably uniform, which indicates that an apparently sound weld joint was obtained. [t thus appears that inert-arc plug welding may be a feas- ible method of fabrication of these assemblies. UNCLASSIFIED DWG 22216 DRILLED HOLE // z" - - Lo e g S e Aoy~ 0.067=in~dia i | H 7 " 1] 7 11 “\ N 0.062 in, ~~-HEADER SHEET s SECTION B-8 E T i 1 0.084 1 A : N ‘*\.‘\\ N [ TYPE 304 STAINLESS STEEL-CLAD SHEET /% : SO ONN, r. D.026 inqad-t=-— SECTION A~A Fig. 7.7. Assembly of Type 304 Stainless Steel Components Prior to Inert-Arc Plug Welding, B UNCLASSIFIED [ eHOTO Y 10336 Fig. 7.8. Section of ‘‘Rivet'' Formed by Initial Application of a 30-amp (DCSP) Arc Current for an Arc Time of 2 sec at an Arc Length of 0,140 in. Etchant: aqua regia. 21X ' PERIOD ENDING DECEMBER 10, 1953 However, ‘ the application of these techniques to the fabrication of a multiweld unit should be investigated further, ' MECHANICAL PROPERTIES OF INCONEL R. B. Oliver D. A. Douglas K. W. Reber J. M. Woods Metallurgy Division Stress-Rupture of Inconel in Fluoride Fuel Stress-rupture tests of coarse- and fine-grained Inconel in argon and in fluoride fuel NaF-ZrF -UF at temperatures of 1300, 1500, and 1650°F are be- ing made, Data for the higher stress ranges are complete, and the tests being made currently are expected to run from three to eight months, Test. ing in the stress range from 500 to 1500 psi will commence in about six months when the six new units designed for low loads are completed.. The results obtained to date for fine-grained Inconel are summarized in Table 7.2. Tube-Burst Tests of Triaxially Stressed Tubes In the past, data obtained in the tube-burst tesis of triaxially stressed Inconel tubes have shown considerable scatter and interpretation has been difficult. However, data obtained in recent tests have shown reasonable reproducibility, and it is believed that most of the operating difficulties have been overcome. The present procedure is to UHCLASSIFIED | PHMOTO Y 10335 Fig. 7.9. Section of Completed Weld Formed by Application of a 30-amp Arc Current for 4 sec at an Arc Length of 0.060 in. to the ‘““Rivet’’ Shown in Fig. 7.8. Etchant: oqua regia. 21X N ANP QUARTERLY PROGRESS REPORT TABLE 7.2, RESULTS OF STRESS-RUPTURE TESTS OF FINE-GRAINED INCONEL IN NaF-ZrF4-UF4 RUPTURE TIME RUPTURE TIME IN TEMPERATURE STRESS IN ARGON L UORIDES (°F) (psi) (hr) (hr) 1300 10,000 20600 1500 10,000 50 30 1300 5,000 7000 1500 5,000 540 260 1650 5,000 25 25 1500 2,500 4000 1200 1650 2,500 450 175 test the tube with the fluoride fuel NaFoZrF4-UF4 inside and purified argon outside. The stress range from 500 to 3500 psi is currently being in- vestigated ond tests have been completed in the vpper half of the range. In comparing the data with creep data at the same stresses in the same fuel, it is found that the tube-burst time is about the some as the time for 2% elongation in the tensile creep tests, T he tube specimens looded by internal pressures give an axial-to-tangential stress ratio of ]/2, which results in minimum ductility; thus the results probably represent the minimum life ex- pectancy, other factors being equal. In the near future, these tube specimens will be tested with the fuel inside the tube and sodium, instead of argon, outside. Environmenta! Effects on Creep of Inconel Creep tests of Inconel specimens in nitrogen and in sodium environments at 1500°F have recently been completed. Thus it is now possible to com- pare the effects of six types of environments that Inconel structures might encounter. The results of these creep tests are summarized in Table 7.3. Table 7.3 shows that hydrogen is the most detri- mental environment and that acir is the most bene- ficial, as far os rupture life is concerned, It is interesting to note that, with the exception of the tests in air and sodium, all the specimens reached about the same total elongation regardless of the rupture time, The increased ductility in air and in sodium is probably the result of decarburization. Chemical analyses are being mode to test this hypothesis., The results presented in Table 7.3 are for Inconel specimens that were annealed at 92 TABLE 7.3. RESULTS OF CREEP TESTS OF INCONEL. ANNEALED AT 1650°F AND TESTED UNDER A STRESS OF 3500 psi ot 1500°F IN VARIOUS ENVIRONMENTS LIFE FINAL ENVIRONMENT | RUPTURE | ELONGATION () (%) \ Hydrogen 446 13.0 NaF~ZrF4-UF4 550 12.0 Sodium 1333 35.0 Argon 1467 12.0 Nitrogen 1770 11.6 Air 2567 50.0 1650°F hefore being tested. Specimens annealed at 2050°F and then stressed in hydrogen or Naf- ZrF4-UF4 have shown more resistance to the environmental effect. Tests of this type of speci- men in the other environments are now being made. HIGH-CONDUCTIVITY METALS FOR RADIATOR FINS E. S. Bomar J. H. Coobs H. Inouye, Metallurgy Division The investigation of the fobrication of high- thermal-conductivity fins was confined during this period to a study of diffusion barriers for use be- tween Inconel and copper, the fabrication of clad- copper fin material in sufficient quantities for the assembly of experimental radiators, and the investigation of special materials, Diffusion Barriers Barrier material for use between Inconel and copper was selected on the bosis of low solubility of the element or alloy in copper or Inconel. In addition, it was desired to fabricate thin composite clads to determine their rolling characteristics. The barrier materials used were tungsten, molybde- num, titanium, zirconium, tantalum, vanadium, types 310 and 446 stainless steel, and iron plus silver, ‘ Several composites were hot rolled approxi- mately 50% ot 1800°F and then finished cold to 0.008 in, total thickness. Bonding was not achieved between molybdenum and Inconel or between va- nadium and Inconel, and the following composites could not be successfully rolled: (1) Inconel-Ag- Fe-Cu-Fe-Ag-Inconel and (2) Inconel-W-Cu-W- inconel, The first of these composites was un- successful because of improper design, and the ‘second showed severe stretcher marks, Oxidotion and diffusion fests of the other composites are being made. . ' Some of the earlier tests of Inconel-clad copper showed the absence of diffusion voids, and there- fore a few tests are now being made to determine the effect of surface preparation on void formation and diffusion. The composites for this study have been fabricated; they consist of a layer of oxide on Inconel formed by heating in wet hydrogen and in air at 1300°F, Clad Copper As reported previously,? types 310 and 446 stain- less steel are suitable cladding material for copper for use at 1500°F, while Inconel cladding couses severe diffusion. In order to establish which of the claddings is most suitable on copper and also to determine the effect of diffusion on thermal con- ductivity, opproximately 15-ft2 samples of copper clod with Inconel, type 310 stainless steel, and type 446 stainless steel have been or are being made. The fin material is to be strips 4 in, wide and 0.010 in, total thickness; the copper core will be 0.006 in. thick and the cladding material will be 0.002 in. thick. This material is to be used for the fabrication of experimental heat exchangers. Electroplated Copper Plates of chromium plus iron and of iron plus nickel plus chromium to give compositions corre- sponding to types 446 and 310 stainless steel, PERIOD ENDING DECEMBER 10, 1953 respectively, did not protect copper from oxidizing at 1500°F. A composite made entirely by electro- forming showed the same results. The inability of these plates to protect copper ot 1500°F is ascribed to the lack of a metallurgical bond between the copper and the plate and to the brittleness of the deposits. Although the ductility can be restored by heating to 1600°F, the domaging cracks are present in the as-received or annealed material, Solid Phase Bonding An calternate investigation for obtaining uniform thin claddings is being made., A bond between the copper and the cladding can be obtained by seam welding maoterial of the final desired thickness. The Glen [, Martin Company proposed this method and has supplied samples up to 4 in. long which appear satisfactory. Joint efforts have, to date, produced claddings 0.002 and 0,001 in. thick. The thermal properties of the aiuminum bronzes at low temperatures indicate that the conductivity of alloys containing about 6% aluminum is greater than that of the stainless steels or Inconel by a factor of about 6 at 1500°F. To verify this, meas- urements are now being made on & and 8% aluminum bronzes. These alloys are oxidation resistant, and if the expected thermal properties are correct, the need to fabricate composite bodies will be eliminated because these alloys may be suitable for use as radiator fins. This is especially im- portant because difficulties have been experienced in obtaining clad material from commercial suppliers, FABRICATION OF SPECIAL MATERIALS E. S, Bomar J. H. Coobs H. Inouye, Metallurgy Division Extrusion of Inconel-Type Alloys Extruded Inconel tubing was redrawn from 1 in. in outside diameter with a 0.125-in. wall to 0.500 in. in outside diameter with o 0.035-in. wall by the Superior Tube Company., These moterials were drawn satisfactorily at reductions as high as 36% per pass with annealing after each stage at 1900°F, The properties of the tubing are listed in Table 7.4, The extrusion of near-lnconel alioys from seven other vacuum meits was dttempted with varying degrees of success., Temperatures from 2150 to 3E. S. Bomar et al.,, ANP Quar, Prog. Rep. Sept, 10, 1953, ORNL.-1609, p. 96. ' 93 ANP QUARTERLY PROGRESS REPORT TABLE 7.4. PROPERTIES OF ANNEALED HIGH-PURITY INCONEL TUBING 0.505 in. IN OUTSIDE DIAMETER WITH A 0.035-in. WALL GRAIN SIZE TENSILE STRENGTH YIELD STRENGTH ELONGATION MELT NO.* . . (mm) (psi) (psi) (%) DPL.7 0.090/0.130 £1,000 35,000 52 DPI-8 0.075 83,400 36,000 50 DP9 0.090/0.110 83,500 40,000 48 DPIi-10 0.090/0.130 80,400 40,800 46 *Analyses of melts 8, 9, and 10 were given in the Metallurgy Division Progress Report for the Period Ending April 10, 1953, ORNL-1551, Table 22, p. 54. 235)°F were employed to produce the tubing (or rod), ond, in some instances, extrusion rotios as low as 10:1 were required, All the extruded tubing has been sent to the Superior Tube Company for further reduction., A molybdenum-nickel alloy (20% Mo-~80% Ni, nominal composition) did not extrude at a reduction of 13:1 at 2350°F. An attempt will be made to extrude this alloy at a lower reduction ratio, Rolling of Chromium-Cobalt Alloy A chromium-cobalt alloy ingot 6 in. long and 1 in. in diameter (45% Cr—55% Co, nominal composi- tion) was cost, and attempts to roll it at 2350°F showed that this alloy was hot-short, Additions of 0.5% manganese and 0.5% aluminum did not im- prove the hot-rolling properties. A solid rod of this alloy which was precision cast was readily machin~ able, but it cracked severely, The present plans call for precision casting the alloy tubing. The hardness of the cast alloy is Rockwell C-50, Rolling of Cebalt A ]l/z-in.-dics, 6-in.-long cobalt ingot containing 0.5% moanganese and 0.5% aluminum was success- fully rolled to Y -in.-dia rod at 2300°F, Previously, it was found that cobalt without the manganese and aluminum additions was hot-short, The alloy contgining aluminum and manganese has anas-cast hordness of Rockwell A-49, Rolling of lron-Chremiuym-Nicke! Aloy A 'Il/-in.»dm, 6-in.-long ingot of iron-chromium- nickel alloy (45% Fe-40% Cr-15% Ni) was rclled o yz-in.-dia rod at 2350°F, This alloy was found to be hot-short at 1850°F, even though it contained 94 1% of a nickel-manganese-titanium master alloy, Previously, attempts to hot roll this alloy at 1850°F without the alloy addition were unsuccessful be- couse of hot-shortness. The as-cast alloy has @ hordness of Rockwell A-62, Cold-Ralled Columbiem Columbium sheet 0,20 in. thick obtained from the Fansteel Corporation was annealed to a grain size of ASTM 7 at 1100°C in a vacuum. The sheet was cold rolled to 0,025 in., which represented a reduction of 87.5% i thickness. The room-temper- ature mechanical properties were determined on specimens vacuum annealed for ]/2 hr at various temperaiures and are summarized in Table 7.5, The microstructure of columbium under these con- ditions shows partial recrystallization at 1000°C, and, at 1100°C, recrystallization is complete. Vacuums of the order of 107> mm Hg were obtained in the annealing furmnace. The analysis of the sheet shows 0.75 + 0,3% tantclum, and traces of copper, iron, and titonium, Analysis of the cxide content of the sheet is not complete, but it is expected to be greater than 0.3%. TUBULAR FUEL ELEMENTS J. H. Coobs, Metallurgy Division Drawing of tubular fuel elements has been con- tinved, Plans called for the reduction of six tubes on each of two schedules in steps of 15% and 2% reduction per pass, respectively, All tubes were to be reduced from 0.750 in. in diometer with o 0.042-in. wall to a final size of 0.250 in. in dicmeter with a 0.015-in. wall, or o total reduction of 87%. Plug drawing was used in preference to drawing on a mandrel, as explained previously.? However, be- cause of failure of the tubes at intermediate steps, the drawing schedules could not be carried to com- pletion. Reduction of the tubes was continued as long as was considered practical, that is, until each tube had failed within the core region or near the starting end in such a way as to make further drawing difficult or impossible. The final sizes of the tubes processed on the 15% schedule are given in Table 7.6, along with core composition and total reduction. : | Because the results with the 20% schedule were discouraging, only three tubes were processed; dll 41bid., p. 99. PERIOD ENDING DECEMBER 10, 1953 tubes failed on the second pass. Total reduction of these tubes was 36%. In general, the tubes reduced by both schedules developed excellent finishes on both inner and outer diameters, with no evidence of rippling or folding in the core region. In most cases, failure occurred near the frailing end of the core. The failures were atiributed mainly to inability to keep the plugs positioned in the die properly, with the result that the reductions did not conform strictly to the schedule. The reductions were somewhat higher than scheduled on severcl passes and caused galling in the dies and foilure of the tubes because of severe tensile stresses, Samples have been cut from the nine tubes drawn and are now being examined metallographicalily. TABLE 7.5. MECHANICAL PROPERTIES OF COLUMBIUM COLD ROLLED TO A REDUCTION OF 87.5% ANNEALING YIELD POINT AT | TENSILE | ELONGATION | MODULUS OF SPECIMEN HARDNESS NO. TEMPERATURE | = 0.1% OFFSET | STRENGTH | IN1in. | ELASTICITY ’ (°C) (psi) (psi) (%) (< 10~¢ psi) 1 As rolled 145 76,500 4.0 2 300 154 69,500 80,700 2.6 17.66 3 600 _ 147 62,300 73,400 8.0 16.60 4 900 136 57,300 63,100 8.3 17.30 5 | 1000 | 100 37,500 50,800 12,6 17.88 6 1100 85 23,400 42,200 26.6 19.30 7 ; 1200 86 24,000 45,000 18.6 17.20 TABLE 7.6. FINAL SIZES OF TUBES PROCESSED ON THE 15% SCHEDULE TUBE fi NO. OF | FINAL SIZE TOTAL REDUCTION COMPOSITION (BY VOLUME NO. _ ( ) PASSES (in.) (%) 1 80% type 302 stainless steel-20% U02 5 ’ 0.562 56 2 70% Ni-30% UO, 9 0.396 77 3 80% Ni-20% UO, 9 0.396 77 4 70% type 302 stainless steei-30% UO, 10 0.355 80 5 80% Fe—20% UO, 3 0.625 ' 39 6 70% Fe~30% UO, 5 0.562 56 95 ANP QUARTERLY PROGRESS REPORT CERAMIC RESEARCH L. M. Doney J. A, Griffin J. R, Johnson, Metallurgy Division Glass«Type Pump Seals Work continued on the development of a suitable, viscous, high-temperature pump seal, Ten test rings, 2) in. ID, 3% in, OD, % in, thick, of a beryllium fluoride glass® were prepared by melting the glass batch in a platinum crucible and casting the glass in grophite dies (Fig. 7.10). The rings were cooled slowly in o blanket of glass wool to minimize cracking, The finished rings were sent to the Experimental Engineering Group for testing. Ceramic Container for Fuel Further work is being done on the fabrication of a ceramic container for yse in the investigation of the electrical resistivity of the fluoride fuels (cf., sec, 8, ‘‘Heat Transfer and Physical Properties'’). Three specimens, 5]/2 x 1 x 1 in, were cut from hot-pressed beryllium oxide blocks (ARE blocks), A hole z’] in. in diameter was drilled 4‘/2 in. deep along the fong axis of the specimen and then tapered to 0.158 in. in diameter for the remaining length of the specimen. These specimens were tested and found to be resistant to attack by the fuel; how- ever, they were too short to satisfy the test condi- tions. Work has commenced on the fabrication of a SL. M. Doney, J. A. Griffin, and J. R. Johnson, ANP Quar, Prog, Rep. Sept, 10, 1953, ORNL-1609, p. 98. ~< I PHOTO 20810 Fig. 7.10. Giass-Type Pump Seal. Composition (by weight): BeF,, 50%; KF, 25%; MgF,, 16%; AlF 5, 9%. 