£ brary wi ni, cume do ,. e i , doumvany &g h G Mao in name wi >W?§F?PFW??P?OP?POFP?T?PWNPWOfif o0z MASTOMIrYQoEwMIPAMSETOAO> MU T E e . H. Jordan & . W. Kellholtzfr . Adamson . Affel . Baldock . Barton . Callihan . Cardwell Cathcart . Center (K-25) . Chapman . Charpie . Clewett . Clifford . Cottrell . Cowen omer Crouse . Culler . Emlet (K-25) . Ferguson . Fraas Frye . Furgerson . Grimes . Hoffman . Hollaender . S. Householder T. Howe W. Johnson P. Keim . T Ke[ley . Ker'resz I;j:son M. Klnj‘ A, . E. E!L erne ORNL-1896 Progress INTERNAL DISTRIBUTION 46, R. S. Livingston 47. R. N, Lyon 48. F. C. Maienschein 49. W. D. Manly 50. L. A. Mann 51, W. B. McDonald 52. F.W.McQuilken g 53. A. J. Miller & 54, K. 55. E. 56, J.F 57. G. 58. R. 59. P. 60, A. 61. H.4& 62, P 63 '-» W. Savage ? A. W. Savolainen E. D, Shipley _ . O. Sisman %67. G. P. Smith "88. A. H. Snell 6 R. . Strough 70.‘%% D. Susano Whltmun R . Wigner {cons . Williams . Wilson 83. C. E. Winters 84, C. D, Zerby 85.94. X-10 Document Reference ki 95-114, Laboratory Records Depar'rm- 115. Laboratory Records, ORNL R-:. 116-118. Central Research Library fln‘uo |6 A EXTERNAL DISTRIBUTION 119. AFDRD Jones § 120. AFDRQ . 4, 121, AFSWC W 122, Aircraft Lab WADC (WCLS)# 4123, Argonne National Laboratéfi%y 04 ATIC ]25—1' 12‘“ BAGR - WADC 129\.‘ah‘e||e Memorial In&'afute 141 142, 143. ofd 144. i boratory — WADC (WCLE) 145-148. | 149. 150. 151, Knollgg'A\‘omlcfif_‘E ower Laboratory 152, Loc heed Bugbank 153. Loc .; ... r, | : 172, SAM }73-174. Technical Information Serv , Oak Ridge Operations Office ,.[}175-177 WADC - Library E 178. WAPD — Bettis Plant 179. Wright Aero ORNL-528 ORNL-629 ORNL-768 ORNL-858 ORNL-919 ANP-60 - ANP-65 ORNL-1154 ORNL-1170 ORNL-1227 ORNL-1294 ORNL-1375 ORNL-1439 - ORNL-1515 ORNL-1556 ORNL-1609 ORNL-1649 ORNL-1692 ORNL-1729 ORNL-1771 ORNL-1816 ORNL-1864 Reports previously issued in this series are as follows: Period Ending November 30, 1949 Period Ending Fébrudry. 72'8, 1950 Period Ending May 31, 1950 Period Ending .Augus-fr- 31, 1950 Period Ending December 10, ]950 Period Ending March 10, 1951 Period Ending Junelr'lfl, 1951 Period Ending September 10, 1951 Period Ending December 10, 1951 Period Ending March 10, 1952 Period Ending June 10, 1952 | Period Ending September 10, 1952 Period Ending December 10, 1952 Period Ending March 10, 1953 Period Ending June 10, 1953 Period Ending September 10, 1953 Period Ending December 10, 1953 Period Ending March 10, 1954 Period Ending June 10, 1954 Period Ending September '170, 1954 Period Ending December 10, 1954 Period Ending March 10, 1955 1. iy ) FOREWORD Th:s quarterly progress report of the Aircraff Nuclear Propulsron Project at ORNL records the technical progress of the research on c:rculchng-fuel reactors and all other ANP research at the Laboratory under its Contract W- -7405-eng-26. The report is divided into three major parts: [. Reactor Theory, Componen’r Developmenf and Construction, I, Materials Research, and IHl. Shielding Research The ANP Project is comprlsed of about 475 technical and screnhflc personnel engaged in many phases of research directed toward the achievement of nuclear propulsion of aircraft. A considerable porhon of this reseqrch is performed in support of the work of other organizations participating in the national ANP effor’r. However, the bulk of the ANP research at ORNL is directed toward ’rhe development of a circulating-fuel type of reactor. The design; construction, and 6pércrtion of the Aircraft Reactor Test (ART), with the coopera- tion of the Pratt & V‘Whitney Airc':rafi-D'ivision, are the specific objectives of the project. The ART is to be a power plant system that will include a 60-Mw circulating-fuel reflector-moderated reactor and adequate means for heat disposcl.' Opercrfion of the sysferfi will be for the purpose of determining the feasibility, and the problems associated with the design, construction, and operation, of a high-power, circulating-fuel, reflector-moderated aircraft reactor system. T ST T T P T SRR T SUMM/\I EY ............................................................................................................................................................ PART I. REACTOR THEORY, COMPONENT DEVELOPMENT, AND CONSTRUCTION 1. REFLECTOR-MODERATED REACTOR oo e eeer e | Aircraft Reactor Tesf Design e ettt et et e b et s eee et ens ART Control o, et byt eeenres rerrre s et eareabaeaaenne et naes Control Principles ..ottt Operating Procedure ........ OO OO OSSOSO PRO P PRTUOROSIN e Reactor Physics ..., e ettt e e e - Probable Effect of Replacing Inconel by Columbium in the ART Core Shells................o ART Temperature Coefficient ......co.iviieeiiricirieis ettt sssb st Reactivity Effect of a Heat Exchanger Leak ..., Burnup and Gamma-Ray Heating of Control Rod ... 