96 suitably sized beryllium oxide shape by isostati- cally pressing and sintering. High-Density Graphite Specimens of high-density graphite (Graph-i-tite) supplied by the Graphite Specialties Corporation were obtained and sent to the Metallurgy Division’s Corrosion Group for compoatibility tests with the fluorides and other molten salts and molten metals. Typical properties of this graphite are given by the manufacturer as: Apparent density, g/cm® 1.85 to 1.92 Tranverse breaking strength, psi 4000 to 4500 Modulus of elasticity, Ik/in.2/in./in. Electrical resistance, ohm-in. 20 x 10° to 25 x 10° 0.00032 to 0.00040 Permeability 1.85-g/cm® density Impermeable to water at 40 psi and room temperature for 10 min minimum 1.92-9/cm® density Impermeable to air at 40 psi and room temperature for 10 min minimum COMBUSTION OF SODIUM ALLOYS G. P. Smith M. E, Steidlitz Metallurgy Division Studies of the flammability of jets of sodium alloys in wet and dry air at temperatures of up to 800°C have been continued. Sodium-mercury alloys containing less than 34 mole % sodium ond sodium- bismuth alloys containing less than 40 mole % sodium did not burn. All other clloys tested burned in degrees ranging from slight to violent, The com- bustible alloys tested were: binary alloys contain- ing, in addition to sodium, 90% aluminum, 50 to 60% bismuth, 90% indium, 90% lead, 0.6 to 66% mercury, 90% silver, or 90% zinc; ternary alloys having a 45% sodium—50% bismuth base and containing 5% each of calcium, copper, magnesium, mercury, potassium, or silver; and ternary alloys having a 0% sodium—65% mercury base with 5% each of aluminum, bismuth, calcium, copper, magnesium, potassium, or silver. The humidity of the air was noted to have a very slight effect on the rate of combustion. PERIOD ENDING DECEMBER 10, 1953 8. HEAT TRANSFER AND PHYSICAL PROPERTIES RESEARCH H. F. Poppendiek Reactor Experimental Engineering Division The physical properties of o number of fluorides and other moterials of interest to aircraft reactor technology have been measured. Preliminary density and viscosity measurements have been obtained for the ARE fuel concentrate over the temperoture range of &60 to 1000°C; the viscosity varied from about 17.5 cp at 725°C to 9.9 cp at 975°C. The vapor pressure equations for two NaF-ZrF, mixtures were determined. The thermal conductivity of the solid fluoride mixture NaF-KF- LiF-UF, (10.9-43.5-44.5-1.1 mole %) was deter- mined to be 1.7 Btu/hrft°F, while that of the solid heat transfer salt NaNO «NaNO ,-KNO, - (40-7-53 wt %) was only 0.6 Btu/ilr'ff’oF. The heat capacity of NaF-ZrF -UF, (50-25-25 mole %) was 0.17 cal/g-°C in the solid state and 0.27 cal/g°C in the liquid state over the temperature range of 610 to 930°C, with a heat of fusion of 42 cal/g. The heat copacities of two special samples of type 310 stainless steel and of G-E No. 62 brazing alloy have also been measured. Additional forced-convection heat transfer experiments with NaF-KF-LiF eutectic in a nickel tube were made, and the data obtained were found to be in agreement with data from the former experi- ments. No additional thermal resistance was present and no film deposits on the inner surface of the nickel tube were cobserved, in contrast to the conditions found with this fluoride mixture in Inconel. _ Temperature and velocity measurements were obtained in o glass thermal convection loop over a Grashof modulus range from 2 x 10% to 67 x 104, The experimental velocity data are compared with predicted values which were obtainedby the numeri- col solution of the laminar-flow heat-conduction equation. The measured values diverge radically from the theoretical values, undoubtedly because of the turbulence which was introduced by the radial temperature gradient and which was observed at Reynolds numbers as low as 100. Forced-laminar-flow volume-heat-source velocity and temperature solutions have been derived for the case in which the fluid viscosity is temperature dependent. These solutions are functions of a new dimensionless modulus which is a measure of the importance of the viscosity variation of the system. Experimental apparatus is being assembled for measurements of fluid flow in an annulus ond for studies of transient surface-boiling phenomena. The annulus flow will be determined by photo- graphing dust particles; the essential photographic technique is being developed. The boiling studies will be conducted on a flat metal filament sus- pended in water. HEAT CAPACITY W. D. Powers G. C. Blalock Reactor Experimental Engineering Division The enthalpy and the heat capacity of NaF-ZrF - UF, (50-25-25 mole %) were determined' in the liquid and the solid state with a Bunsen ice calo- rimeter. The data can be represented by the fol- fowing equations: ' H (solid) — H o .(solid) ~18 + 0.177, c, = 0.17 + 0.02, at 280 to 610°C, HT(quuid) - Hfloc(soiid) -39 + 0.27T1, c, = 0.27 +0.02, at 610 to 930°C, where H is the enthalpy in cal/g, T is the tempera- ture °C, ¢, is the heat capacity in cal/g'°C. The it 0°cC -heat of fusion is 42 cal/g at 610°C, The enthalpy and the heat capacity of two special samples of type 310 stainless steel and of brazing compound G-E No. 62 (69% Ni-20% Cr-11% Si) were determined for the General Electric (.:<;'om;:u:m),v:2 1. for type 310 stainless steel (heat 64177) from 238 to 858°C, f{T — HOOC f ~7.2 + 0.143T, 0.143 + 0.007; “p it 'W. D. Powers and G. C. Blalock, Heat Capacity of Fuel Composition No. 33, ORNL CF 53.11-128 {Nov. 23, 1953), 2W. D. Powers and G. C. Blolock, Heat Copocity of Two Samples of 310 Stainless Stee! ond of a Brazing Compound, ORNL CF 53-9-98 (Sept. 18, 1953). 97 ANP QUARTERLY PROGRESS REFPORT 2. for type 310 stainless steel (heat 64270) from 240 to 834°C, f T "HOOC c P ~5.5 + 0.1397, 0.139 + (0.005; 3. for G-E No. 62 brazing alloy from 211 to 840°C, } il -4.3 + 0.142T, 0.142 + 0.012. THERMAL CONDUCTIVITY OF SOLIDIFIED SALTS W. D. Powers R. M. Burnett S. J. Claiborne Reactor Experimental Engineering Division The thermal conductivities of solid flucrides have been measured by a transient cooling methed, A sphere of the material being studied is allowed to until a uniform initial The sphere is then rapidly transferred to a large circulating bath at a lower temperature, A fine-gage thermocouple at the center of the sphere is used to determine the time-temperature curve as the sphere cools. Clas- sical transient temperature solutions in terms of remain in o calorimeter temperature is established. the physical properties of the sphere are available and may be usedto extract the thermal conductivity of the sphere from the experimental time-tempera- ture dota. In this way the thermal conductivity of the solid fluoride mixture NaF-KF-LiF-UF, (10.9- 43.5-44.5-1. 1 mole %) was foundto be 1.7 Btu/hrft? (°F /1), as compared with a value of 2 Biu/hrft? (°F/ft) for the same fluoride mixture in the liguid state. The thermal conductivity of a solid heat transfer salt (NaNO,-NaNO,-KNO,; 40-7-53 wt %) is also being determined by the transient cooling method., Preliminary measurements indicate a value of 0.6 Btu/hr ft2(°F/ft) over the temperature range 80 to 170°F. DEMSITY AMND VISCOSITY OF FLUORIDES S. |. Cohen T. N, Jones Reactor Experimental Engineering Division Preliminary viscosity and density measurements on Na,UF ( were made® on a Brookfield viscometer and displacement apparatus, respectively, The density, as plotted in Fig. 8.1, is given by the equation p = 5598 - 0.00119T, where p is in g/cm3 and $60°C < T < 1000°C. The viscosity varies from about 17.5 cp at 725°C 3& l. Cohen, Preliminary Measuremant of the Density and Viscosity of Composition 43 (N02UF6), ORNL CF 53-10-86 (Oect. 14, 1953). S DWG. 22217 50 ¢ 4.9 - 4.8 4.7 plg7em3)=5598-0.001197, NSITY (g/em3} 2 B \ | i 1 T E60°C < 7T < 1000°C D 650 800 900 1000 TEMPERATURE (°C) Fig. 8.1, Density of Ma UF as o Fuactior of Temperature, 98 to about 9.9 cp at 975°C (Fig. 8.2), and the experi- mental data are represented by the equation p o= 0.92 e2936/T , where yiis in centipoises and T is in °K. pr— 50 DWG. 22218 1 . e dameiaeaaan e e becrmebeee s 2O b - 8 > 40 5 B & _ 3 - @ >y 1 i B - 1 (cp)=0.92 £2938/7 ook 2 . T e e e e E ! | f ! ‘ 300 560 000 1100 1200 1300 1400 TEMPERATURE {°K} Fig. 8.2. Viscosity of No,UF , as a Function of Temperature, The following solid density measurements at room temperature were made on NaF-ZrF, mixtures in support of phase studies of these materials: DENSITY (g/cm?) Batch B-62 (53-47 mole %) 4.00 Batch R-132 (52-48 mole %) 4,103 Batch R-133 4.124 A method for predicting densities of fluoride mixtures have been developed and will be described in a forthcoming report. Plots of liquid densities at any temperature vs, the calculated room-tempera- ture densities have been developed that correlate all the experimental data available to within 2%. Similar plots have been devised for viscosity datq, but the correlation is not dependable. VAPOR PRESSURES OF FLUORIDES R. E. Traber, Jr. R. E. Moore C. J. Barton Materials Chemistry Division Additional measurements of vapor pressures by the method of Rodebush and Dixon? were madg. PERIOD ENDING DECEMBER 10, 1953 Chemical analysis has shown that the material used for the previously reported® measurement of the vapor pressure of NoF-ZrF,, nominally 50-50 mole %, actually contained 47 mole % Z¢F,. In- stead of being the pure compound NaZrF,, it was probably a solid solution of NaZrF, and Na ZrF The vapor pressure equation for fl'us 47 mole % ZrF (~53 mole % NaF mixture is 7213 log P = — + 7.635, where P is in mm Hg and T is in °K. The vapor pressure equation for the mixture NaF-ZrF , (50-50 mole %) is 6827.7 log P = ~ + 7.503 over the temperature range 812 to 942°C. ELECTRICAL CONDUCTIVITY OF FLUORIDES N. D, Greene Reactor Experimental Engineering Division Preliminary electrical conductivity measurements of NaF-ZrF , (50-50 mole %) and NoF-ZrF, (57-43 mole %) hqve been obtained for the tempera'rure range 1000 to 1800°F, For these measurements, a thick-walled cylindrical tube of beryllium oxide is mounted vertically in an open Inconel crucible contdaining the molten fluoride. An electrode is inserted through the tube, and the electrical path is defined from the electrode just under the fiquid surface to the Inconel at the bottom of the crucible. However, the beryllia tube was not sufficiently long to prevent noniscthermal regions in the cell, and therefore accurate measurement of the salt temperatures were difficult. A longer tube has been ordered. An attempt was also made to measure the con- ductivity of NaF-KF-LiF by using the apparatus described above, but this salt penetrated ond eroded the tube so badly that the data obtained were deemed unrelioble. A new cell in which only platinum is in direct contact with the salt haos been designed and is now being constructed. This cell will be used to measure the conducti» vities of salts which are not compatible with 4W. H. Rodebush and A. L. Dixon, Phys. Rev. 28, 851 {1925). R. E. Moore and R. E. Traber, ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL-1609, p. 106. 99 ANP QUARTERLY PROGRESS REPORT beryllia. Since slight contamination of the melts occurred as a result of the constituent materials of the crucible going into solution, a more noble material than either Inconel or stainless steel is indicated. Accordingly, a platinum crucible has been designed and constructed for this purpose. FORCED-CONVECTION HEAT TRANSFER WITH NoF-KF-LiF EUTECTIC H. W. Hoffman J. Lones Reactor Experimental Engineering Division Additional data have been obtained for NaF-KF- LiF eutectic (11.5-42.0-46.5mole %)} flowing through a heated nicke! tube. The experimental results are presented in Fig. 8.3 in terms of the Colburr j-function. The previously reported results for NaF-KF-LiF in both nickel and Inconel tubes are also shown. In review, NaF-KF-Lif heat transfer data were first obtained by using a nickel tube, The results (shown by the inverted triangles in Fig. 8.3) indi- cated that the anticipated heat transfer correlation was followed. The experiment was then repeated with the eutectic flowing through an Inconel tube, and the data {circles in Fig. 8.3) were foundtolie approximately 50% below the expected correlation. The discrepancy was believed to be due to the formation of a film of corrosion products at the fluid-metal interface. The heat transfer results and the existence of the film were both definitely verified for NaF-KF-LiF in Inconel tubes in a subsequent experiment (squares in Fig. 8.3). The film was analyzed and found to be K,CrF [ plus a small amount of Li CrF .. The data reported ot this time were obtained to establish the validity of the original results for NaF-KF-LiF in nickel tubes. The results (normal triangles in Fig. 8.3) are essenticlly in agreement with those for nicke! tube No. 1;visual observation of the tube showed no film. The experiment was terminated because of fatigue failure of the nicke! tube at the high operating temperatures used. Experimental apporatus is being constructed for studying fluoride films formed on Inconel as a function of time and temperoture, Because of the unavailability of pure K CrF, it has not yet been possible to measure the thermal conductivity of this material. Several other fluoride mixtures are being studied to determine the thermal conductivity of such films because they may form on Inconel tubes. UNCILLASSIFIED DWG. 22219 0.010 e ey . - e - - 0 B ¥ NICKEL TUBE NO I | 3 | A NICKEL TUBE NO.2 T T " B 1}* 1 - " ® INCONEL TUBE NO.1 R —1- -— ® INCONEL TUBE NG 2 g | = | \ ‘ _ b L ~ bl S ,,,,,,] Lgl i 0.005 — ‘ ‘ _ I lL { [ ‘ | : I‘ 5 v i !G """" I ¥ 1 /.«—-‘t“‘*'::;;_l____\i """"" o ' ; | - ; 4-— 5 ‘ /,7" A b‘“"‘*--._._\\ ; ‘ | ! w i v A ‘ | ' /=0023 Re™9? | l '\‘ — —_— e ——— ——— —;}J EE -‘ ‘ ,,,,,,,, 5 i, s % 1 // ‘ ! ‘ ‘ { i % v'/ | ! i ! ‘ E)J /v | | [ ! t Co S 0002 L 7+7a oL Lol e foo b | e o o e - | 5 | | ‘ el - 0 | | | | || Lo ‘ | | \ | 1 ‘ | o ! 0.001 ‘ L L I L 103 2 5 10% 5 103 REYNOLDS MODULUS Fig. 8.3. The j-Modulus vs. Reynolds Modulus for NaF-KF-LiF {i1.5-42,0-46.5 mole %}. 100 FLOW IN THERMAL-CONVYECTION LOOPS D. C. Hamilton F. E. Lynch L. D. Palmer Reactor Experimental Engineering Division Thermal-convection loops have been used for observing the nature of flow of ordinary fluids and to determine the accuracy and range of usefulness of an analytical method for predicting the mean velocity, or Reynolds modulus, from wail tempera- ture measurements, A discussion of the velocity measurements ond flow observations was given in a previous report.