2. EXPERIMENTAL REACTOR ENGINEERING ... In-Pite Loop Component Development ..o F lUX MEOSUREIMENES o iiiiiiiiiieeiieit et eeiorte s teiseeeeasseeeesanraeeeestsarteas e b e et s sbab e e esermbaas s abatsseansm e ee s s baaaeeaabiaees Fission-Gas Holdup.....ccooocovenninini et — et e e e bt e nbe it e e nreens e et ete e et e b e e e neaas Bench Test........... e e e e e ereeeeba et e e s e st et sae e renee ettt Horizontal-Shaft Sump Pump ...ccoveieiie e, eetteere it i bo et e ersenenaeataesanats IS rUMENTATION ..ottt ettt bbb e e " Assembly of MTR In-Pile Loop Nou Tt seesesssssomsenes s srons Development and Operation of Forced-Circulation Corrosion and Mass Transfer Tests ................. Operation of Fused Salt-Inconel Loops ... Sodium in MUBIMETal LoopS oot e i Pump Development ..o e et Water Performance Tests ... s Bearing and Seal Tests ... i s " 'Mechanical Shakedown Tests......c.ccocvrnnnne. s bttt fShorf—C:rcunf F’ump Test Stcmd.......' ..... ettt bbbttt a st ettt i 'Heat Exchcnger T eSS o e et eee bt enes Heat Exchanger Tube-Spqcer Pressure Drop Tests ..................................................... et benes ................................................................................. arsasssncen sseuss 4teE N Lt NPT It e s eaassansb et seentactndnnebbbbbatat b Thermal Cyclmg Te ts of Sodlum-inconel Berylhum Sysfem ettt et Gas-Flred Hecn‘ Source ettt eraeen e e o Room Temperature Ref | , “Reactivity Measurements ... | - Power Distributions................ et e eetteereitteteeniareteee hrateeiae_aeeeeieteteeeitnsaeeannteaeeaaneteseanteesee s tanreseneraes Critical Experiments.....cccocooniieirinnacn. Assembly for High- Temperature Critical Experament ............................................................................ [ OREWORD ........................................................................................................................................................ 15 15 17 18 19 20 20 20 21 22 26 26 26 26 27 27 29 29 30 30 31 32 32 33 34 35 35 35 37 37 37 39 42 43 43 43 46 48 Vil | - | PART lI. MATERIALS RESEARCH | b B 4. CHEMISTRY OF REACTOR MATERIALS .....ooovvvocrurreemiemmnienececensssssemsimsesseeseseresessssssssssbosssss s 51 Phase EqUiliBrium STUBIE@s ...o.oouieeee oo ee ettt st st e .. 51 b The Binary System LiF-ZeF f.oooo s 51 T The Ternory System NaF- LlF Z e 52 ' " The Quaternary System NaF-LiF- ZrF UF s e 55 _BeF ~Bearing SysStems ... e e arre e 56 Visuai Observation of Fluoride Melts ... 59 - Phase Separation by Zone Melting ...t 59 Chemical Reactions in Molten Salts ..., et e 60 Equilibrium Reduction of FeF, by H, in NaZrF oo e 60 Reduction of UF by Structurcl Metals ................................................................................................ 60 S’rablhfy of Chromlum and Iron Fluorides in Molten Fluorides ......ccoooniviiniininicecincee, 63 The Disproportionation Pressure of U g e 64 Reduction of UF, with Uranium in Alkali Fluorides ..o, e ereeans 67 . Effect of Filter Medium on Stability of U e, 69 * Stability of PbF -BeF, Melts in Inconel ..o 71 Soiublhfy of Metals in Molten SIS ot ne e s ee ey 71 __ Produchon Of PUrified FIUOFIAES ..oveoeeeeeeeeeeeeeeeeeeeeeeeeeerse et ssn s v essseesesneneannnes 12 3 - Fuel Purification and Preparation Research ..., et 72 Pilot-Scale Purification Operations .........ccociiviervieiviieniiee et esseeeeesre st e seesaasen e saesseanes 73 i Production-Scale Operations ...ttt e ettt e st e eatean e baaten 73 ] Batching and Dispensing Operation ..........cccieeriieriireie ettt e O 75 Loading and Draining Operations ...........c.ccoioiiiiiiiiiiiiiii ittt et 75 P A SOIVICES cuuiiiviiiie ettt ettt ettt ge et at e s aateseeane e eraeeneesaeenseenneetans 76 Experimental Preparation of Simple Fluorides ... 76 Fundamental Chemistry of Fused Salts ..ot 77 EMEF MeaSUrEMENTS ..oiviiiiieiieiiii ettt e s bb s e b e bbbt et es b st e sb e nb b 77 Vapor Pressures of LiF-ZrF , Mixtures ... s 80 Solubility of Xenon in Fused Salts ... 81 X-Ray Diffraction Study of Liquids ..o 8] High-Temperature X-Ray Spectrometer Studies ....oooviiioiiiiioiiie e eeeer e srasssine e 81 5. CORROSION RESEARGCH ..ottt ettt bbbkttt ese s enenee 83 4 Forced-Circulation Studies ..o ettt 83 ; Fluorides in Inconel . ..., e - 83 Sodium in INCONEI .. oo et cereeeieereeens 86 5 Thermal-Convection StUIEs .....ooooiiiiiicee ettt eae e et srs e srs e eans et 88 Alkali-Metal Base Mixtures with UF and UF in lnconel oo et eeeeteeraaaaareeaans 88 3 Zirconium Fluoride Base Mixtures W|th UF, and UF, in Inconel oo, 90 - 1 Effect of Temperature on Mass TranSFEr ..o ovoiieeeeeee ettt enns 91 E i Effect of Loop Size and Shape on CorroSion .......c.ccccoeerivieieeisresienseseesesscesss et ssssssss e 1 { Evaluahon of Control Loop Results ..ttt e 93 1 Generdl Corrosion STUGIES ..ot e 94 s b ; Brazing Alloys on Type 310 Stainless Steel and **A’’ Nickel in Sodium and ‘ o F in Fuel Mixtures ... s . 94 : Screening Tests of Solid-Phase Bonding ..ot e s 9% v Sodium in INCONE] .. ..o e e ettt b ettt eenes e 101 O vildi Lithium in Type 347 Stainless Steel. ... 101 - , Versene Cleaning of Beryllium-Inconel Sysfems ................................................................................ 102 Fundamental Corrosion Research et e bRk SRR bbb e et s 105 = Film Formation on Metals ... cvvreernsnneenes 105 | Mass Transfer and Corrosion in Fused Hydroxide ... 110 Chemical Studies of Corrosion .......ccccoevvvvevienvvinencnneenn, e et et e e ettt e e ara e st e e et 114 Corrosion of Inconel by LiF-BeF, and by LiF-BeF -UF , s 114 Effect of UF ;-UF , Mixtures on Corrosion of Inconel by Various Solvents ... 114 Studies of the Sodium Hydroxide—Nickel Reaction ........cocooiiinininiiiii, 115 6. METALLURGY AND CERAMICS . oot sre s er e e eb e st 119 Development of Nickel-Molybdenum Aoy s .......cociviiiiviiieieieiee st 119 Fabrication SHUAIES .ovv ottt ereb ettt e er e st bbb b anes s 119 Oxidation Studies . ....o.ocevoiiiiieiice e, e 123 Stress-Rupture Studies of Nickel-Molybdenum Alloys ... 124 HOSTEIOY B oottt 124 Modified Nickel-Molybdenum Alloys ... 125 Tensile Properties of Hastelloy B ..o 125 Development of Brazing AHoys ..., ettt 128 High-Temperature Oxidation Tests ..., 128 Phy sical Property Tests oo e 129 Fabrication of Test Components.....c.ccoociiiniiiiicicciiccrric e e 129 Twenty-Tube Heat Exchanger ... 129 Intermediate Heat Exchanger No. 2. ...t 131 - NAK 10-ATr RAAIGEOIS ...oiiiiiiiiiiiiiicieet et e ettt s e sttt et et s s e st ebe s sbe e b b eas s enan 134 " Cornell Radiator No. T oot ettt es bbbt s i ea e s s ensn st s e 139 - - Special Materials Studies......cccoovirrennn. ettt ettt ettt anenes 140 3 Special Aoy EXIUSIONS ottt 140 ' Clad-Columbium Fabrication .ottt et bbb e st s ene s 140 B6C~Cu Shielding Material .. ..c.ooiviiiii e 142 Magnesium-Lithium Aoy .o 143 Welding of Columbium Thermal-Convection Loops ..o, 144 Dimensional Stability Test on an Inconel Spun Core ..o 144 Brazing of Cermets to Inconel ... ... 145 Ceramic Research .....c.ioovevennenne. ettt teeststateasateansiteteseshaEeEetat e aeate et e Rt et et e see e e R e sarat e e bearerearennerene 145 ’ RAre=Earth CeramiCs oo eeeieeeeeiesree e et st e e s ta ettt etaseese s as e e ebeabe e smecamebeeaaesnesssressasarn s bn st e arabasabaeabes 145 UO, Particles Coated with ZrO ..o 147 Graphite-Hydrogen Reactions and Erosion lnvestigation ..., 147 7. HEAT TRANSFER AND PHYSICAL PROPERTIES ... 149 Fused Salt Heat Transfer.................... s et e et 149 ART Fuel-to-NaK Heat Exchcnger ..... ettt e e s 149 ART Core Hydrodynamlcs e ' ' 151 Reactor Core Heat Trcmsfer s 154 - Free Convection in Fluids Having a Vo[ume Hecf Source et eibeeas e b ettt et er et b s st en ettt s et e 154 3 HEG CaPACHY .ooiiiiiii et bk sa s e S0 h bbb 156 ix 10. 1. 12. VESCOSIEY oottt ettt ettt ettt bbb h st st et et b e bbbttt bs bt e aa bt s bbbt a ettt be Rt e et b e s - 157 Thermal COndUCHIVITY «.eomeeereeeeeeseeeeeeeesseeeeseeeeeeseeeseeeeresessoeseeseeeeees et 159 Electrical ConducHiVity ..ot ....... 160 RADIATION DAMAGE .ottt bbb es bbb s ettt 161 MTR Static Corrosion Tests .............. e ettt eteanes e 161 LITR H'ofiiontal-Beam-Hole Fluoride-Fuel Loop ....cocoooiii e e . 163 Deposition of Ru'03 in LITR Fluoride-Fuel Loop .ooooivorieeeeeeeeeoeseoseeoes e eeeeeeeeeseeeeesesssese s eseeer e 167 Miniature In-Pile 00D ettt et ettt ettt e 168 Delay of Fission Gases by Charcoal Traps ..o 169 ~ Creep and Stress-Corrosion Tests ..o, e 170 " A Theoretical Treatment of Xe 135 Poisoning in the ARE and the ART ... 171 ANALYTICAL- CHEMISTRY OF REACTOR MATERIALS ..... 174 ' Défé'r'min'ation of Uranium Metal in Fluoride Salt Mixtures ..o..oooooevooovevoooeosees e 174 Det.err.niinction of Trivalent Uranium in Fluoride FULS w.....ovvovrorooeoeeoeeoeoeoeoeoee 175 Oxidation of Trivalent Uranium by Methylene Blue ..o, 175 Simultaneous Determination of Trivalent Uranium and Total Uranium........c.ooooii 176 Determination of Lithium in LiF-BeF, and LiF-ZrF4-UF4 ................................................................ 176 Volumetric Determination of Zirconium in Fluoride Fuels with Disodium Dihydrogen Ethylenediaminetetraacetate .. ..ottt ettt e 177 Determination of Oxygen in Fluoride Fuels ........cocoooiiiiiii e 178 Determination of Oxygen in Metallic Oxides by Bromination .........c.cccoeiiiiiinnnnccececnnes 178 Determination of Oxygen in Beryllium Oxide by Acidimetry ........c.ooiiiiiiiiiieceee 179 Determination of Trace Amounts of Nickel in Fluoride Fuels with Sodium Diethy Idithiocarbamate ... e e e b e er e s 179 ANP Service Laboratory . ..ottt et eb e ere et ebe et e asEr st et ar et ans 180 RECOVERY AND REPROCESSING OF REACTOR FUEL .ooooooooooooooeoeooeeeeeeeeeeeeeeeeeee oo 181 Pilot Plant Design ....coiiieiiiiiceiee ettt ettt st s et e enere e te e se s eneeneteensens 181 Process DeVElOPMENE o .ottt ettt et n et et e ettt eaee e e 181 Corrosion Studies 182 PART Ill. SHIELDING RESEARCH SHIELDING ANALY SIS Lottt et eass et s e st e aeas - 191 Gamma-Ray Distribution in a Circulating-Fuel Reactor and ShIEld oo 191 Enérgy Absorption Resulting from Gamma Radiation Incident on a Multiregion Shield | ._ - With SIab GeomMEtry ..ottt e nes e, 192 Energy and Angular Distribution of Air-Scattered Neutrons from a Monoenergetic Source......... 192 Analysis of the Constant-Velocity Transport EQUation..........ccocooovmeeeoeeccrecereereorereseesecessisscercrssine 193 LID TANK SHIELDING FACILITY ........... . 194 | Reflector-Moderated Reactor and Shield Mockup T SES oo 194 st | ) Gamma-Ray Dose Rate Measurements ...t e ve e 194 NeUtron MeaSUFEMENTS ...ttt ettt se st s et et e eae st st reee e e aeene 197 Sodium Activation in Heat Exchanger Region ... 198 13, BULK SHIELDING FACILITY oottt eb ettt s 200 GE-ANP Air-Duct Mockup EXperiment ......c.ocooioiiiiiii et 200 The Spectrum of Fission-Product Gamma Rays ... 203 14, TOWER SHIELDING FACILITY Lot 205 The Differential Shielding Experiments at the TSF: Phase | ... 205 Measurements in the Detector Tank ...t e 205 Measurements in the GE-ANP Crew Compartment ..., 205 Analysis of the Differential Shielding Experiments ... 206 Definition of Dose Scattering Probability ... 208 Evaluation of Probability from TSF Experiments ..o, 213 Evaluation of Direct-Beam Integral ..o e 214 Calculation of Scattered Dose ... 