® Additional velocity and wall temperature data have been taken, fluid temperature traverses have been made, the analytical analyses have been completed, and ¢ report of the results is being prepared.” ' Measurements of temperature and velocity were taken for five specific conditions over the range of Grashof modulus from 2 x 104 to 67 x 104, The corresponding variation of Reynolds modulus was from 50 to 270. The Grashof and Reynolds moduli are, respectively, a’?’figfiw(max) Gr, = ———————— and where d is the tube diameter, v is the kinetic viscosity of the liquid, 3 is the thermal coefficient of expansion of the liquid, g is the acceleration of gravity, ! - 6 (max) is the maximum variation in wall tempera- ture, is the mean fluid velocity, and The flow ond heat transfer observed in these experiments were characteristic of free convection rather than forced convection, and thus the signifi- cant parameter which dictates the character of the flowis the Grashof modulus rather than the Reynolds modulus. | At Reynolds moduli of up to about 100, the flow was essentially laminar. Above this value, the Grashof modulus was superceritical, ond the flow was laminar over part of the circuit ond turbulent 8p. €. Hamilton, F. E. Lynch, and L. D. Palmer, ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p. B9. p. Hamilton, F, E, Lynch, end L. D. Palmer, The Nature of the Flow of Ordinary Fluids in « Thermal Convection Harp, ORNL-1624 (to be published), PERICD ENDING DECEMBER 10, 1953 in other parts. As the Grashof modulus {and thys Reynolds modulus) was increased, the length of the circuit in laminar flow decreased. In addition, when the Groshof modulus was high, the radial temperature gradient was sufficient fo cause vigor- ous secondary flow cells in the laminar part of the circuit, The method for predicting the Reynolds modulus from wall temperature measurements employed the numerical solution of the laminar flow equation for heat conduction. The problem waos solved on the ORNL high-speed computer, the ORACLE. In Fig. 8.4, the predicted and measured values of Reynolds moduli are compared. For the first two sets of datq, for which the flow was essentially l[aminar, the correlation is no worse than 30%; for Reynolds moduli grzater than 100, the flow was observed to be partly turbulent, and o laminar flow solution is obviously not applicable. The predic- tion is satisfactory for the laminar flow regime, but a similar prediction for the turbulent flow regime cannot be made because the turbulent diffusivity is not known for this type of turbulent flow. UNCLASSIFIED DWG 22220 1000 i l J , 800 R - l ,,,,,,,,,,, ...._A_ ....... ................. ‘ POSTULATED VELOf‘ITY e 4o 600 | ...l DISTRIBUTION: 4 elbo ................. o -U_m_ =4 | 400 |m—ed A @--efi-u—m s e // 30O o e fi—r—- seesquesqenpeepasgeen e Dt 3 o ] ° A @ 200 w#ul - // /. o = 1 / 2 | 5 9 | +30 % // O \ ] £ 00 e YR [ '1' /-( a ‘ ',_ o o Ll 1 . . 'z J ________________ 20 40 80 BO GO 200 MEASURED REYNOLDS MODULUS 10 20 300 40@ Fig. 8.4. Comparison of Predicted and Measured Reynolds Moduli for a Therma! Convection Loop. 107 ANP QUARTERLY PROGRESS REPORT The method used in the previcus work for measur- ing velocity is not applicable to higher flows Therefore a velocity measuring device is being developed which will be applicable at higher velocities and which will not offer resistance to the flow. FLUID FLOW IN AN ANMULULIS J. O, Bradfute Reactor Experimental Engineering Division The experiment to be used for measuring fluid velocity profiles in annulus systems was described previously.? In essence, the method consists of photographing dust particles which are entrained in air flowing in an annulus. Most of the recent effort on this research was devoted to phatographic experimentation to develop a procedure for photo- graphing the grid system and the dust particles with the light available from a single flash of a Strobolume. This has been accomplished. The problem stemmed from the fact that colimating a beam of light into a sharply defined region in space, G necessary inefficient, is extremely A pair of parallel slits, each about 2 mm wide and opproximately 25 cm apart, with the requirement, Strobolume placed immediately behind the last one, produced what seemed to be the optimum combina- tion of intensity and definition in the usable por- tion of the beam. Although about 700 lumen-seconds per flash of the light energy was dissipated by the light source, the developed image of the grid sys- tem obtained by using Kodak Tri-X panchromatic film ond o stop opening of f:4.7 was so under- exposed that even the doubly accented centimeter lines were barely visible upon careful examination. Several film sensitization and latensification materials, as well as high-speed developers, were tried with varying degrees of success and failure, A :1.8 lens was obtained, and, by employing this lens and following a complex developing process, negatives were produced which have sufficient contrast for satisfactory printing. Chalk-dust particles were also photographed by applying this procedure, TRANSIENT SURFACE-BOIJLING STUDIES M. W. Rosenthal Reactor Experimental Engineering Division In the event o water-cooled reactor suddenly become supercritical, the time required for produc- 102 tion of a certain volume of steam would be of im- portance, Of primary interest would be the effect of generation of heat ot rates increasing exponen- tially with time in vertical plates (such as the fuel plates of an MTR-iype reactor) submerged in water that was initially at below the boiling point. To obtain information on this phenomenon, an experimental study of heat transfer to water under transient conditions has been undertoken. The basic experimental system is a flat metal filament positioned vertically in o tank of water with heat generated electrically in the filament. The resecrch program consists of twe phases: a study involving exponential generation of heat, which will be started when the necessary electrical equipment has beern developed, and a preliminary general study of boiling from a vertical filament, which will begin immediately. Arrangements have been made for the Instrumenta- tion and Controls Division to develop and construct equipment to regulate the flow of current in the filament and to produce an exponential heating curve. Thisequipment will also measure and record instantaneous values of the variables of interest. The filament and the adjacent water will be photo- graphed by using a high-speed camerg imotion-picture syiichronized with the heat generation system. CIRCULATIMG-FUEL HEAT TRANSFER H. F. Poppendiek .. D. Palmer G. M. Winn Reactor Experimental Engineering Division Some preliminary experimental temperature meas- urements have been obtained in a forced-flow volume-heot-source system under !aminar flow conditions. The volume-hect-source is generated within the flowingelectrolyte by passing an electric current through it. In general, the experimental results are in ugreement with the previcusly de. veloped laminar flow theecry., At low flow rates (Reynolds number of 600 or less), the effect of the free convection which is superimposed on the forced convection was observed. The variable ronges within which theoretical and experimental forced-flow volume-heat-source behavior have been SJ, 0. Brodfute and J. I. Lang, ANP Quor, Preg. Rep. March 10, 1953, ORNL.-1515, p. 160. compared to date are the following: 600 < Re < 14,000, 4.6 < Pr < 8.7, to~— 1§ 3 x 10—+ <6 x 10-2 , m 2 Yo k The next set of experiments is to be made with the test section in a vertical rather than a horizontal position; also, concentrated sulfuric acid is to. be used as the electrolyte so that a new Prandtl modulus range con be studied. ' Currently, the forced-turbulent-flow temperature solution for the case of fluid flow between parallel plates and annular spaces with volume heat sources in the fluids is being evaluated, This solution will be useful in estimating fluid temperature distributions in reactor systems such as the reflector-moderated reactor. PERIOD ENDING DECEMBER 10, 1953 Forced-laminar-flow volume-heat-source velocity and temperature solutions have been derived for the case in which the fluid viscosity is temperature dependent. A simple, algebraic viscosity-tem- perature expression which closely opproximates experimental viscosity data was substituted into the hydrodynamic equation. A power series with unknown coefficients was substituted into the heat transfer equation, and two integrations were carried out. The resulting temperature solution was then substituted into the hydrodynamic equation, and an integration: was carried out. The coefficients of the resulting power series velocity solution were determined by relating them to the initially proposed velocity power series and the integral equation describing the mean fluid velocity. The resulting velocity and temperature solutions are functions of a new dimensionless modulus, as well as radial position. = This dimensioniess modulus is a measure of the importance of the viscosity variation in the flow system. 9. RADIATION DAMAGE J. B, Trice, Solid State Division A. J. Miller, ANP Division Inconel capsules containing the fluoride fuel NoF-ZsF ,-UF , were irradiated in the LITR and in the MTR. Examination of the capsule walls showed a tendency toward intergranulor corrosion that does not occur in unirradiated static tests. The dato on depth of penetration showed considerable scatter, and therefore no correlation with power production rate and exposure time could be made. Design and construction ‘continued on the inepile circulating- fuel loops and the MTR creep test equipment. IRRADIATION OF FUEL CAPSULES G. W. Keilholtz C. C. Webster J. G. Morgan M. T. Robinson H. E. Robertson W. R. Willis W. E. Browning Solid State Division Inconel capsules containing various compositions ‘of the fluoride fuel system, NaF-ZrF 4-UF,, have been irradiated in both the Lowwlntensny Test Reactor (LITR) and in the Moterials Testing Re- actor (MTR). These tests have included irradiation times of from 53 to 800 hr with power production from the 230 watts/cm3 with a 4 mole % UF , fuel in the LITR to the 8000 watts/cm® with a 12 mole % UF , fuel in the MTR. In general, Inconel sur- faces show more tendency toward intergranular corrosion when in contact with irradiated fuel than they do in out-ofepile tests. However, the data so far available indicate no definite trend in corrosion penefration either as a function of rate of power production in the fuel or of irradiation time. Six Inconel specimens containing the fluoride mixture NoF-ZrF -UF (50-46-4 mole %) which were irradiated in the LITR for 53, 140, 270, 565, 685, and 810 hr each are shown in Fig. 9,1. In each case, the power dissipation in the fluoride was 230 watts/em3. Out-of-pile control capsules for this series of tests are shown in Fig. 9.2. Inconel capsules which were irradiated in the MTR with power production in the fluoride mixture of 2500, 3900, and 8000 watts/cm? are shown in Figs. 9.3, 9.4, and 9.5, respectively. The Naf- 103 ANP QUARTERLY PROGRESS REPORT Fig. 9.1. Static Corrosion of Inconel by NuF-ZrF4-UF4 (50-46-4 mole %) After Exposure at 1500°F in the LITR to 230 watts/cm> for Various Irradiation Times. ZrF -UF , system wos used in each of these tests, although the composition was varied and the UF content was 4, 6.5, oand 12 mole %, respectively, in the three tests. The sample shown in Fig. 9.5, which was contacted for 575 hr in the MTR with fuel producing power at a rate of the order 8000 watts/em3, has only 3 or 4 mils of corrosion pene- tration and is similar to the sample shown in Fig. 9.1, which was exposed at only 230 watts/cm3. In a few samples in which the Inconel had a very large grain size, occasional penetration to a depth of 12 mils has been observed. This phenomenon appears to be fairly independent of power density or time of irradiation, and, as with the other effects observed, there is no certainty as to how much of the effect is due to radiation, 104 In confirmation of the analytical werk reported previously, petrographic examinations of irradiated fuels showed no decompasition or segregotion of the fuel as a result of irradiation. CREEP UNDER IRRADIATION W. W. Davis J. C. Wilson N. E. Hinkle J. C. Zukas Solid State Division No new creep tests were mode during this quarter, The recent increase in power of the LITR causes excessive microformer operating temperotures, and the increasing incidence of microformer foilures (which occurred even before the power was in- creased) prevents relicble strain measurements. | uncLAssiFieD | | PHOTO 11985 Fig. 9.2. Static Cotrosion of Inconel Exposed to NoF-ZrF -UFJ {(50-46-4 mole %) ot 1500°F for Various Times. Accordingly, no more creep tests will be operated in the LITR until a svitable strain transducer is found. Both the Bourdon-tube extensometer and a bimetal expansion extensometer now being de- veloped show promise of becoming reliable meas- uring instruments for use in the LITR ond in the MTR. | Several models of an in-pile stress-corrosion apparatus are being built to determine the best system for conducting stress-corrosion tests. The discovery of an effect of straining on the corresion of Inconel! and the recent observation of inter- granular corrosion in static, inspile fuel irradiations 'R, B. Oliver et al., Met. Div. Semiann. Prog. Rep. April 10, 1953, ORNL.1551, Figs. 43 and 44, p. 52, 53. PERIOD ENDING DECEMBER 10, 1953 | UNCL ASSIEIED E PHOTO 11961 Fig. 9.3. Static Corrosion of Inconel by NaF- Zrl=4-l.ll"'4 (50-46-4 mole %) After Irradiation in the MTR at 2500 wotts/cm3 for 575 hr ot 1500°F. UNCLASSIFIED PHOTO 11962 OUT-0OF~PILE SPECIMEN IRRADIATED SPECIMEN Fig. 9.4. Static Corrosion of Inconel by NaF- ZrFA-UF - {53.5-40-6.5 mole %) After 419 hr at 1500°F \éith and Without lrradiation in the MTR at 3900 watts /cm3. 105 ANP QUARTERLY PROGRESS REPORT Fig. 9.5. Static Corrosion of Inconel by NaF. Zer'UF4 (63-25-12 mole %) at 1500°F With and Without Irradiation in the MTR at 8000 watts/cm3, have pointed to the need for data on the effects of in-pile stress corrosion. The MTR creep test rig has been welded shut, but considerable difficulty has been experienced with legks in the helium supply system and in the seals where the lead wires exit from the pressurized portion of the experimental plug. The complete 106 instrument assembly has been tested ond found satisfactory. IN-PILE CIRCULATING LOOPS 0. Sisman C. Ellis W. E. Brundage M. T. Morgon F. M. Blacksher A. S, Olson W, W. Parkinson Solid State Division In-pile circulating loops for studying the corrosion of Inconel by fluoride fuels under dynamic con- ditions in a neutron flux are being built for insertion in hale HB-2 of the LITR. It is expected that similar loops con be used in the MTR. A loop consists essentially of U-shaped sections of 0.225-in.