215 Effects of Direct-Beam Collimation ... e 216 Effect of Neutron Energy Spectrum ... e 216 Application of the Differential Shielding EXPEriments ....coooiiooiooeoeeee oo eeeeeeee oo 217 PART IV. APPENDIX 15. LIST OF REPORTS ISSUED FROM FEBRUARY 1955 THROUGH MAY 11, 1955 __..coooooooovceeeeenee. 227 ORGANIZAT'ION CH AR T S et e 231 Wfif xi it o \"16 5%, on:4 k savmg Il"l urcmlum concen’rra’rlon “of the fuel cnd a shREY ANP PROJECT QUARTERLY PROGRESS REPORT SUMMARY PART I. REACTOR THEORY, COMPONENT DEVELOPMENT, AND CONSTRUCTION _1.' Reflector-Moderated Reactor The development of the reactor layout is con- tinving. New features that have been incorporated because of stress, fluid flow, or fabricability considerations include an elliptical fuel expansion tank, a rounded dome to enclose the top of the reactor, a newly designed sodium pump impeller, and other related items. Recently completed heat exchanger tests yielded consistent data from which a series of heat exchangers is being designed. The most promising of these will be chosen for the ART. The preliminary layout of the interior of the ART test cell which shows the major items of equip- ment and the recommended provisions for support was completed. A similar layout for the NaK piping and radiators was also completed so that drawings for the building and concrete work could proceed. More detailed drawings of the cell that will show the disposition of the small items and the instrumentation are being prepared. Information block diagrams showing the basic- control actions desired for the ART have been prepared and will be used as a basis for selection of the hardware types and the control techniques to be used and for determination of areas of con- trol that will require new component development. The room-temperoture critical experrmenfs for ‘the ART have shown ’rhot 'rhe critical mass de " creases’ very rapldly as the neutron obsorphon in ~ _the core shell decreases. Hence, replacemenf of - the hlghly absorbmg Inconel by a low-cross- sectlon ' “:'material, such os columblum, would brlng about 0 substantial scvmg in uranium ‘investment and, mudenfclly, make the temperature ‘coefficient more ~ ““negative. For fabrication | reasons, only about 50% of the Inconel could be replaced by columbium. Itis esflmofed _that the effect of a 0% replacemenf ical mass by about 3 wh C:h 'o : onds Yo a 0.7 mole % U235 ivestment. Calculations were made of the temperature coef- ficients of the ART that check well with previous 7-kg saving in total multigroup calculations. The results support the postulated negative over-all temperature coef- ficient. It has also been found that, if as much as 5 wt % lithium were added to the NaK in the secondary coolant circuit, the reactivity of the ART would be a fairly sensitive indication of a NaK ledak into the fuel circuit. Burnup and gamma- ray heating of rare-earth oxides being considered as control rod materials are being studied. 2. Experimental Reactor Engineering A series of design changes have been made in the nose, or heater, section of the in-pile loop as more information concerning the neutron flux in the MTR beam hole and the flux depression of the materials of loop construction has become avail- able, The present nose section consists of a 2V sturn coil (an increase of 1 turn) with its axis parallel to the beam hole center line. The flux seen by the fuel in the in-pile loop is now expected to be 30% of the unperturbed value, and the average power density in the nose section will be 0.7 kw/em3. A working mockup of the in-pile loop has been completed and is being operated with the fuel mixture NaF-Zer-UF4 (53.5-40-6.5 mole %) at a Reynolds number of about 5000, a temperature differential of 175°F, and a maximum fuel temper- ature of 1500°F. Resistance heating is employed. The experience gained will aid in operation of the in-pile loop at the MTR. The fission gas holdup system for the loop is " being tested with flow rates of 0.15 and 0.03 scth of helium with about 0.13% krypton, After 15 days “"of operation, no significant amounts of krypton were getting through the liquid-nitrogen-cooled charcoal-adsorption traps. Assembly of the first loop for the MTR in-pile experiment is under way, “ with operation scheduled for the next quarter. Three additional test stands were installed for ‘the operation of resistance-heated high-velocity - forced-circulation |oops with iorge tempero’rure d:fferenfluls for mveshgahng corrosmn "and mass “transfer of Inconel by fluoride fuel mixtures. Also, three more gas furnaces were placed in service as heat sources for these loops, bringing the total o il o e bifig - Golhio . FEiieobio S ANP PROJECT PROGRESS REPORT humber of test stands to ten. Operation of 16 loops was terminated during the quarter. The majority of these loops operated for 1000 hr in the Reynolds number range of 1,000 to 15,000 and with temperature differentials of 100 to 300°F. The ‘éx‘éés's'iv'e' high wall temperature in the bends of electric- resnstonce heated loops was corrected by relocation of the heating elements so that heat is applied in only s?rmghf sections of tubing. Four test stands are now in operation for studying éotrosion and moss frcmsfer of sodium-Inconel and sodlum—lnconel-s’rcmless steel systems. Oper- ation of six such loops was terminated this quarter at the conclusion of periods of either 500 or 1000 hr with maximum sodium temperatures of up to 1500°F. Appreciable deposits of mass-transferred material were found in the cold legs of these loops, and fherefore a controlled series of experi- ments’ was “started. The loops will provide infor- mation on ‘fHé‘éffecfs of the oxygen content of the sodlum, the use of a cold trap, the use of a lower temperature, and the use of an all-stainless-steel system, - The ART fuel pump (model MF-2) was operated in water performance tests. Some cavitation noises were present in all tests, but no serious effect of the apparent cavitation on performance could be found. It is estimated that the efficiency of the pump at design point, exclusive of seal and bearing losses, is 70%. New inlet volute configurations are currently being designed and tested. The bearing and seal, cold mechanical shakedown, and high-temperature test stands are being fabricated and assembled. An extensive program of heat exchanger testing is under way for obtaining reliable heat transfer data on fuel-to-NaK heat exchangers of the general type and configuration for ART application. Con- siderable information will be obtained on the effect of corrosion and mass transfer on materials of fabrication in high-temperature, high-heat-flux, NoK and fluoride systems. An opportunity will also be provided for ascertaining the structural integrity and reliability of fabrication of heat exchangers and radiatou:s supplied by outside vendors. _The program involves the operation of three intermediate heat exchanger (IHE) test stands and two small heat exchanger (SHE) test stands, such as the SHE stand now in operation. The IHE stands will be used to test large tube bundles (about 100 tubes) of the general size and configu- ration of ART heat exchangers (arranged for regenerative operation), while the SHE stands will be used to test smalier, more easily fabricated, tube bundles (20 to 50 tubes). The SHE stand now in operation will be modified for use as a general test loop following termination of the current test. Design of all IHE test stands is complete, and procurement and assembly are well under way. Design of the SHE test stands and procurement of equipment are under way. - Water tests with a full-scale aluminum model of the 21-in. ART core and the entrance header region were initiated. Without inlet guide vanes, the fluid was observed to enter the core at an angle of about 70 deg from the vertical. Flow reversal at the island was observed. Data from these tests are being used in the design of turning vanes and vortex generators to correct this unacceptable flow condition, ' I . A second thermal-cycling test of a sodium- Inconel-beryllium system was completed. Over 100 thermal cycles were applied to the beryllivm piece between the range of the high power level, 61 w/cm® to the beryllium, and the low power level, 2.5 w/ecm®. The sodium temperature from inlet to outlet at the high power level ranged from 1050 to 1200°F. Inspection of the beryllium piece after the test {total operating time, 1030 hr) re- vealed three axial cracks on the outer surface of the hot end of the beryllium, The 100-kw gas-fired heater was tested and was found to perform very satisfactorily. Minor modi- fications can be made to this heater that will increase the capacity. 3. Critical Experiments The critical assembly of the reflector-moderated reactor consisting of the beryllium island and reflector enclosing the fuel region and having axial extrusions simulating the exit and entrance flow channels was loaded with sufficient uranium to give several per cent excess reactivity. This overloading was used to evaluate some of the materials of interest in the design of the ART prior to dilution of the fuel to the critical uranium density. ' o Measurements were made with samples of a mixture of the oxides of the rare-earth elements being considered for the absorber material of the control rod of the ART. A cylinder of the mixture FRERET T =f ) e cal properhes ‘l_oeh‘ :..,f:v_{:NaF ZrF4-UF4*system' has been’ found . - Study of the analogous NaF-LiF-BeF, system “has shown fhcn‘ adequateiy low’ melflng pomfs are “Yespect 16 pure UF, and uranium metal at temper- ” cn‘ures below 1400°C 0.79 in. in diameter and 21 in. long decreased the reactivity nearly 2%; measurements with shorter lengths of the material indicated that the total value of the rod could be mcreosed by 60% if the diameter were increased to 1% in. Tests made on tubes of several different dimensions gave data for the design of the control rod guide thimble. In other experiments, columbium was shown to be somewhat less poisonous than Inconel in a neutron spectrum similar to that expected in the ART, and it was found that a layer of beryllium in the center of the fuel channel of the critical assembly reduced the critical uranium concen- tration by a few per cent. Fission rate distri- butions were measured across the flow channel at several locations in the central reactor region and in the end duct. A high-temperature critical experiment embodying the nuclear characteristic of the ART has been designed which will operate at zero nuclear power at about 1200°F. The purposes of the experiment are to measure the critical uranium concentration, the temperature coefficient of reactivity, and the effectiveness of control rods. PART |.