-0D tubing, which will be in the neutron flux, thick-walled 1/2-in. tubing thot will connect the irradiated section with a Venturi flowmeter and an air-cooled heat exchanger just inside the reactor shield, and a centrifugal sump pump outside the shield, All these components, except the shaft and seal portion of the pump, will be enclosed in a helium atmosphere. All in-pile components, other than the pump, will be surrounded by a water- cooled jacket. _ The fabrication of components for the first loop is 80% complete and most of the work remaining to be done on this loop consists of welding or brazing together the finished parts. The components for a second loop are 40% complete. The withdrawal shield is being fobricated, but work on the shield for the pump has not started. The design of the instrumentation is 90% complete, and about 85% of the instruments and controls are on hand. PERIOD ENDING DECEMBER 10, 1953 10. ANALYTICAL STUDIES OF REACTOR MATERIALS C. D. Susano, Analytical Chemistry Division J. M. Warde, Metallurgy Division R. Baldock, Stable lsotope Research and Production Division Developmental work was mainly concerned with the investigdtion of methods for the determination of the oxidation states of the constituents and the metallic corrosion products in fluoride fuel mix- tures and fuel solvents. Divalent iron fluoride was ieached from samples of fluoride mixtures with 0.1 M H,50,, and the iron was determined as the o- phenanthroline complex. Divalent chromium fluoride was found to be complexed by 0,17 M (NH )2C204 and could be distinguished from trivalent chromium fluoride, which is readily complexed with a dilute solution of disodium dihydrogen ethylenediamine- tetraacetate (EDTA) at pH 4.0. Studies were continued on the use of ammonium oxalate for the determination of traces of UF ., in the presence of UF . A separation based on solubility difference is possible but not for low concen- trations of UF,. The absorption spectra of so- lutions of UF, in EDTA and HCl were determined. The effectiveness of solutions of ammonium oxalate, oxalic acid, and acetic acid as solvents for removing fluoride fuels, which contain zirconium, from metallic thermal convection loops was in- vestigated. Ammonium oxalate proved the most effective solvent tested; however, it was not con- sidered so satisfactory as the nitric and boric acid mixtures which are currently used. Petrographic examinations were made of irradiated samples of fluoride fuel mixtures. No difference was noted under the petrographic microscope be- tween the irradiated specimens and the comparable control samples. The mass spectrometer method for quantitative analysis of the uranium content of irradiated fluoride fuel mixtures has recently been extended for use as a measure of uranium burnup during irradiation. A procedure has been developed for calculating the percentage of uranium by weight in both the un- irradiated and the irradiated conditions. Hence, the loss in uranium content can be determined to the accuracy that the isotope abundances are known in the originol material. The portion of this total uranium loss that is due to burnup is then determined by mathematical analysis., The activities of the Analytical Service Laboratory have increased markedly during the past quarter. A total of 1,378 samples was received as compared with 878 for the previous quarter, and 1,201 samplies invelving ‘o total of 11,034 determinations were reported. i ANALYTICAL CHEMISTRY OF REACTOR MATERIALS J. C. White Analytical Chemistry Division The developmental work during this quarter con- sisted almost entirely of a search for methods for determining the oxidation states of the constituents and the metallic corrosion products in ARE fuels and fue! solvents. Three phases of the problem are currently being studied: the determination of the oxidation states of iron in fuel solvent and in the fuel; the deterinination of the oxidation states of chromium in similar media; and the determination of the oxidation states of uranium in the fuel. The ultimate aim of these investigations is the determi- nation of each specie in the presence of the other species. The reducing power of specially prepared NaZrF_ with zirconium additive was also de- termined. The effectiveness of certain solvents for disselving proposed fuels from the walls of metal contciners was investigoted. No additional work was conducted on the determination of metallic oxides in reactor fuels because of the urgency of other problems. ' Oxidation States of {ron P. V. Hottman Analytical Chemistry Divisien A number of samples of the fuel mixture NaF- ZrF ~UF | (50-46-4 mole %) and iron (5 wt %) were heated and filtered. Divalent, trivalent, and total iron contents were determined, in addition fo vranium frifluoride. The finely divided samples were leached with 0.1 M H2504 on the steam bath for 1 hr to remove iron. Ferrous iron in the filtrate 107 ANP QUARTERLY PROGRESS REPORT from the leach was determined with o-phenanthro- linz;! iron was determined in the leach filtrate by first reducing the iron present with hydrogquinone and then determining it as the o-phenanthroline complex. In order to test the efficiency of the leaching process, a separate sample was dissolved, and the iron was then determined by the o-phen- anthroline method. The results, with a relative standard error of 10%, show that a single acid leach is sufficient to leach iron quantitatively from the fuel mixture, Further evaluation of the data shows that only o small fraction of the added iron reacts with the fuel and passes in solution through the filter. The oxidation state of this iron is predominantly, if not com- pletely, bivalent. {t should be recognized, however, that if FeF_ and UF, exist as such in the mixture, reduction of trivalent iron to the bivalent state vndoubtedly occurs on contoct with sulfuric acid. The absence of UF_ was indicated by the hydrogen evolution method.2” However, this method is not sufficiently sensitive to determine the probable small concentration of UF, which may be present in these samples. The need for 0 sensitive method for determining UF, in reactor fuels is evident. Oxidation States of Chromium D. L. Manning Analytical Chemistry Division Work has continued on the determinotion of the oxidation states of chromium following its reaction with UF& in fuel solvent NcquFs. The tormula for the reaction is NaZrF === Cr® + UF, . 800°C It has besn shown that chromium trifluoride can be separated from chromium meta! by leaching with an acetate-buffered solution (pH, 4,0) of disodium ditydrogen ethylensdiaminetetraacetate (EDTA). Chromium metal is insoluble in this medium. Several samples were leached with EDTA at 60 to 70°C for 3 hr to determine the extent of reduction of CrF, to metallic chromium during the heating peried. No residues (1) CrF, *+ 3UF, indicotive of metallic chromium were 15, C. White et al., AMP Quor, Prog, Rep. June 10, 1952, ORNL-1294, p. 177. 2D, L. Manning, W. K. Miller, and ®&. Rowan, Js., Methods of Determingtion of Urenivm Trifivoride, ORNL. 1279 (Apr. 25, 1952). 108 obtained. This suggests that chromium trifluoride is either not reduced to metallic chromium or that the metal obtained is in such a finely divided and reactive state that it dissolves in the leach so- lution, However, previous tests weuld indicate incomplete reduction, probably to divalent chromium fluoride. Since EDTA solutions complex both di- and tri- valent chromium, other complexing agents were studied to distinguish between CrfF, and CrF . Preliminary experimenis have indicqteé that a 0.17 M (2.5%) solution of ammonium oxalate will readily dissclve CrF_ at the refluxing temperature, and a purple complex will form. Metallic chromium, CrF,, and Cr 03 appear to be insoluble, and they show no te«néency to form complexes. The ab- sorption spectrum of the complex formed by divalent chromium fluoride and ammonium oxalate is identical with that of the complex formed by the recction between trivalent chromium chloride and ammaonium oxalate. Further investigation of the nature of this complex is ynder way. The work on the determination of the oxidation states of chromium in reactor fuels has shown that (1) divalent chromium fluoride can apparently be selectively leached with 0,17 M ammonium oxalate, (2) divalent and trivalent chromium fluoride can be leached simultanecusly with EDTA, and (3) the metal and the trivalent oxide are not attacked by either of these solvents. Determination of UF3 and UF’4 W. J. Ross Analytical Chemistry Division Work on the development of a msthod for the de- termination of UF, and UF_ in NeZrF, hos been continued, and the more complete studies show, as reporfed previously, that EDTA dissolves UF, but apparently hes little effect on UF,. The cftects of temperature and the concentration and acidity of EDTA on the recction between UF, and EDTA were investigated. Concentration o? the reagent was not critical over the range 0.08 to 0,15 M; a concentration of 0.1 M was used in most of the tests. Solutions of EDTA with acid concentrations of pH 4 were not appreciably more effective in dissolving UF, than thoss with acid concentrations approaching a pH of 7. Use of solutions less acidic tharn pH 7resulted inhydrclysis of uranium. Temper- ature was the most critical condition. The rate of reaction at 100°C was approximately ten times that at 25°C; however, the data obtained at 25°C were not reproducible because of variations in particle size. ‘ - Under various conditions of temperature uand acidity, the absorption spectra of UF -EDTA are essentially identical with those for UF, dissolved in HCl, except that the spectra for the 4UF4 EDTA systems are shifted by approximately 20 my toward the infrared by the presence of EDTA. The maxi- mum molar extinction coefficients for the four spectra shown are nearly equivalent; they range from 35.7 for UF -HC| to 25.0 for UF -EDTA at a pH of 6.8 and a tempemture of 100°C. Petrographic and chemical analyses of the three batches of UF:3 used for the study of solubility in 0.1 M EDTA showed UF, contents of 92.4, 90.4, and 90.7%, respectively. %ampies of these batches were treated with 0.2 M ammonium oxalate at reflux temperature and filtered, and petrographic exami- nation revealed that the residues were essentially free from UF,. The data indicate that during the first hour of reflux the greater portion of UF , present is removed but that quantitative separation is questionable. The solubility of UF, oxalate is low; however, UF, may oxidize slowly upon contact with this solvent, as indicated by the appreciable concentration of uranium in the solution when allowed to stand overnight. The UF content of the sample which had been treated wn‘h oxala’re is being determined and is expected to be near 100%. On the basis of the data obtained thus far, UF and UF, can be separated by ammonium oxuiate. However, the method is not sufficiently precise for the separation of traces of UF, from UF,. Similar tests are under way to determine the feasibility of using 0.1 M EDTA for separating UF . in ammonium Reducing Power of NaZ(F . with Zirconium Addition D. L. Manning Analytical Chemistry Division The ftotal reducing power of NaZrF, specially prepared by treatment with metallic zlrconlum was determined. In the hydrogen evolution method?® ysed for this determination, the sample is treated with 0.2 M HF to liberate the hydrogen which is formed by the reaction between zirconium and hydrofluoric 3. C. White and W. J. Ross, ANP Quar. Prog. Rep. March 10, 1953, ORNL-1515, p. 173. PERICD ENDING DECEMBER 10, 1953 the gas is collected and measured; and the reducing power is calculated. Essentially no re- ducing power was observed in these somples. A second method which involved complete dissolution of the sample in standard ceric suifate solution in 9 M H2504 and back-titration with standord ferrous sulfate solution was used to confirm the results. Further tests by the hydrogen evelution method on synthetic samples of NaZrF_ and zirconium demonw strated that the method is quantitative, acid; Dissolution of Fluoride Mixtures Containing Zirconium P. V. Hoffman Analytical Chemistry Division Nitric and boric acid mixtures have been used successfully as solvents for fluoride mixtures, but, in an effort to find a less-corrosive solvent, small- sized lumps of NaF-ZrF --Ut':4 (50-46-4 mole %) were contacted with varlous concentrations of ammonium oxalate, oxalic acid, and acetic acid at 70 to 80°C for various periods of time. The optimum concentration of ammonium oxalate was 1 M, which is approximately o saturated solution at 50°C. [(One hundred ml of T M (NH ) C O M O will dissolve 1 g of small lumps of Fuel un upproxnmcately 2 hr.] A 2 M solution was the most effective oxalic acid concentration and was approximately equivalent to 1 M ammonium oxalate in dissolution effectiveness. The optimum acetic acid concen- tration was 0.2 M, which yielded only 50% of the solubility strength of the oxalates. A comparison of the effectiveness of the three reagents tested is shown in Fig. 10.1. The rate of dissolution with 1 M ammonium oxalate is greater than the rate with 2° M oxalic acid, although their effectiveness is nearly equivalent. Ammonium oxalate (1.0 M), because of its higher pH {6.5), is to be preferred over 2 M oxalic acid (pH (.7), however, with respect fo possible corrosiveness. None of these solvents are svitable, however, for dissolving large-sized lumps of Nan-ZrFA-UF4 from metallic containers. PETROGRAPHIC EXAMINATION OF FLUOR?DES G. D. White Metallurgy Division T. N. McVay Consultant, Metaliurgy Division Petrographic examinations were made of fluoride fuel mixtures NaF-ZrFé-UFd (50-46-4 mole %) and 109 ANP QUARTERLY PROGRESS REPORT i 100 OWG. 22221 i | T 14 (I“JH4)2C204""*"‘-\\‘>// ap __..__‘;_________i // // ________ ] | | / //2 M 1,0,0, 80 7 — 7o // R .. 60 // : S SR % ‘ £ 0.2 M HC,H30, ‘ 2 '3¥e | a | S 50 — e w a n 3 / 3 40 HH—-- / ______ 30 // ————————— - 20 R —_ 10 H— e o ‘ 0 0.50 100 1.50 2.00 TIME (fr) Fig. 10.). Comparisen of Effectiveness of Solu- tions of Ammonium Oxclate, Oxalic Acid, ond Acetic Acid as Solvents for Nr.aF-Z’.rFA-UF4 (50-46-4 mole %). (53.5-40-6.5 mole %) which had been irradiated a maximum of 492 hr in the MTR. Unirradiated conirol samples were examined at the same time. These fuels are normally composed of Nqu(U)Fs, which is green, and perhaps a very smallamount ofa lower refractive index complex, which is designated E-2. The irradiated samples were normal, except that the NaZr(U)F_ had a slightly low refractive index and was slightly reduced. There were no differences between the control samples and the irradiated samples which could be observed with a petrographic microscope. MASS SPECTROMETER IMYESTIGATIONS OF IRRADIATED FLUORIDE FUELS C. R. Baldock Stable |sctope Research and Production Division The mass spectrometer has been successfully employed in the quantitative anclysis of the uranium content of irradiated fluoride fuel mixtures. Initially, the mass spectrometer method was used to test the validity of chemical analyses, but it was recognized that the method could be extended to obtain an accurote measure of fuel burnup. In addition, by using accepted values of cross sections,an average valve of fthe flux in which the fuel has been ir- radiated can be determined. This gives an in- dependent check of values predicted from other types of measurements and of the internal con- sistency of the celculations. The isotope dilution method as applied to the analysis of irradiated fuel has been used extensively with a routine accuracy of £1%, which is not the ultimate limit of accuracy. Calcuvlation of UF4 in Unirradioted Fuels For the calculation of the percentage of UF, (by weight} in unirradiated fluoride fue! samples of known isotopic composition, the exact values for the isotopic masses are used to compute the atomic weight of the uranium in the fuel: (1) At.wt = 235124, + 234124, where A_ denotes abundance of uranium isotope of mass number m. The ratio, R, of U235 to U?38 js obtained from the mass spectrometer measurement, and since the atoms of U235 gnd U238 gre obtained from both the vranium in the fuel and the natural U3OB spike materia! added, (atoms of U235 from fuel) + (atoms of U235 from spike) (2) R = (atoms of U238 from fuel) + (atoms of U238 from spike) 110 The atomic weight of uranium in the spike is 238.07, and therefore W1F1YA235 W2F2A235 @) 2 at. wt 238.07 = ? W F YA, . WoFaAaag af. wt 238.07 where W, = weight of uranium-bearing fuel, W, = weight of U308 spike material, F, = U/UI"*"4 = gravimetric factor for uranium in fuel, F, = 3U/U,0; = gravimetric factor for urani- um in spike, Y » 100 = percentage of UF, (by weight) in fuel. if Eq. 3 is solved for Y, WoFdsss WoF 4,38 ) v 238.07 238.07 W1F1A233 W1F1A235 at. wi at. wt (szz) (at. wt) (Azas - RA238> Calculation of U235 Lost from Irradiated Fuel if C is defined as the number of gram atomic weights of U235 in W, grams of fuel before ir- radiation, then (5) c WIF!YA235 at. wt and, for the irradiated spiked fuel sample, the following equation must hold: W,F A 2235 + — s @ R = 238.07 Wi F\ Y435 Waladysg at. wt 238.07 where X is the gram atomic weight of the U235 present in W, grams of fuel after irradiafion. Solving Eq. 6 for X gives PERIOD ENDING DECEMBER 10, 1953 WrFIYAzaa WzeAzas (7) X = R ————~+ R ———me at. wt 238.07 wzeAzss 238.07 Then, the gram atomic weight of the U233 lost from W, grams of fuel, from Eqs. 5and 7, is ] W1 F1 Y RA @) C - X = (Ayg5 — RAygy) af. wt W2F2 238.07 and the percenfage of U235 lost from the fuel follows as C - X —x 100 . C Determination of U235 Burnup Equation givesonly the total percentage of U235 lost from the fuel. However, this loss may be due to several factors, including burnup, radiation damage, and replacement of chromium leached from the Inconel walls, A graphical method has been developed to test this loss to determine whether it can be completely ascribed to the capture of neutrons to produce fission preducts and U236, that is, to burnup. The solution of the differential equation for the first-order rate expression for burnup is found to be ug?s (a, + a) et 9 IO = 7 ® ¥ 235 2.303 i where a, = fission cross section =580 barns, a, = capture cross section = 105 barns, N e = neutron fluxdensity=2.17 x10"¥n/cm?.sec, - G- H irradiation time, sec, and the slope is (a, + az) b 10 e = 6.4544 x 1078 (10) 2.303 111 Therefore U(2)35 (1) log = U235 ! 6.4544 x 10~8; . If a plot is made of Eq. 11, which might be called the theoretical curve, it is possible to compare the experimental results on burnup with the results predicted by theory, In addition, Eq. 11 can be used to check the product of the three constants Qyp Oy and ¢. Or, if the (a, * a,) values are assumed to be correct, Eq. 11 can be used as a check on ¢, the averoge flux density. The values of a, = 580 barns and a, = 105 barns are rather well established. The value of ¢ = 2.17 x 1014 n/cm2.sec has been tentatively chosen because it gives the best fit to the data based on samples from four test capsules and it is consistent with values determined by other means, SUMMARY OF SERVICE CHEMICAL ANALYSES J. C. White A. F. Roemer C. R. Williams Anclytical Chemistry Division Sixty-four samples of UF, and ten samples of NGQUF6 (the ARE fuel concentrate) have bheen analyzed to date. The UF | samples were analyzed for uranium only, but comp?ate analyses were made of the fuel concentrates, including determinations TABLE 10.1. of iron, nickel, and chromium. The cooperative program with the {_aboratory Division of Y-12 was continued on the determination of uranium, as described in the previous report.? The relative difference between loboratories on the determi- nation of uranium in the fue!l concentrate was 0.04%, ond the difference on the UF4 analyses was 0.18%. The larger difference was ottributed mainly to the heterogeneity of the samples of UF . which contained black particles and comparatively large hardpieces ofa lighter hue than that of the powdered salt; the large pieces were probably UO,F,, Uranium metal that was considered to be 99.94% pure was used as the standard. The relative standard error of 12 samples at a 95% confidence level was 0.02%. Some specimens of Inconel metal used in corrosion tests were analyzed for uranium, zirconivm, sodium, and fluoride and the major constituents nickel, iron, and chromium. Traces of sodium carbonate in the range of 0,05% and less in sodium hydroxide were determined. The major portion of the work wos concerned with the anolysis of fluoride fue! mixtures. A decided increase occurred in the number of samples received during the quarter. A total of 1,378 somples waos received, 1,201 were reported, and 11,034 determi- nations were made {Table 10.1). 4. C. White er ol., ANP Quar. Prog. Rep. June 10, 1953, ORNL«1609, p- 121. SUMMARY OF SERVICE ANALYSES REFPORTED NUMBER OF NUMBER OF SAMPLES DETERMINATIONS Experimental Engineering 595 6,411 Corrosion Studies 271 3,247 Reactor Chemistry 256 1,000 ARE Fluid Cirguit 74 328 Heat Transfer and Physical Properties Studies 4 48 1,201 11,034 112 INTRODUCTION AND SUMMARY E. P. Blizard, Physics Division The spectra of gamma rays and of fast neutrons from the Bulk Shielding Reactor were measured (sec. 11). The neutron spectrometer was altered for this work to improve the sensitivity at some sacrifice water shields can now be studied. The gamma-ray spectrum of the Bulk Shielding Reactor without a shield showed the usual broad peaks at 2.2 Mev that are due to water capture and those at 7 to 8 Mev that are due to capture in aluminum and possi- bly other materials., The fast-neutron leakage spectra of the BSR were obtained from profon-recoil measurements for the water-reflected reactor {oad- ing No. 22, Data obtained with the nuclear plate camera have been used to extend the spectrometer data from 1.3 Mev down to 0.5 Mev. A measure- ment of the angular distribution of fast neutrons in water was also made. At neutron energies above 8 Mev, the intensities at various angles converge. The Lid Tank Facility was used for studying the slant penetration of fission neutrons in a hydro- genous shield (sec. 12). This problem is encoun- tered in specifying a crew-shield side-wall thick- ness. 1he uncertainty lay in the importance of the short circuiting of the shield by scattering into paths more nearly normal to the shield surfaces. For the shield thicknesses which have been meas- ured, it appears thot this short circuiting is un- important, and the attenuation is characteristic of the slant thickness. This is not unexpected for the thin shields (< 7.5 ¢m) which have been meas- ured to date. There is as yet nothing conclusive about the behavior of very thick shields with slant-incidence radiation, but it will be advan- tageous if the slant thickness gives the ottenuation for the thicker shields, The Lid Tank Facility was also used for air duct experiments designed to give some insight into the interaction between paralle! ducts (sec. 12). Data from three sources, (1) a duct consisting of three sections, 22 in. long and 3,5/] in. in diameter, joined at angles of 45 deg, (2) a second duct con- sisting of two such sections, and (3) the siant penetration experiments, were put into the Simon- Clifford formula, and the formula was solved for the reflection factor. The values obtained were in agreement with those obtained from previous measurements and thus constituted an in energy resolution; however, thicker thermal important confirmation of those measurements. It is now felt that engineering designs based on the Simon-Clifford theory are on a sound basis when applied to-a single duct, especiaily when the di- mensions are similar fo those studied. Since it was observed that the peak values of dose meas- vred at the end of the 19-duct array were a factor of approximately 10 higher than those observed with the single duct, o study of interference be- tween adjacent ducts has been started. Tests are planned to give more quantitative information on the interaction between neighboring ducts. and possibly to improve the formula for streaming through ducts. The effective removal cross section work for which the Lid Tank Facility is used was con- tinved, and old experiments have been reviewed. A new set of values is reported (sec. 12). A summary of the first 30 L.id Tank experiments is being prepared (sec. 12). The summary will be useful because of the large role the Lid Tank Facility has had in the development of various reactor shield designs and because of the constant need for reference to the data. The stee! structure of the Tower Shielding Fa- cility has been erected, and the ground structures are approximately 75% complete (sec. 13). The optimistic date for completion of the facility is Jonuary 15. Some basic shielding research which is not re- ported here because it is supported by another activity is, nevertheless, of vital interest to ANP. The development of a scintillating fast-neutron spectrometer has received considerable impetus recently with the growth of o few clear europium- activated lithium iodide crystals. These crystals have been tested on the ORNL Van de Graaff generator and show encouraging energy sensitivity, but an anomalous lack of energy resolution indi- cates that the development is far from complete. Most of the theoretical shielding work has been reported in the Physics Division Semiannual Re- port with the exception of o few problems which originated in the experimental program (sec. 14). An unusual calculation has been made of the vis- ible light to be expected from divided-shield aircraft. An estimate of the neutron reflection coefficient for water was obtained by using @ simple model with a coincident source and a nondirectional receiver located above the water surface. 115 ANP QUARTERLY PROGRESS REPORT 11. BULK SHIELDING REACTOR R. G. Cochran F. C. Maienschein H. E. Hungerford E. B. Johnson T. A. Love G. M. McCammon Physics Division During this quarter, the main effort at the Bulk Shielding Facility was directed toward obtaining neutron and gomma-ray spectroscopic data. The gamma-ray leakage spectrum of the Bulk Shielding Reactor has been meosured becouse it is typical of the spectrum of gamma radiation incident upon a shield. | In addition, considerable fast-neutron spectral data were obtained on the present loading for the water-moderated reactor. The measurement of the energy spectrum of neutrons attenuated by various thicknesses of water was repeated and extended to a 50-cm distance from the face of the reactor. A meosurement of the angular distribution of fast neutrons in water has been made, and the attenuated neutron spectrum through o shield mockup was measwed, SPECTRUM OF GAMMA RAYS EMITTED BY THE BSR! Results of measurements of the gamma-ray energy spectra and angular distributions through lead and water were published ;::ria‘vic.n,:sly.2 in the following, the results of measurements of the gamma-ray spectrum of the Bulk Shielding Reactor without a shield are presented. These measurements may be considered as an extension of the earlier work. The measurement of the spectrum from the un- shielded reactor was necessarily delayed until the BSR was loaded with “‘cold’’ fue! elements, that is, elements which had not been used when the reactor power was higher than 3 watts. The low power level of the reactor was dictated by the counting-rate limitation of the spectrometer. With 1F. Maienschein and T. A. Love, Spectrum of Gamma Rays Emitted by the Bulk Shielding Recctor, ORNL CF 53-10-16 (to be published). 2r, Maienschein, Gommo-Ray Spectral Measure- ments with the Divided Shield Mockup, Part {, ORNL CF 52-3-1 (Mar. 3, 1952); Part 1l, ORNL CF 52-7-71 (July 8, 1952); Part |1}, ORNL CF 52-8-38 (Aug. 8, 1952); and T. A. Love, Part IV, ORNL CF 52-11-124 {(Nav. 17, 1952). 116 the reactor ‘‘hot,’’ the power level from fission- product activity alone would have been too great. For this particular experiment, a 6 by 5 fuel element arrangement was used in on unreflected reactor. The total power and the power distribution were determined with gold foils in the usudl manner. Spectrometer Arrangement. One end of on air- filled oluminum tube was placed ogainst the face of the reactor, and the tube extended to the three- crystal gamma-ray spectrometer® (Fig. 11.1). For the first run, the air tube was only 5 in. long ond was surrounded by a 5-in. thickness of lead which shielded the spectrometer proper from gamma radia- tion from other parts of the reoctor. The results of this run showed that the neutron-induced back- ground was so high os to moke a meaningful in- terpretation of the dota impractical, In o second run, o 40-in. air column surrounded by water, except for the 5-in. thickness of lead at the end next to the spectrometer, served as a collimator. The additional water shielding reduced the neutron level sufficiently for the nevtron-induced background to be negligibly small. The total back- ground was measured by plugging the collimator within the spectrometer with lead. It was also found that the background from reactor fission products was small, Results, The absolute photon (gamma-ray) flux obtained is plotted as a function of photon energy in Fig. 11.2. The flux volues were determined from the known sensitivity of the spectrometer® and from the solid angle defined by the collimator shown in Fig. 11.1. Since the water surrounding the aluminum tube was not very effective in pre- venting gomma rays from entering the sides of the air column, the spectrum shown in Fig. 11.2 is too large by a small fraction, perhaps 20%. 3F. C. Maienschein, Multiple-Crystal Gamma-Roy Spectrometer, ORNL-1142 (July 3, 1952). PERIOD ENDING DECEMBER 10, 1953 UNCLASSIFIED DWG. 24656 THREE-CRYSTAL GAMMA-RAY SPECTROMETER- 7 '\\. / ; /7 \\- ///’ S . SIS W, . ! /{ i t5 in. BSR T T I ””{_ l ZiALUMINUM TUBE FHLED WITH AR MREMOVAE!& LEAD PLUG Fig. 11.1. Experimental Arrangement for Measurement of Bulk Shielding Reactor Gamma-Ray Spectrum, UNCLASSIFED OWG. 2657 ,,,,,,,,, o n r {photons femEssec/ Mev wott Ssteradian) o 0 1 2 3 4 5 6 7 8 GAMMA-RAY ENERGY (Mev) Fig. 11.2, Spectrum of Gamma Rays from the Bulk Shielding Reactor. The spectrum shape shows the usual broad peaks at 2.2 Mev that are due to water capture and those at 7 to 8 Mev that are due to capture in aluminum and possibly other materiadls. The couse of the 0.4-Mev pedak, which appears to be redl, has not been found. The 0.4%-Mev pedk, however, may well be due to gamma rays from the 0.478-Mev level in the Li’ formed by neutron capture in the boron liner of the spectrometer, ‘ The smooth curve in Fig. 11.2 represents the spectral shape (not intensity} obtained by Mot z4 4). W. Motz, Phys. Rev. 86, 753 (1952). with a Compton-recoil magnetic spectrometer look- ing at a U233 slug in the center of the Los Alamos water boiler and one ot the active core of the fast reactor. The curves cannot be expected to corre- spond closely because of the differences in neutron- capture gamma-ray sources and because the low resolution of the spectrometer used by Motz would smooth out any structure present. The generdl similarity in slopes, however, is interesting. FAST-NEUTRON LEAKAGE SPECTRA OF THE BSR The neutron spectrum of the BSR obtained with the reactor surrounded by cans of beryllium oxide as a reflector® was reported previously. For the data reported here, the previous spectral measure- ments were repeafed and additional measurements were made; however, a waterweflected reactor (loading Né. 22) was used. The measurements were made with the BSF proton-recoil fast-neutron spectrometer,’ Except for the spectrum at the face of the reactor, the spectral measurements in water out to 30 cm from the fuel® are in agreement with the previous meas- urements. The spectrum at the face of the reactor SR. G. Cochran aond K. M. Henry, Fast Neutron Spec- trum of the Bulk Shielding Reactor, Part |, ORNL CF 53-5-105 (to be published). ' 5R. G. Cochran et al., ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL-1609, Fig. 11,1, p. 127. TR- G, Cochran and K. M. Henry, A Proton Recoil Type FastNeutron Spectrameter, ORNL.-1472 {Apr. 2, 1953). 8p. 6. Cochran and K. M. Henry, Fuast Neutron Spec- trum of the Bulk Shielding Reactor, Part 1I, ORNL CF 53-11-45 (to be published), 117 ANP QUARTERLY PROGRESS REPORT In addition to the proton- data obtained with the have been used to extend the spectrometer data from 1.3 Mev down to 0.5 Mev. is shown in Fig. 11.3. recoil measurements, nuclear plate camera’® Meoswements were also mode out to 50 em from the reactor. The spectrum measured at 50 cm is shown in Fig. 11.4; the statistics on these data are quite poor, since the neutron intensity was approximately at the limit of sensitivity of the speciromater. An attempt was also made to measure the spec- trum of neutrons emerging from the side of an air- craft divided-shield mockup.'