| MATERIALS RESEARCH 4, Chemlstry of Reactor Mufenals Interest in obfqmmg fuel mixtures more suitable for use in a C|r¢u|ahng -fuel reactor than those available in the N'cx'F-Zi'F“'-UF“ system has led to evaluation of the NaF-LiF-ZrF, ternary and the NaF-LiF-ZrF,-UF, quaternary systems. The NaF- LiF-ZrF, system has been reasonably well de- fined, but much work remains to be done on the quaternary system. Phase-equilibrium data show that quite low melting pomfs are available in the ternary system ‘and that adequa'rely Tow meltmg"-'}"’ points are available at Zrl':4 cohcentrations as low . as 21 mole %, however, no composition ‘with physr- ' fter than those avollable in fhe_'_ ; available over wide areas. Phys:cal properfy dafa - -have not yet been ‘obtained in sufficient detm[ for :"compquson wu’rh avo_[lable fuel. mo’rertal Pre-v: - .llmmary dofo on the éo '{fblllfy of UF3 in BeF k becrmg composmons e » data show a trend of increasing UF3 SOIUblIIfy-M- obfumed The sc cm‘ered with increasing temperature and decreasing BeF, content, but in no case does it appear that the "PERIOD ENDING JUNE 10, 1955 solubility of UF, at 600°C is sufficiently high to provide more than. a fraction of the concentration needed for an ART fuel. Previous measurements of the partial pressure of HF at equilibrium during the reduction of FeF, by hydrogen in NaZrF, showed higher valuves than would be predicted from thermochemical data and ideal solution behavior, and it was postulated that the higher values were due to a lowered activity of the metallic iron because of alloying with the nickel apparatus. The postulated alloying has been confirmed by low values for final FeF, con- tent of the melt and by chemical analyses of portions of the nickel apparatus. Additional data were obtained on the reduction of UF, by structural metals. Data on the reduction of UF, by metallic chromium with NaF-Zer-UF“ (51-45-4 mole %) showed that, in comparison with the data obtained with NaF-ZrF“-UF4 (48-48-4 mole %) as the solvent, an increase in the final mole fraction of NaF from 0.48 to 0.51 in the melt containing uranium causes a significant decrease in the equilibrium CrF, concentration. Data for the reaction of UF, with metallic iron in these mixtures and in NaF-KF-LiF (11.5-42-46.5 mole %) agreed very closely and were somewhat higher at 600°C than ot 800°C. Some, as yet inconclusive, data were obtained on the reaction of UF4 plus UF. with chromium metal and Inconel in NaF-KF- LiF (11.5-42-46.5 mole %) at 600 and 800°C. Ad- ditional data were obtained that confirmed previ- ous findings that FeF, is relatively stable in the NaF-KF-LiF eutectic at 600 and 800°C and that CrF, is not stable. Previous evidence that UF, was more stable at elevated temperatures than free energy estimates ‘"had indicated was substantiated by vapor pressure measurements on UF, in the temperature range 1270 to 1390°C. The disproportionation pressure "curve that was ob'romed shows that the dispro- ~ “portionation of UF_ is far from complete under the “conditions prevallmg in the vapor pressure cell and that UF, is thermodynamically stable with In the’ mvestlgahon of varmbles ‘affecting the “reduction of UF4 with-uraniom in alkali fluorides, the effects of the surface area of nickel or copper “exposed to the melt and of adding excess uranium metal were studied. |t has become apparent that alloying of nickel and uranium can occur ot temper- oI ANP PROJECT PROGRESS REPORT atures far below the minimum nickel-uranium liquidus temperature (732°C) and, consequently, that disproportionation of UF, can be expected to occur at the temperatures of interest at nickel surfaces or at any metallic surface with which uranium can alloy. Preliminary evidence was obtained which indicated that UF, dissolved in an alkali fluoride mixture is more stable in copper than in nickel. Investigations of methods for rapid purification of fuel mixtures included attempts to use electroly- sis under a hydrogen atmosphere to remove oxides in order to avoid the container corrosion that results when HF is used and to use metallic zir- conium to replace most of the hydrogen in the stripping operation. The use of zirconium metal was demonstrated on a 5- and a 50-1b test scale and was found to be a quite rapid and effective method for purification if small quantities of UF in the product are tolerable or desirable. In - electrolysis experiments, the ZrF ,-bearing mixtures could be electrolyzed smoothly, but the alkali fluoride mixtures gave variable results. Attempts to prepare mixtures containing UF, and no UF, were unsuccessful, the largest UF, content at- tained being 85% in an NaF- ZrF, base. Fifty-six pilot-scale preparonons totaling 630 ib of material were produced in various compositions for small-scale corrosion studies, for physical property determinations, or, in many cases, for use as purified intermediates in phase-equilibrium studies. Uranium trifluoride was a component of nearly 25% of the materials requested. Production operations were resumed on March 1, 1955, on a three-shift, five-day-week basis to provide test material for the greatly accelerated ANP engi- ~ neering program. A total of 4800 b of purified material was prepared during the quarter. Attempts to flnd a commercial source