® The mockup con- sisted of a slab of lead 4% in. thick placed 16 in. from the reactor. The shield therefore extended out through the water to a position 113 cm from the reactor. Unfortunately, the neutron spectrometer was not sufficiently sensitive to detect neutrons at 113 em; however, a spectral measurement was made at 70 cm, but the statistics (Fig. 11.5) were rather poor. A measurement of the angular distribution of the fast neutrons in water was made during the M. P. Huydon, E. B. Johnson, and J. L. Meem, Measurement of the Fast Neutron Spectrum of the Bulk Shielding Reactor Using Nuclear Plates, ORNL CF 53-8-146 {10 be published). Repori‘ of the ANP Shielding Boord for the Aircraft Nuclear Propulsion Program, ANP.53 (Oct. 16, 1950). L DWG 2184C S O RECOIL SPECTROMETER DfxTA END OF * COLLIMATOR AGAINST REACTOR O NUCLEAR PLATE CAMERA DaTA | ] PLED (nsuhons/cm?/sec/woH/Mev 103 & 7 9 10 ENERGY {(Mev) Fig. 11.3. Meutron Spectrum ot the Face of the B3R, 118 quarter. H The experimental setup for this meas- urement is shown in Fig. 11.6, and the spectra vs. angle data are plotted in Fig. 11.7. At neutron energies above 8 Mev, the intensities af the various angles converge. This effect is probably due to neutrons which bypass paort or all of the collimator. A distance greater than 10 cm from the reactor would be more desirable for this measurement, but then it would be difficult to obtain sufficient in- tensity. Figure 11.8 shows the variation of flux with direction for each of several neutron energies. ]R. G. Cochran and K. M. Henry, Angulor Distriby- tion of Fast Neutrons in Warer, ORNL CF 53-11-46 (ic be published). DWG. 24921 e . . .o - e 50 cm Hy(? BETWEEN I IREACTOR AND COLLIMATOR, | 4-ft COLLIMATOR | &(£) ineutrons/cm2/sec/ watt/Mev X 1073) ENERGY [Mev) Fig. 11.4. Neutron Spectrum 50 cm from the BSR, 1 ey e _,,,,DWG.,EE?Oi ; | 70 em W0 BETWELN ! E XPERIMENT i " |1 | REACTOR AnD COLUMATOR. . en @ LEAD =7 2-ft COLLIMATOR T e o o | : ' : 3SR 4 1,0 r———ij 2 | ‘ b ST .\\ S ; NEUTRON /| g 8 SPECTROMETER” | =~ [ g ‘ < T E < = g 5 g, é 4 = E ' w2 ! ENERGY (Mev) Fig. 11.5. Nevtron Spectrum 70 cm from BSR. FUEL ELEMENT -~ _ 10 ¢m per—- BULK SHIELDING REACTOR PERIOD ENDING DECEMBER 186, 1953 OWG, 22199 4 ft emaasa -- e e - - e f/" 1-in. C_OLLIMATOR NEUTRON T e SPECTROMETER N HOUSING \\ . e e \\\ ; Yol T - ’Q deg e T T e -~ - \\\\ \\ - S - 2 T L0 . £ ~. =5 o A .k ", \EQN ™~ DIRECTION ©OF ROTATION TOP VIEW Fig. 11.6. Experimental Arrangement for Me asurement of Angular Distribution of Neutrons from BSR in Water. $EY { PEUITORS / GMT/gec/ watt/ Mey i END OF COLLIMATOR 10 cm FROM :REACTCR NEUTRON ENERGY {Mev) A DWG. 21843 Fig. 11.7. Angulor Distribution of Neutrons from BSR in Water, W DWG. 21842 b (£) (neutrons /cm/sec/ watt/ Mev) >/ / "; L /z, i EAP - \\ /\ // \/ /// // - li i’:"t‘i»-/_‘ ’f /:’/ 207 [REACTOR N\ T S o / 7\./ /7T o0° 50° 30° 25e Fig. 11.8. Energy Angular Distribution of Neutrons from BSR in Water, 119 ANP QUARTERLY PROGRESS REPORT 12. LID TANK FACILITY C. L. Storrs - GE-ANP G. T. Chapman J. M. Miller D. K. Trubey F. N. Watson Physics Division Tentative results of a preliminary experiment performed at the Lid Tank indicate that most of the air-scattered neutrons that penetrate an aircraft crew shield follow the slant path. In another experiment, the study of neutron streaming through air ducts was resumed; data from the initial experi- ments have presented on important confirmation of the Simon-Clifford theory. A re-evaluation of all measurements for the determination of removal cross Refer~ ence to Lid Tank neutron and gamma data is facili- tated by a recent survey of the first 50 experiments. Approval has been received for the use of enriched vranium for a new Lid Tank source plate, and a design has beendeveloped which will make possible sections has improved some of the values, a far more accurate power calibration. The source plate, as well as its appurtenant power-measuring equipment, is at present being constructed in the ORNL Research Shops. It is planned to install the source plate in January. SLANT PENETRATION OF NEUTRONS THROUGH WATER A preliminary experiment has been performed to determine the slant penetration of fission neutrons through water. An air-filled cluminum duct, 40 in. long and 2 in. in diameter, was used as a collimated “point’’ source of fast neutrons, and the fast- neutron dose was determined in the water around the end. The duct was tilted from the normal to the source plate to reduce background of fast neutrons from the ORNL Graphite Reactor. Since the fast neutrons are collimated in the outward direction, if not reduced they would overshadow the fission-plate neutrons which are, of course, isotropic. The data, plotted as isodoses oround the duct, are shown in Fig. 12.1. By integrating the dose along planes through the radiation thicknesses of water at various angles was de- termined (Fig. 12.2). This information must be regarded as tentative until the dose determinations field, the dose penetrating different 120 have been repeated with more care, but the results obtained so far indicate that most of the dose comes from neutrons that penetrate the shield along rthe slant path. This indication is similar to that obtained for gamma rays for comparable attenu- ations.! The observed relaxation lengths range from 4.9 to 6.3 cm. As would be expected, these values are smaller than those obtained with thick shields, since a fission spectrum was used as a source and the shields were quite thin. IF. 5. Kirn, R. ). Kennedy, and H. 0. Wyckoff, Oblique Attenuation of Gamma-Rays from Cobalt-60 and Cesium- 137 in Pclyethylene, Concrete and Lead, MNBS-2125 (Pec. 23, 1952). RN DWG. 2iB74 No |- e z, DISTANCE FROM SOURCE (cm) T 9o | oot o 40 30 20 o Fige 12,1. Fast-Neutron Dose Distribution Off the End of a Duct 101.6 cm Long by 5,08 e¢m in Diameter. 7 s— DWE. 21876 RELATIVE NEUTRON DOSE (orbitrary units) 7, SLANT THICKNESS (em) Fig. Slant Penetration of Neutrons in Water. 12.2. To determine the total air-scattered dose in a divided shield, it is necessary to integrate over all angles. Such an integration would presumably give a straight-through relaxation length comparable to that observed in the BSR air-scattering experi- ment.? 1t would be desirable to extend the measure- ments to greater thicknesses of water and to harden the spectrum by putting water between the source and the duct entrance. Intensity problems limit the extent to which this can be accomplished, however, AIR DUCT TESTS The use of the Lid Tank Facility for the study of neutron streaming through air ducts has been resumed with emphasis on fast-neutron measure- ments. A flow-type dosimeter with somewhat greater sensitivity than that available heretofore was used to determine the dose around the ends of several ducts. One of these ducts was taken from 2H. E. Hungerford, The Skyshine Experiments at the Bulk Shielding Facility, ORNL-1611 (to be published). PERIOD ENDING DECEMBER 10, 1953 the 19-duct array studied a year ago.? |t had three sections, 22 in. long and 3% joined at angles of 45 degrees. A second duct consisted of two such sections. The slant pene- tration experiment described in the preceding sub- section provided a third measurement. Putting the data from these three measurements into the Simon-Clifford formula? and solving for the reflection factor, a, gave values of 1.0 and 0.9, respectively, for the first and second bends. These values are in agreement with the velue of about 1 obtained from the previous thermal measurements® made in water at a distance of 30 c¢m from the ends of the ducts (thus the emergent flux was filtered and was representative of the hard neutrons that de- termine the dose), and they constitute an important confirmation of those measurements. The value of 1 has been assumed in recent direct-cycle design studies. It is now felt that engineering designs based on the Simon-Clifford theory are on a sound basis when applied to a single duct, especially when the dimensions are similar to those studied. Since it was observed that the peak values of dose measured at the end of the 19-duct array were a factor of approximately 10 higher than those ob- served with the single duct, a study of interference between adjacent ducts was indicated. This study has been started, with a single section and o three-section duct being used. In the first con- figuration, one of the possible streaming paths was reproduced in the array by placing the short duct parallel to the first section of the long duct and 12!{2 in. from it. In this position, it lined up, approximately, with the last section. The result was an increase by a factor of 4,5 in the dose at the end of the long duct. 4 The short duct was then moved to a position that was again parallel to the first section but con- siderably closer (§ in.) and above it in a different plane so that there was no direct streaming path. The data (which must be considered tfentative) showed no increase in the peak dose beyond the long duct but did show a broadening of the emergent nevtron beam. The results obtained from these experiments are shown in Fig. 12.3. in. in diameter, 3,. . Flynan and C. L. Storrs, Radiation Measurements Aroundan Array of Cylindrical Ducts, ORNL CF-52-12-187 (Dec. 15, 1952). 7 4A, Simon and C. E. Clifford, The Attenuation of Neutrons by Air Ducts in Shields, ORNL-1217 (Nov. 11, 1952). SC. E. Clifford, private communication. 121 ANP QUARTERLY PROGRESS REPORT DWG. 21918 current C F] measurements, gives a removal cross section of 1,31 barns/atom for fluorine as compared with the value of 1,36 barns/atom reported in the previous report. A re-examination of all the Lid Tank measurements for the determination of remeval cross sections has improved some of the values. A summary of the presently accepted values obtained from direct measureaments is given in Table 12.1. Additional values based on the direct measurements are given in Table 12.2. TABLE 12.1. EFFECTIVE FAST-NEUTRON REMOVAL CROSS SECTIONM VALUES OBTAINED FROM LD TANK DIRECT MEASUREMENTS EOSL oot T 0. E ,,,,,, E L 2 e w Q O .».1 ~ z 10 bt o = - w5 TWO-BEND DUCT A+ - —— 1+ .. = WITH SINGLE SECTION g DUCT IN LINE /T ———T7 2 b TWGO-BEND DUCT 1072 WITH SINGLE SEC-- . TION DUCT NOT IN LINE TWO-BEND DUGT ol L b 1072 -20 -18 -2 -6 o 6 12 18 20 ». DISTANCE FROM CENTER OF END OF DUCT (cm) Fig. 12.3. Fast-Neutron Dose Beyond Yarious Configurations of Ducts with 45-deg Bends. Further tests are planned to give more quantitative information on the inferaction between neighboring ducts and possibly to improve the formula for streaming through ducts. REMOVAL CROSS SECTIONS The measurement of the effective fasteneutron removal cross section of C_F . has been repeated with the use of 15 in. of the fiquicl instead of 28 in. as in the previous measurement.® Despite this large difference in thickness, the measured values of the molecular removal cross section agreed to within 5%, which is within the experimental error, and thus they provide an interesting check on the validity of the geometrical corrections that were made. The revised value for carbon, 0.80 barn/atom instead of 0.84 barn/atom, when used with the 6C. L. Storrs et al., ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL-1609, p. 145-148. 122 MATERIAL gn Al 1.31 barns/atom Be 1.05 Bi 3.43 C 0.80 Cu 2,08 Fe 1.95 Ni 1.85 W 2.6 + 0.2 C7F16 26.6 barns/molecule C,FyCi 6.50 34(: 4,26 Hzo* (2.90) 520* (2.78) * Neither hydrogen nor deuterium haos, properly, an effective remova! cross section, since the total cross section is not nearly constant in the high-energy region. Rough values have been obtained for H20 and 020 bosed on 140~cm shields. rentheses, These are reported In pa- TABLE 12,2. EFFECTIVE FAST-NEUTRON REMOVAL CROSS SECTIONS BASED ON LID TANK DIRECT MEASUREMENTS MATERIAL URI {(barns/atom) B 0.87 D* (0.92) F .31 H* (0.98) 0 0.94 1'rCf., footnote to Table 12.1. w SURVYEY OF LID TANK EXPERIMENTS Shielding experimentation at the ORNL Lid Tank Facility began in June 1949, and since that time approximately 60 major experiments have been performed. In view of the large role this facility has had in the development of various reactor shield designs and because of the constant need for reference to the data, it was felt that the experi- mental resuvlts should be compiled in one publication. Accordingly, o summary of the first 50 Lid Tank experiments will socon be presented by E. L. Czapek.”:8 ‘ As described in several previous reports, the Lid Tank Facility consists of a large water tank adjacent to an opening in the concrefe shield of the TORSORT student from Electric Boat Division of General Dynamics Corporation. 8. L. Czapek, General Survey of Lid Tank Experi- ments 1 Through 50, ORNL-1636 {to be published). ' PERIOD ENDING DECEMBER 10, 1953 ORNL. Graphite Reactor (Fig. 12.4). A converter plate of utanium covers the opening between the water tank and the reactor shield. This plate absorbs most of the incident thermal neutrons emerging from thereactorand causes fission, which, in turn, produces the known uranium spectrum of neutrons, as well as gamma rays of various energies. Shielding samples are inserted in the tank, and the transmitted radiation is measured behind the shield. The geometry of the Lid Tank permits the use of very large samples, which, in turn, help to reduce uncertainties due to boundary conditions. The water in the tank protects personnel from radiation and allows easy positioning of test samples. The Lid Tank Facility was built to test both compositioris and geometries of shields. With the resuiting experimental data, difficult attenuation problems could be quickly solved and shield thick~ nesses couid be predicted {to within an accuracy of Fig. 12.4. View of Lid Tank Focility from Above. 123 ANP QUARTERLY PROGRESS REPORT +5%) without having to rely too much on a knowledge of the attenuation processes. Experimentation at this focility has primarily been limited to investigations of the ‘‘unit shield’’ in which all material that functions primarily as shielding is located around the reactor. All the experiments covered in the survey can be grouped as follows: 1. Water Shield (experiments 1, 4, 10, 13) 2. Preliminary |nvestigations” (experiments 2, 3, 5,6, 8,10) a. Lead-water shield b. lron-water shield 3. Boron Carbide Shields (experiments 11, 12) 4. Thermal Shields? (experiments 10, 20, 36) a. lLead-water shield b. lron-water shield 5. Shield Mockups (experiments 7, 14, 15, 16, 17, 18, 29, 45) d. Submorine Thermal Reactor b. Submarine |ntermediate Reactor c. Reflector-Moderated Reactor (ANP) 6. Shield OQOptimizations (experiments 9, 13, 24) a. Two-component shield of lead and borated water b. Suppression of capture gammas in lead-water shield c. Optimum position of a lead-iron gomma shield in relation to pressure shell YA small amount of boron was present in a majority of these shield mockups. 124 7. Shield Penetrations (experiments 19, 21, 22, 25, 26, 37, and 38 and special experiments) a. Cylindrical air ducts with or without bends b. Array of cylindrical air ducts with or without bends ¢. Annularduct (specialized mockups simulating GE-R1 air outlet and inlet sections) d. Structural members in shields (streaming effects) e. Instrument plug (modification in shield to improve control instrumentation) 8. Removal Cross Sections (experiments 23, 30, 31, 32,33, 34, 35, 39, 40, 41, 42, 43, 46, 47, 49, 50) 9. Special Investigations . a. Effects of boration of water (experiment 10) b, Effects of water and air gaps in shields (experiment 48) c. Induced radiocactivity outside an air duct (experiment 27) d. Activation expected in a coolant (sodium) at various locations in a boren carbide shield (experiment 10) e. Gamma scattering around submarine bulkhead (sodium source) (special experiment) The situation at present in regard to unit shields or reactor parts of divided shields appears to be that the basic mechanisms are understood and that, except for the very simplest problems, shield mockups are essential, although the effects of small changes in design can be quite well predicted. The Lid Tank method of mockup testing is highly developed, and large amounts of data can be col- lected in relatively short times. PERIOD ENDING DECEMBER 10, 1953 13. TOWER SHIELDING FACILITY C. E. Clifford T. V. Blosser L. B. Holland Physics Division The steel structure of the Tower Shielding Facility has been erected (Fig. 13.1), and the ground structures are approximately 75% complete. The underground building for the reactor controls and for personnel will be available for occupancy by December 10; access to the control area of the building was authorized in November. The mechanical components of the reactor have been constructed and are now being assembled, and the reactor tank has been received. Construction of the crew-compartment tank is nearing completion. The last hoist is scheduled for delivery by January 1, 1954. : The optimistic date for completion of the facility is January 15, after which there will be a one-month period for critical experiments and shakedown operation. This will be followed by differential shielding measurements in the reactor and the crew-compartment water tanks. A mockup of a GE airecooled reactor shield design is scheduled for delivery by March 15 by the General Electric Company, and for testing in conjunction with this, a crew-shield mockup will be supplied by ORNL. PHOTO 11959 Fig. 13,1. Tower Shielding Facility, November 13, 1953. 125 14. SHIELDING ANALYSIS E. P. Blizard J. E. Faulkner M. K. Hullings F. H. Murray Physics Division H. E. Stern Consolidated Vultee Aircraft Corporation Much of the shielding analysis work for this quarter was reported in the Physics Division Semi- annual Report! and is not repeated here. The work reported in the Division report included two treat- ments of geometrical transformations with methods of estimating reactor leakage. |n addition, some calculational methods for obtaining neutron at- tenuation were presented. On comparison of the calculated attenuations with measured attenuations, it was possible to derive ‘‘effective removal cross sections’’ that when compared with other measure- ments showed reasonable agreement. In addition, gamma-ray absorption coefficients which have been mecsured for some elements have been used to interpolate the coefficients for other elements. The visible light from an aircraft nucleor power plant has been estimated, and the calculations will be compared with measurements made on the BSR. The measurements of visible light are made in air and are not to be confused with the Cerenkov radiation (blue glow) which is so obvious in water. An estimate of neutron albedo was obtained to supply a simple model for ground and structure scattering problems. VISIBLE LIGHT FROM A NUCLEAR POWER PLANT The results of some experiments carried out at l.os Alames on the visible light from a polonium source? have been used in obtaining an estimate of the ionizing radiotion that would have to be present T, P. Blizard et al., Phys. Div. Semiann., Prog. Rep. Sept. 10, 1953, ORNL-1630 (in press). 126 around the aircraft reactor shield to just make it visible. If the airplane were approximately 200 ft from the observer, the ionizing radiation near the shield would have to be about 5 x 10% r/hr. If the airplane were very far from the observer, say 50,000 ft, the ionizing radiation would have to be about 10% r/hr. These estimates will be compared with measurements made on the BSR. NEUTRON REFLECTION COEFFICIENT FOR WATER An estimate of the neutron reflection coefficient for water has been obtained by using a simple model with a coincident source and a nondirectional receiver located above the water surface, If iso- tropic oxygen scatter and attenuation governed by the removal cross sections are assumed, the re- flected flux F at height 5 for source strength N is NO scotter F = em———— || 48”}.72 removal The albedo appreach gives No a F = — 10752 for a re-emission distribution between isotropic and cosine, where a is the reflection coefficient. A value of a = 0.085 is thus obtained which is in fair agreement with the experimental value of 0,08. 2, G. HoHman, Ruadiation Doses in the Pajarito Accident of May 21, 1946, L A-687 (May 26, 1948). 4% REPORT NO, CF 53-9-15 CF 53-9-53 CF 53-10-62 CF 53-11-95 CF 53-12.9 ORNL-1650 CF 53-9-32 CF 53-9-102 CF 53-9-180 CF 53-10-208 CF 53-9-19 ORNL-1615 ORNL-1634 CF 53-5-239 CF 53.9.4 CF 53-9-16 CF 53-8-146 CF 53-9-161 CF 53-10-1 CF 53-10-16 CF 53-11-2 CF 53-11-45 15. LIST OF REPORTS ISSUED DURING THE QUARTER TITLE OF REPORT l, Aircroft Reactor Experiment Experimental Procedure on the ARE (Preliminary) Supplement to ARE Hazards Summary Report (ORNL-1407) ARE Fuel Recovery Report and Recommendations of Second ARE Cleaning Committee ARE Design Data Stress Anclysis of the ARE il Experimental Engineering ANP Experimental Engineering Quarterly Status Report Design Data and Proposed Test Schedule for Sodium-to- Air Radiators LITR Fluoride Fuel Test Loop A Flat Plate Heat Exchanger for Reactor System Uses 1. Critical Experiments Static Analysis of the ARE Criticciit):r Experiment Critical Experiments on Reactor Preliminary Critical Assembly for the Aircraft Reactor Experiment IV. Shielding Nuclear Dose Measurements on the Divided Shield Mockup at the Bulk Shielding Facility Lid Tank Shielding Test of the Reflector-Moderated Reactor Minimum Shield W_eighf Penalty for Air Ducts Measurement of the Fast Neutron Spectrum of the Bulk Shielding Reactor Using Nuclear Plates An Estimate of the Neutron Reflection Coefficient, Using the Concept of Removal Cross-Section Computation of Effective Removal Cross-Section Meaosured at ORNL Lid Tank : Spectrum of Gamma Rays Emitted by the Bulk Shielding Reactor The Shielding of Nuclear Radiations, Lecture | Fast Neutron Spectrum of the Bulk Sh:iellding Reactor, Part 11 AUTHOR (s) jc Ln Meem £. S5, Bettiz W. B, Cottrell F. N. Brewder Gs A- Cristy L. A. Mann F. F. Blankenship G, M. Adamson W. B. Cottrell R. L. Maxwell Jl w. Wfliker H, W. Savage H.s Jo Stumpf O, Sisman R. W, Bussard . Cq Ba Mi“S D. Callihan: Rl Cc Keen . Callithan D. Scott Hs E. Hungerford Fc No WQ‘Ison E, P, Blizard M. P. Haydon E. B. Johnson J. L. Meem Ho El Sfefn. L. S. Abbott F. C, Maienschein Te Al Love E. P. Blizard R, G, Cochron K. M. Henry DATE ISSUED 9-3-53 9-25-53 10-5-53 11-10-53 to be issued to be issued 9-2-53 9-8-53 9-25-53 10-26-53 9-1-53 10-22-53 10-28-53 5-20-53 9-30-53 9-2-53 to be issyed 9.29-53 no date fe be issued 11-2-53. te be issued 129 ANP QUARTERLY PROGRESS REPORT REPORT NO, CF 53-11-46 CF 53-11-53 CF 53-11-54 ORNL-1636 CF 53-10-117 CF 53-10-228 ORNL-1565 ORNL-1633 ORNL-1647 MM-147 MM-151 MM-157 CF 53-10-78 ORNL-1626 UA-PR-13 BMI-852 BMI-864 CF 53-8-106 CF 53-8-217 CF 53-9-98 CF 53-10-86 CF 53-11-128 ORNL-1624 130 TITLE OF REPORTY IV, Shielding (continued) Angular Distribution of Fast Neutrons in Water The Effect of Some Liquid Metal Ducts on Reactor Shislds Gamma-Ray Spectrum of the Bulk Shielding Raactor General Survey of Lid Tank Experiments 1 Through 50 V. Metallurgy Excomination of LF Pump Loep Metallographic Examination of Second Heat Exchonger from Bi-Fluid Pump Loop Scaling of Columbium in Air Fabrication of Spherical Particles Interim Report on Static Liquid Metal Corrosion Third Progress Report on the Flash Welding of Malybhdenum. Part | — Temperature Distribution During the Flashing Cycle Redrawing of Spacial High Purity Incone! Progress Repori and Final Report to Carbide and Carbon Chemicals Company Yl. Chamistry Fused Salt Compositions Determination of Zirconium by the Chloranilic Acid Method Progress Report for the Period April 1, 1953 Through June 30, 1953 Vapor Pressures of Beryllivm Fluoride end Zirconium Fluoride Potential Liquid Fuels for Nuclear Power Reactors VIl. Heat Transfer and Physical Properties Preliminary Results on Flinak Heat Transfer Preliminary Measuremants of the Density ond Viscosity of Fluoride Mixture No, 44 Heat Copacity of Two Samples of 310 Stainless Steel and of a Brazing Compound Preliminar y Measurement of the Density and Viscosity of Composition 43 (NazUFé) Heat Capocity of Fuel Composition No. 23 The Nature of the Flow ot Ordinary Fluids in @ Thermal Convection Harp AUTHOR (s) R. G. Cochran K¢ M: Henry Mo K. I'iU“ingS F. Ci Maienschein El Ln Czapek G. M. Adamson R. S Crouse Gc Mn Adqmson R. 5. Ciouse H. lnouye A, Levy W. D. Manly Rensselaer Polytechnic Institute Superior Tube Company Commonwealth Engineering Co,. of Ohie C, J. Barton 0. Menis University of Arkansas Battelle Memorial Institute Baitelle Memorial Institute H. W. Hoffinan S. Is Cohen T+ N. Jones W. D, Powers Go Ca B]OIOCk S l. Cohen Te N. Jones W. D, Powsers G. C. Blalock D. C. Hamilton F. E, Lynch L. D. Palmer DATE ISSUED to be issued to be issued 11-8-53 to be issued 10-6-53 10-27-53 ?-1-53 11-12-53 to be issued 9-30-53 10-7-53 11-9-53 10-9-53 10-28-53 7-30-53 7-13-53 9-8-53 8-18-53 8-31-53 92-18-53 10-14-53 11-23-53 to be issued REPORT NO. CF 53-9-8 CF 53-11-147 ORNL-1609 ORNL-1632 PERIOD ENDING DECEMBER 10, 1953 TITLE OF REPORT Yitl. Miscellaneocus The Small Hydrogen Moderated Reactor Directory of Active ANP Reseorch Projects at ORNL Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending September 10, 1953 ORNL Contrel Computer AUTHOR (s) DATE (SSUED C. B, Mitis G-2-53 W. B. Cottrell 11-27-53 W. Bo Come“ 10‘15"53 Je J. Stone to be issved 16. DIRECTORY OF ACTIVE ANP RESEARCH PROJECTS AT ORNL NOVEMBER 27, 1953 l. REACTOR THEORY AND COMPONENT DESIGN A, B. E. Aircraft Reactor Design L 2, 3. 4, Reflector-Moderated Reactor Studies Fluid-Fuel Flow Studies Gamma Heating of the Moderator Design Consultants ARE Design 1. 2, 3. 4, Building Electrical Power Circuits Fluid Circuit Design Fuel Recovery Equipment ARE Installation 1. 2 3. 4, 5. 6. Plumbing Electrical System Controls instrumentation Expeditors Coordinators ARE Operations 1. 2. 3. Staff Day Shift MNight Shif? ARE Control Studies .‘. High-Temperature Fission Chamber 9704-1 9704-1 9704-1 9704-1 1000 1000 7503 9201-3 7503 7503 7503 7503 7503 7503 7503 7503 7503 4500 Fraas, L.aVerne, Bussard Stumpt Bussard, Wilner Wislicenus, JHU; Maines; Chambers, U.T. Browning Walker Cristy Estobrook Webster et ol. Packard et =i, Epler, Mann Affel Perrin, West Wischhusen, Watts Bettis, Meem, Cristy, Mann Whitman, Cottrell, Grindell, Croley, Schafer, Williams, Southern Affel, Perrin, Cobb, Leslie, Allen, Reid, Morrison, Huntley, Gregory, Addisen Hanauer 131 ANP QUARTERLY PROGRESS REPORT 2, Control System Design 3. Control Studies on Simulator F. Reactor Physics 1. Anclysis of Critical Experiments 2. Kinetics of Circulating=-Fuel Reactors 3. Computation Techniques for ARE-Type Reactors 4. Statics of Reflector-Moderated Reoctors 5. Machine Computations 6. Interpretation of Critical Experiments G, Critical Experiments 1. Reflector-Modzrated Reactor Critical Assembly 2. Preliminary Assembly of Supercritical-Water Reactor (Prott & Whitney) 3., Preliminary Assembly of AC-1 (General Electric ANPP) H. Besaring, Secl, and Pump Development 1. Mechanical Pumps for High-Temperature Fluids 2, High-Temperature Seals for Rotary Shafts 3. Pump for Hydroxide Systewms 4, Seals for Hydroxide Systems 5. Bearings for High-Temperature Application 6. Canned-Rotor Pump Development 7. Materials Compatibility in Molten Salts at High Temperatures ' . Valve Development 1. Valves for High-Temperature Fluid Systems 2. VYalves for FHydroxide Systems J. Heat Exchanger and Radiator Development 1. High-Temperature Fluid-to-Air Heat Exchange Test 2. Fluoride-torLiquid Metal Heat Exchange Test 3, Heat Exchanger Fabrication 4, Boeing Turbojet with Sodium Radiator 5. Radiator and Heat Exchanger Design Studies 6. FulleScale Test Facilities K. Fluid Dynamics of Liiquid Fuel Reactors 1. Pump-Core Shell Component Test .. Instrumentafion 1. Pressure-Measuring Devices for High-Temperature Fluid Systems ]Buflelle Mewmorial Institute, 132 4500 4500 9704-1 9704-1 9704-1 9704-1 4500 4500 9213 9213 2213 9201-3 9201-3 BMI! BMI 9201-3 9201-3 9201-3 9201-3 BMI 9201-3 9201-3 9201-3 9201-3 9201-3 9201-3 9201-3 9201-3 Epler, Ruble, Mann, Green, Honauer QOakes, Mann, Green Mills Ergen Bengstan, Ergen LaVerne, Burtnette, Spencer Charpie, Coveyou, Sangren, Gerberich, Given, Atta Prohammer Callihan, Williams, Scott, Zimmerman, Lynn, Noakes (P &W) Callihan et al, Callihan et al, McDonald, Cobb, Huntley,Grindell McDonald, Tunnell, Smith, Mason Dayton Simons, Allen Stair, Tunnell, Smith, Mason McDonald, Stair, MacPherson, Grindeil Stair, Tunnell, Smith, Masen McDonald Simons, Allen et al. MacPherson, Salmon, Frasos, Stumpf Salmon, Fraas, MacPherson, Bussard Petersen, Fiaas Fraas, Coughten Fraas, Stumpf, Bailey, U.T. Coughlen Bussard, Gallagher MecDonald, Smith 2. Flow-Measuring Devices for High-Temperature 3. 4. Fluid Systems High-Temperature Strain Gage Highe Temperature Calibration Gage Il. MATERIALS RESEARCH A. Phase Equilibrium Research 1. 2, 3‘ 4, 3. 6. 7 8. 9. 10. Thermal Analysis of Fluoride Systems Therma!l Analysis of Chloride Systems Phase Equilibria Among Fluorides by Quenching Equilibria in BeF -Bearing Systems by Quenching Differential Thermal Analysis of Fluorides Differential Thermal Analysis of BeF -Bearing Systems Equilibria Among Fluorides and Chiorides by Filtration X-Roy Diffraction Studies of Complex Fluorides Phase Equilibria in Hydride-Hydroxide Systems Phase Equilibria Consultant B. Chemistry of High-Temperoi:fure Liquids 1 - 2, 3. 10. 1L 12. 13. 15. Equilibrium Pressures iin Hydroxide-Metal Systems Equilibrium Pressures in Hydroxide-Metal Systems Equilibrium Pressure in Hydride-Halide Systems Chemical Reactions in Fused Salts Solutions of Metals in Fused Salts Thermagalvanic Potentials in Fused Systems Decomposition Potentials in Fused Systems Spectrophotometric Investigations of Fused Salts Preparation, Phase Behavior, and Reactions of UF, Rates of Reduction in Fused Salts Solubility of Uronium Compounds in Hydride- Mydroxide Systems Frae Energy of Fluorides Preporation of Pure Hydroxides Preparation of Complex Fluorides Chemistry Consultants C. Fluoride Production and Handling 1. 2, PiloteScale Production of Fluorides | aboratory-Scole Fluoride Production 2Ba|dwin-l..ima Hamilton Corp. IMetal Hydrides, Inc. PERIOD ENDING DECEMBER 10, 1953 9201-3 BLH? Cornell 9733.2 9733.2 9733-2 Mound 9733-2 Mound 9733-2 4500 9766 MHI® U. Ark. 9766 Mound 9733-3 4500 9733-3 97333 9732-3 9733-2 97332 aml! 9201-3 9733.3 9733-3 9928 97332 McDonald, Smith Insley, Barton, Bratcher, Truitt Borton, Boyer insley, Moore, Metcalf, Truitt, Thoma, White Heik, Orban, Joy et al. Bolomey Orban et of. Barton, Sheil Bredig Agron, Thoma Banus et al. McVYay, U. Ala, Smothers et al. Knox, Kertesz Otto, Orban Overholser, Redman, Weaver Bredig, Johnson, Bronstein Cuneo Topol, Overholser Friedmon, Blankenship Barton, Truitt, Bloed, Watson, Blankenship Blood, Watson, Blankenship Simons et al. Maan, Petersen, Cisar Ketchen, Overholser Sturm, Overholser Gibb, Tufts U. Hill, Duke U, Carter, Carter L ab. Nessle, Blakely, Eorgan, Croft, Didloke Blood, Watson, Blankenship, Boody 133 ANP QUARTERLY PROGRESS REPORT G, H. 134 3. 4, 5. 5. 7. 8. Preparation of Materials for Radiation Damaoge ARE Fuel Carrier Production ARE Fuel Concentrate Production ARE Filling Operation Sampling Techniques Sampling Techniques ARE Fuel Reprocessing 1. 2, 3. 4, 5. Fuel Recovery Process Development Fuel Recovery Design Studies Unit Operotions Studiss Pilot Plant Qperations Fuel Recovery Process Studies Corrosion by Liquid Metals 1. 2 3- = 4 Sl 6. 7. 8. 9. 10. 11, 12. Static Corrosion Tests Dynamic Corrosion Research in Convection Loops Effect of Crystal Orientation on Corrosion Effect of Carbides Dynamic Corrosion by Liquid Lead Diffusion of Molten Media into Solid Metals Structures of Liquids by Diffraction Techniques Protective Coatings for Corrosion Resistance Corrosion Inhibitor Studies Handling of Liquid Metal Samples Compatibility of Cermets in Liguid Metals Mechanism of Isotherma! Mass Transfer Corrosion by Fluerides — 2, 3. 