of ZrF, are under ~way. |f a commercial source is not found, it will be necessary to expand the Y-12 production facili- " ties immediately, Two batches of enriched fuel were prepared for use in an in-pile loop, and prepa- rations are being made for the production of the materials for the proposed high-temperature critical experiment. ' Potential measurements were made with combi- * nations of s'-ever‘ol__ half cells consisting of metal - electrodes bathed in solutions of the corresponding metal ion in the molten salts. The temperature ) rdh’éé"'stqdi'éd.' was, in general, 550 to 700°C. Cells consisting of zirconium rods immersed in various NaF-ZrF, melts and cells consisting of metallic nickel electrodes immersed in solutions of NiF, in molten NaF-ZrF, melts were studied. Vapor pressure measurements of mixtures in the LiF-ZrF, system were started. The data showed the vapor pressures of the LiF-ZrF, mixtures to be considerably higher than those of the corre- sponding NaF-ZrF, mixtures. An x-ray diffractometer for studying the structure of liquids has been constructed and is undergoing final testing. A high-temperature attachment for an x-ray spectrometer has been used for studies of compositions in the systems NoF-ZrF,, LiF- ZrF,, and NaF-Ber. 5. Corrosion Research Several Inconel forced-circulation loops that were operated with fluoride mixtures and with sodium as the circulated fluids were examined. The fluoride mixtures included ZrF ,-base mixtures with UF, and with combinations of UF3 and UF and an olkch-mefol base mixture containing UF Favorable results were obtained with the ZrF - base materials in that the depths of attack were no deeper than have been found previously in thermal-convection loops. Attacks as low as 5 mils in 1000 hr appear to be obtainable. The conversion of some of the UF, to UF, decreases the attack. The most important variables appear to be maximum wall temperature and the hot-zone surface-to-loop volume ratio, whereas large vari- ations in velocity and Reynolds number have very little effect on the depth of attack. The alkali- metal base mixture containing UF, caused exces- sive mass transfer and a very heavy intergranular concentration of subsurface wvoids to a depth of 35 mils. The mass-transferred deposit in the cold zone was up to 85 mils thick. Mass transfer of large amounts of nickel metal was found in the Inconel forced-circulation loop - that circulated molten sodium at 1500°F. Layers of dendritic metal crystals up to 26 mils thick were found to have formed in 1000 hr. The use of type 316 stainless steel in the cold portions was found to reduce the mass transfer slightly, but further study of the variables in the process is needed to confirm this finding. : Alkali-metal base mixtures containing combi- nations of UF, and UF, were circulated in Inconel thermal-convection loops, and, when about 2 wt % 4y ) kY yranium was present as U3+, low depths of attacks and no deposits were found. Higher U** concen- trations resulted in decreased attack, but hot-leg layers were found. Inconel thermal-convection loops in which ZrF ,- base fuels containing about 2 wt % uranium as U3* were circulated did not show so large a re- duction in attack as was found in the loops that circulated alkali fluoride fuels also containing 2 wt % uranium as U3¥, However, some reduction in depth of attack and a fair reduction in amount were found. The effect of the hot-leg temperature (1200 to 1600°F) on mass transfer was investigated in several Inconel thermal-convection loops oper- ated for 1500 hr. A definite increuse in depth of attack with an increase in hot-leg temperature was noted that may be attributed to mass transfer, inasmuch as loops operated previously for 500 hr did not show the effect of temperature on depth of attack. The mass transfer effect is masked by the effect of impurities and nonequilibrium conditions during the first 500 hr. Considerable work is being done in an effort to find the cause of the increases in depth of attack and the nonuniformity of results now being ob- tained in Inconel thermal-convection loops operated as control loops under standard conditions. Con- tamination during filling or operation does not appear to be the cause of the difficulty. Corrosion tests of brazed type 310 stainless steel T-joints in static sodium and in static NafF- ZrF4-UF4 (53.5-40-6.5 mole %) showed the brazing alloy 9% Si-2.5% P-88.5% Ni to be satisfactory in both mediums. Similar tests of brazed ‘A" nickel T-joints showed the following brazing alloy's to be satisfactory in both mediums: 90% Ni~10% P, 80% NI—IO% Cr—]O% P, and N:crobrqz Seesaw _corrosmn fesfs in both mediums on brazed |ncone| ‘ T-|01nfs showed the Coast Mefcls alloy No 52t0 - have the best resus'rcmce in sodium and in 'rhe"_," \ ol uorlde mixture. | " Since fhe sfrucfurai mefcl alloys ‘that have beenhw : *'-*:proposed for use in fl'\e fabrication of reactor fuel and coolant lines have a tendency to form solid- phase bonds at elevated tempefatures in liquid metals and in fused salts, they are unsun‘ob[e for use in valves bearlngs,' and seals. Therefore cermets and ceramics that do not bond to each ~ other but that can be bonded to the structural metal alloys are being investigated. The cermets (metal- bonded ceramics) appear to be the more promising ~ PERIOD ENDING JUNE 10, 1955 because