4, S 6. Static Corrosion of Metals Static Corrosion of Metals Corrosion in Small-Scale Dynamic Systems Dynamic Corrosion Tests in Thermal-Convection Loops Corrosion of Nonmetallic Materials Forced-Convection Corresion Loop Corrosion by Hydroxides 1. 2, 3. 4, Static Corrasion of Metals and Alloys Physical Chemisiry of Hydroxide Corrosion Static and Dynamic Corrosion Static and Dynamic Corrosion by Hydroxides Physical Properties of Materials 1. 2. 3. 4. 5. 6. 7. 8. Density of Liquids Yiscosity of Liquids Therma! Conductivity of Selids Thermal Conductivity of Liquids Specific Heat of Solids and Liguids Electrical Conductivity of Fluorides Surface Tension of Fluorides Viscosity of Fused 5alt Systems 97332 9201-3 9212 7503 9201.3 9766 4500 4500 4505 3019 3505 7503 2000 9201-3 2000 2000 2000 9201-3 2000 2000 2000 9201.3 9201-3 2000 2000 2000 9766 2000 9201.3 9766 9201.3 2000 2000 9766 BMI! 9204-1 9204-1 9204-1 9204-1 9204-1 9204-1 9204-1 9766 Blankenship, Boody Messle, Eorgan, Blond, Thoma Nessle, Forgan MNessle, Forgan Mann, Blakely Kertesz, Meadows, Didlcke Ferguson, Cathers Goeller, Browder, Ruch Eister, Nurmi, Kackenmaster Jackson Lewis, Matherne Messlz, Blaksly, Croft, Eorgen Yreeland, Hoffman Adomson Smith, Cathcart, Bridges Vreslond, Hoffman Cathcort Adamson Smith, Cathcart Smith VYreelond, Hoffman Adamson Ketchen Yreeland, Hoffman Cathcart Yreeland, Hoffman Kertesz, Buttram, Smith, Meadows ¥reeland, Trotter, Nicholson, Hoffman Adamson Kertesz, Buttram, Meadows, Smith Sc‘.‘ll maon Vreeland, Smith, Hoffman, Cathcart Smith, Steidlitz, Boston Kertesz, Buttram, Smith Simons et al. Joines, Cohen Jones, Cohen Powers, Burnett Cloiborne, Fowers, Burnett Powers, Blalock Greene Cohen, Jones Kertesz, Knox 9. Vapor Pressure of Fluoride Systems 10, Vapor Pressure of BeF ,-Bearing Systems 11. Vapor Pressure of Fluid Systems i» Heat Transfer 1. Heat Transfer Coefficients of Flucride and Hydroxide Systems 2. Heat and Momentum Trapsfer in Convection Loops 3. Heat Transfer in Circulating-Fuel Systems 4. Forced Convection in Annuli 5. Transient Boiling Research 6. Fluid-Flow Phenomena J. Metals Fabrication Methods 1. Welding and Brazing Techniques for ARE 2. Molybdenum Welding Research 3. Molybdenum Welding Research 4, Resistance Welding for Molybdenum and Clad Metals | 5. Welds in the Presence of Various Corrosion Media 6. Nondestructive Testing of Tube-to-Header Welds 7. Basic Evaluation of Welds Metal Deposits in Thick Plates 8., Development of High-Temperature Brazing Alloys 9. Evaluation of the High-Temperature Brazing Alloys 10. Application of Resistance-Welding Methods to ANP Materials K. New Metals Development 1. Molybdenum and Columbium Alley Studies 2. Heat Treatment of Alloys for Special Projects 3. Alloy Development of New Container Materials L. Solid Fuel Element Fabrication 1. Solid Fuel Element Fabrication 2. Diffusion-Corrosion in Sclid Fuel Elements 3. Determination of the Engineering Properties of Solid Fuel Elements 4. Inspection Methods for Fuel Elements M. Ceramics and Metals Ceramics 1. Metal Cladding for Beryllivm Oxide 2. Hot Pressing of Bearing Materials 3. High-Temperature Firing of Uranium Oxide to Produce Selective Powder Sizes 4, Development of Cermets for Reactor Components 3. Ceramic Cooatings for ANP Radiators 4Massachuseh‘s Institute of Technology. SRensselaer Pol ytechnic Institute. PERIOD ENDING DECEMBER 10, 1953 9733-2 Mound BMI] 9204-1 9204-1 9204-1 9204-1 9204-1 9204-1 2000 BMmi! miT4 RPI® 2000 2000 2000 2000 2000 2000 2000 2000 2000 2000 2000 2000 2000 Gerity MiChn 2000 2000 9766 9766 Moore Orban et of. Simons et ol. Hoffman, Lones Hamilten, Palmer, Lynch Poppendiek, Winn, Palmer Bradfute Rosenthal Palmer, Winn, Poppendiek Manly, Housley, Slaughter, Patrigrca Russell Wulff Nippes, Savage Vreeland, Slaughter Hoffman, Patriarca Slaughter, Patriarca Slaughter, Gray, Patriarca Patriarca, Slaughter Slaughter, Potriarca Slaughter, Patriarca Incuye Bomar, Coobs inouye Bomar, Coobs Bomar, Coobs Bomar, Coobs, Woods, Oliver, Douglas Bomar Graaf Bomar, Coobs Bomar, Coobs Johnson, Shevlin, Taylor, Hamner White, Griffin 135 ANP QUARTERLY PROGRESS REPORT 6., Ceromic Valve Parts for Liquid Metals and Fluorides 7. Ceramic Reflactor 8. Ceramic Coatings for Shielding 9. Fabrication of Be( Shapes 10, Cosrosion of Ceramics 11. High-Density Graphite N. Strength of Materials 1. Creep Tests in Fluoride Fuels 2, Creep and Siress Rupture Tests of Metals in Vacvum Gaseous Environments and in Fluid Media 3. High-Temperature Cyclic Tensile Tests 4, Tube-Burst Tests 5. Relaxation Tests of Reactor Materials 6. Evaluation Testing for Other Development Groups 7. Fatigue Testing of Reactor Materials 0. Radiation Damage 1. Fused Salt Fuels lrradiation in LITR 2. Fused Salt Fuels lrradiotion in MTR 3. LITR Dynamic Corrosion L.oop 4. Fluoride Fuel Loops for LITR and MTR 5. Engineering Properties of lrrodiated Metals 6. Meutron Spectrum of LITR 7. Creep of Metals and Stress Corrosion in ORNL Graphite Reactor and LITR 8. Creep of Meials in MTR 9. Thermal Conductivity of Metals in ORNL Graphite Reactor and LITR 10. Remote Metallography for GE-ANP, Pratt & Whitney, and In-Pile Corrosion Tests 11. Radiation Damoge Consultants P. Materials Analysis ond Inspection Methods 136 1. Determination of the Oxidotion States of Metallic Corrosion Products 2. Determination of Trivalent Uranium in Fuels 3. Anclysis of Reactor Fuels and Coolants 4, Determination of Reducing Power of Fuels and Coolants 5. Determination of Oxides in Reactor Fuels 6. High-Tempercture Mass Spectrometry 7. X-Ray Study of Complex Fluorides 9766 9766 9766 U. Ala, 9766 9766 2000 9766 9201-3 2000 2000 2000 2000 2000 2000 3550 3005 3550 3550 3550 3005 3550 3005 3025 3001 3025 3001 3005 3025 3001 9733-4 9733-4 9733-4 9733-4 9733-4 9735 3500 9766 Shevlin, Johnson, Taylor Johnsen, White White, Shevlin Handwerk Doney Curtis, Kertesz, Buttram Vreeland, Hoffman Doney, Corey Adomson Oliver, Douglas Otliver, Woods Olivar, Dougles Oliver et al. Oliver, Woods, Reber Qliver, Reber Keilholtz, Morgan Robertson, Webster, Robinson, Willis, Browning Keilholtz et al. Keilholtz et ol. Sisman, Baumann, Carroll, Brundage, Blacksher, Parkinson, Ellis, Olson, Morgan Sisman, Parkinson, Baumann Carroll Trice, Sismon Wilson, Zukas Davis Wilson, Hinklz Cohen, Templeton Weeks Feldman, Richt Acton, Schwartz Ruoark, U. Ala, Smith, Cornell U. White, Manning White, Ross White, Ross, Roemer, Manning White, Manning White Baldock, Sites Agron, Johnson, Bredig Thoma ' 8. 9. 10. 11. 12, 13. 14. 15. Petrographic Examination of Fuels Chemical Methods of Fluid Handling Metallographic Examinotion ldentification of Corrosion Products from ‘Dynamic Loops Assembly and Interpretation of Corresion Data from Dynaomic Loop Tests Metallurgical Examination of Engineering Ports identification of Corrosion Products Experimental Material Distribution lil. SHIELDING RESEARCH A, Cross-Section Measurements Te Effective Removal Cross Sections B, Shielding Measurements 1. 2, 3. 4. 5. 6. 7. 8, ?. Slont Penetration of Neutrons Air Duct Tests — Fast Neutrons Unit Shield Optimization Mockup Tests for Shield Design Divided-Shield Mockup Tests (GE-ANPP) Duct Tests (GE-ANPP) Gomma Rays from Fission FulleSize Mockup Tests of Aircraft Shields Design Parameter Studies {Divided Shield} C, Shielding Instruments 1. 2. 3. 4. 5. 6. Gammoa-Ray Spectroscopy and Calculations Short-Period Fission-Product Gamma-Ray Spectra Fast-pleutron Dosimeter He® Counter for Neutron Spectrometry Neutron Spectroscopy ~ Proportional Counter Neutron Spectroscopy — Photographic Plotes D. Shielding Analysis 1. 2, 3. 4. Development of New Calculational Methods Application of Stochastic Methods Calculation of Neutron Attenuations Calculations of Gammua-Ray Attenuation E. Experiments on Shielding Facility Reactors 1. 2, Miscellaneous Reactor Experiments { Temperature Coelficient, Xenon Poisoning, etc.) Bulk Shielding Reactor Power Colibration St ower Shielding Facility, yNuclaur Development Associates, Ine. PERIOD ENDING DECEMBER 10, 1953 9766 9201-3 2000 9733-3 2000 9201-3 2000 9201.3 2000 9766 3001 3001 3001 3001 3001 3010 3010 3010 TSFS TSF® 3010 3010 3001 4500 3010 3010 3006 4500 4500 NDA7 NDA’ 3010 3010 McVYay, White Mann, Blakely Gray, Krouse, Roeche Blankenship Smith, Borie, Dyer Adamson, Blankenship Smith, Vreeland Adamson, Gray, Vreeland Y akel Kertesz, Meadows, Didlake Storrs, Trubey, Watson Storrs, Chapman Storrs, Miller Storrs and crew Storrs and crew Cochran and crew Cochran, Fiynn, Hull, Hungerford Maienschein, LLove, Cochran, Henry Clifford, Blosser, Hollond, Watson Clifford, Blosser, Holland, Watson Maienschein, Love Maienschein Campbell, Love Blosser, Hurst Cochran, Henry Cochran, Henry MHaydon, Johnson, Cochran Murray Favlkner, Zerby Goldstein, Aronson, Certaine Goldstein, Wilkins Cochran, Meem, Cole, Johnson, Flynn Johnson, Cochran 137 CHART OF THE TECHNI{CAL ORGANIZATION OF THE AIRCRAFT NUCLEAR PROPULSION PROJECT THE OAK RIDGE NATIONAL LABORATORY AT DECEMBER 1, 1953 ANP PROJECT DIRECTOR SOVIIND SRR — “*Thin material comtaing informotion gHwcting the of tha Unitad Statan within the mucning of the aupi rla 18, U.S.C., Seca, 793 ond 794, the mansmission or revelotion natianel dufm o ianage laws. of which in any maner o on unauthorized person is prohibited by law.** R. C. BRIANT RD D, HILYER,* SEC. ANP ASSISTANT DIRECTOR A4 MILLER ANP TECHNICAL ASSISTANT ASSISTANT TO DIRECTOR W, B, COTTRELL ANP L. APA. fiAogKmu e m: A. W. SAVOLAINEN ANP : i 3EC J. M. CISAR® ANP LIBRARY P. HARMAN, SEC. AN:‘. BYILDING §704-1 ' M. CARDWELL RD E. WEBSTER RD STAPR ASSISTART FOR SHEELDING STAFF ASSISTANT FOR PRYSICS A ) MiLLER ANP STAFF ASSISTANT FOR DESIGN ARE PROJECT ENGINEER STAFF ASSISTANT FOR METALLURGY STAFF ASSISTANY FOR CHENISTRY D. HILYER,* SEC, ANP E. P. BLIZARD P W, K. ERGEN ANP A. P. FRAAS ANP E. 5. BETTIS* ANP R. . BRIANT RO W. D. MANLY M W. R. GRIMES MC SHIELDING RESEARCH REACTOR PHYSICS RADIATION DAMAGE GENERAL DESIGN ARE CPERATIONS EXPERIMENTAL ENGINEERING HEAT TRANSFER AND PHYSICAL METALLURGY CHEMISTRY BUILDING 4500 BUILDING 9704-1 BUILDING 3025 BUILDING 9704-1 BUILDING 7503 BUILDING $201-3 PROPERTIES RESEARCH BUILDING 2000 BUILDING 97333 £. P. BLIZARD P ¥, K, ERGEN ANP D. 5. BILLINGTON* 55 A, P. FRAAS ANP E. 5. BETTIS ANP H, W, SAVAGE ANP BUILDING 5204-1 W. D. MANLY M W. R, GRIMES MC 4. B. TRICE 55 J. L. MEEM ANP H. F. FOPPENDIEK REE L. 5. ABBOT P 1, BENGSTON ANP C. 5. BURTNETTE USAF G. M. ADAMSON M E. S. BOMAR M P. A. AGRON c 1. E. FAULKNER P €. 8. MLLs ANP B O N = R, . BUSSARD ANP R. G. AFFEL ANP J. P. BLAKELY He 1. 0. BRADFUTE REE C. R, BOSTON " C. 1 BARTON e . X. BULLINGS P D. HILYER,* SEC. ANP " 1, 6. GALLAGHZR ALT J. W, ALLEN ANP C. P. COUGHLEN ANP 5. 1. COHEN REE W, H, BRIDGES M F. F. BLANKENSHIP MC , T, BERESOVSKI PW G, F. K. MURRAY p ¥, E. BROWNING by 8. L. GREENSTREET ANP . G. COBB ANP J, E. EORGAN MO N. D. GREENE REE 4. ¥. CATHCART M C. M. BLOOD MC H, E. STERN cy COMPUTERS W E. BRUNDAGE s M. E. LAYERNE ANP W, B, COTTRELL" ANP W. C. GEORGE Y-12 0. €. HAMILTON REE J. M, CISAR* ANP R. A, BOLOMEY MC R, RICKMAN, SEC, P A. FORBES ANP R, M. CARROLL bl W. L. SCOTT ANP G. A. CRISTY EM R.HELTON ANP H. W, HOF FMAN REE J. H. COOBS M F.P. BOODY MC M. TSAGARIS ANP W W, DAYIS 38 R, L. SPENCER USAF J. Y. ESTABROOK ANP E. M. LEES Anp F. £, LYNCH REE 0, A, DOUGLAS M 5. A, BOYER MC CONSULTANT M ). FELDMAN o . J. STUMPF ANP A. G, GRINDELL ANP R, E. MACPHERSON ANP L. D. PALMER REE E. E. HOFFMAN M L. M. BRATCHER MC H. A. BETHE, CORNELL UNIVERSITY R W. HALL NACA B. M. WILNER AGC W, R, HUNTLEY ARF L. A MANK ANP . D. POWERS REE H. INQUYE o D. R, CUNED MC - A . N E. HINKLE se D. HILYER,* $£C. ANP J. K. LESLIE ANP R. N. MASCN EM M., ROSENTHAL REE R.B. OLIVER M H. A. FRIEDMAN MC . W, KEH HOLTZ s E. R. MANN® i #. B. MCDONALD ANP T. SUTTON, SEC. REE P, PATRIARCA M E, E. KETCHEN W CONTRACTOR CRITICAL EXPERIMENTS 1 6. MORGAN s CONSULTANTS E. L. MORRISON USAF G. J. NESSL MC K. W. REBER M R.P. METCALF MC NUCLEAR DEVELOPMENT ASSOCIATES, INC. BUILDING 9212 Mo T, MORCAN s A. H. FOX, UNION COLLEGE E. B. PERRIN AP o 'F‘- N ::; TECHNICIANS G. M. SLAUGHTER M R. E. E MC H. GOLDSTEIN A D. CALLIMAN* p A, S. OLSON 5 R. L., MAXWELL, UNIVERSITY OF TENNESSEE A. L. SOUTHERN ANP g €. 6. BLALOCK REE G. P. SMITH M L. G. OVERHOLSER MC H. 6o . D, L 5 G B WISLICENOS, JORNS HORKINS UNIVERS:TY H. L. WATTS REE D.F. ANP R. M. BURNETT REE . €., STEIDLITZ M J. D. REDMAN MC - ARONSON V. 6. HARNESS* B W, PARKINSON 5 C. F. WEST ANP P. G, SEXTON Y-12 5. J. CLAIBORNE REE 0, C, YREELAND M R. J. SHEIL MC J. J LYNN P A E. RICHT 55 G. D. WHITMAN ANP P\ G, SMITH ANP T. N. JONES REE J. W. WOO0DS M 8. J. STURM MC J. H. MARABLE" P | H. E. ROBERTSON 55 E. WISCHHUSEN EM W, C. TUNNELL ANP J. LONES REE 1. THOMAS, SEC. M R, E, THOMA, JR, MC 1. W. NOAKS PW " T' ROBINSON s P. HARMAN," SEC. ANP D. R. WARD ANP G. M. WINN REE L, E. TOPOL MC E. R. ROHRER* p O SISMAN s D. ALEXANDER, SEC. ANP TECHNICIANS JACK TRUITY NC BULK SHIELDING FACILITY . TECHNICIANS D. STOREY, REC. CL. ANP . M. WATSON e D, SCOTT, JR, ANP W, J. STURM 55 b PR 'ST N 6. D, BRADY M BUILDING 3010 D. V. P. WILLIAMS P L. C TEMPLETON = 1. 5. ADDISON ANP . . STEND, 1 T. EAST M C. F. WEAVER Me R. G. COCHRAN P £. L. ZIMMERMAN P P. M. 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THOMAS M No Vo SMITH Me 5. B ATTA" " F. M. BLACKSHER 5 R A m;ch ANP L J. R, JOHNSON M CONSULTANTS TECHNICIANS R. R. COVEYOU* ANP B L. JOHNSON ANP e ; L‘:'#g“ " METALLOGRARHY J. M. CARTER H. JARVIS Ic C. L. GERBERICH* WP CONSULTANTS 5. W. KINGSLEY ANP " A 0B85 * SEC " R, J. GRAY" u T. R. P. GIBE, WILLIAMS COLLEGE D. J. KIRBY P N, D, GIVEN* MP J. E. GOLDMAN, REACTOR CU“STRDI- D. £, MCCARTY ANP . ‘ . "R, 5. CROUSE* M D. G. HILL, DUKE UNIYERSITY R. M. SIMMONS P F. G. PROHAWMER* ANP e BUHLDING 4500 G. E. MILLS ANF TECHNICIAN : , H. INSLEY, BUREAU OF STANDARDS G. STOUT K W. C. SANGRENY MP CARNEGIE INSTITUTE OF TECHNOLOGY E. P, EPLER* ic . A, REDDEN ANP LA GRIFF:«IC' < ¥ g iggfgv, . T. N MCVAY, UNIVERSITY OF ALABAMA H. WEAVER P D. F. WEEKES, TEXAS A & M COLLEGE D. £, TIDWELL Abie - A M + M ; E SS;EN :g R.G MLEY ANP CONSULTANTS TECHNICIANS CONTRACTORS 1. C. GUNDLACH® ic CONSULTANTS T, N, MCYAY, UNIVERSITY OF ALABAMA M. D. ALLEN® M BATTELLE MEMORIAL INSTITUTE _ S H. HANAUER® i T. 5. 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GRANT, INDICATED ON THE CHART. EACH GROUP, HOWEVER, 15 ALSQ RESPONSIBLE TO IT$ DIVISION DIRECTOR F.L. CULLER® et MASSACHUSETTS INSTITUTE OF TECHNOLOGY MOIJJN-:) ;S:gié‘E’ORY FOR THE DETAILED PROGRESS OF ITS RESEARCH AND FOR ADMINISTRATIVE MATTERS, F. R. BRUCE cr J. L. GREGG, CORNELL UNIVERSITY "AND OTHERS E.F. NIPPES, F. N. BROWDER cT ¥ AS LD TANK THE KEY TO THE ABBREVIATIONS USED IS GIVEN BELOW. IT SHOULD BE NOTED THAT SEVERAL OUTSIDE G. I CATHERS cr RENSSELAER POLYTECHNIC INSTITUTE R T KANS ILDING W. F. SAVAGE e RGANIZAT| VE PERSONNEL PARTICIPATING IN THE ORNL ~ANP PROGRAM . AND OTHERS C. L. STORRS GE ORGANIZATIONS HAVE FERSONKEL PARTIC ORNL RENSSELAER POLYTECHNIC INSTITUTE E. C. WRIGHT, UNIVERSITY OF ALABAMA G. T. CHAPMAN P AC ANALYTICAL CHEMISTRY DIVISION — ORNL R £ AGC AEROJET-GENERAL CORFORATIGN CONTRACTORS - ALC AMERICAN LOCOMOTIVE COMPANY BALDWIN-LIMA-HAMILTON CORFORATION ANALYTICAL CHEMISTRY TECHNICIANS F. 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UNIVERSITY OF MICHIGAN B FHYSICS DIVISION — ORNL AND OTHERS ’ PW PRATT AND WHITNEY AIRCRAFT DIVISION — UAC SPECTROGRAPHIC ANALYSIS R KD RESEARCH DIRECTOR'S DEPARTMENT — ORNL B pe Y TECHNIC INSTITUTE BUILDING 9734 REE REACTOR EXPERIMENTAL ENGINEERING DIVISION - ORML AND OTHERS J.R. MCNALLY sl ] STABLE ISOTOPES DIISION — DRNL SUPERIOR TUBE COMPANY J. A NORRIS* H] . . AND OTHERS 55 SOLiD STATE DIVISION - CRNL HUGH COOFER USAF UNITED STATES AIR FORCE MASS SPECTROME TRY ¥-i2 CARBILE AND CARBOK CHEMICALS COMPANY {Y-12 SITE) BUILDING 9735 PART THHE C. R. BALDOCK* ] 1. R. SITES H] 138