of their high corrosion resistance and other satisfactory chemical and physical properties. Several Kentanium cermets tested in NaF-ZrF - UF, (53.5-40-6.5 mole %) at 1500°F for 100 hr were found to have good resistance to solid-phase bonding if the contact pressure between the speci- mens did not exceed 50,000 psi. The compositions of the cermets tested were: 80 wt % TiC-10 wt % NbToTiCs—IO wt % Ni 70 wt % TiC=10 wt % NbTaTiC3-—20 wt % Ni 64 wt % TiC—6 wt % NbTaTiC,~30 wt % Ni 64 wt % TiC~6 wt % NbTaTiC, - 25 wt % Ni=5 wt % Mo Lithium was circulated in two stainless steel thermal-convection loops for periods of 1000 and 3000 hr, respectively. The hot- and cold-leg tem- peratures were 1000 and 550°F, respectively. Operation was satisfactory throughout the test periods, and macroscopic examination revealed no mass-transferred crystals in the loops or in the {ithium drained from the loops. Metallographic examination revealed subsurface voids and a ferritic surface layer 0.3 to 1.0 mil thick in the loop operated for 1000 hr and 1.0 to 1.5 mils thick in the loop operated for 3000 hr. Lithium metdl had penetrated to the depth of the subsurface voids. The weld zone of the pipe was attacked to a depth of 3 mils in the 1000-hr test and 4 to 5 mils in the 3000-hr test. A few small (0.2-mil) carbide par- ticles were found attached to the wall in the cold- leg sections of both loops. Determinations of the oxidation rate of sodium " have been made at —-79, —-20, 25, 35, and 48°C. The experimental results indicate that, contrary ~to current oxidation theory, the oxide films formed on sodium are highly protechve in the absence of water vapor. The rate curves do not conform to any of the ‘“standard’’ rate equations reported in the fiterature, but they are quchfohvely comparable to low-temperature curves for copper. It is hoped ‘to clarify the oxidation mechanism associated with a linear rate law through a careful study of the structure and composition of the oxide films formed on columbium in the neighborhood of 400°C. Studies of corrosion and mass transfer by fused ’-Thydromdes indicate that both nickel and Inconel may possibly be compatible with hydroxides at temperatures of about 600°C. Hastelloy B is unsatisfactory because of its poor corrosion re- e A i RO s b o i B i s SR ANP PROJECT PROGRESS REPORT sistance. In the temperature range 600 to 700°C, there is evidence of an accelerated rate of mass transfer with both nickel and Inconel, as well as corrosion of Inconel. ‘The Ber-Bearing mixtures LiF- -BeF, (69-31 mole %) and LiF- BeF,-UF, (67.3-30.2-2.5 mole %) were tested in Inconel capsules in 100-hr tilting furnace tests. No evidence of attack by either mixture was found. A study of the effect on cor- ~rosion of the rcfrid of UF3 to UF, in various solvents revealed that increasing the UF_ content ‘up to 50% was beneficial and that any further increase had little effect. v '6. Metallurgy and Ceramics ~Investigations were continued in the study of the properties of nickel-base alloys containing 15 to 32% molybdenum, ternary alloys with a nickel- molybdenum base, and Hastelloy B. Attempts are being made to improve Hastelloy B with regard to fabricability, oxidation resistance, and mechanical properties. Additional evidence has been obtained which indicates that the poor high-temperature fabricability of the commercial material is related to the impurity content; however, it is felt that the superior strength of commercial Hastelloy B may be derived from the impurities. Therefore mechani- cal property tests are under way on an alloy with the nominal Hastelloy B composition, 4% Fe-28% Mo—~68% Ni, but without the tramp elements va- nadium, silicon, manganese, cobalt, chromium, tungsten, and aluminum. A tube blank extruded from a vacuum-melted cast billet of commercial Hastelloy B fractured during the first step of a reduction operation; however, a blank made from a wrought billet was success- fully reduced from 1.5-in.-OD, 0.250-in.-wall to 0.187-in.-0D, 0.017-in.-wall seamless tubing. Two impact extrusions of as-cast vacuum-melted com- mercial Hastelloy B were made at 2000°F with good recovery of sound rod. Attempts to roll the rod at 2000°F were, however, unsuccessful; the material cracked severely. ‘ 'De'sign curves were prepared from the results . Vof creep-rupture tests of Hastelloy B sheet in the ‘solution-annealed condition in an argon atmosphere at 1500 and 1650°F. A comparison of these data wnth preliminary ‘data from tests in fused salts shows that properties of the alloy in the fused salts are actually superior to the properties in an argon atmosphere. Creep tests in air, in argon, and in hydrogen at 12,000 psi showed the effects of argon and hydrogen to be similar, but the effects of air followed closely the pattern observed for Inconel and ““A” nickel in air, insofar as reduced creep rate and longer rupture life are concerned. However, the final elongations of Hastelloy B in air are equivalent to or lower than those in argon, whereas the elongations of other nickel-molybdenum alloys are markedly greater in air than in argon. A program has been initiated to determine whether aging treatment results in serious embrittlement at service temperatures. Typical microstructures have been obtained from specimens heat-treated at temperatures from 1100 to 1600°F for times from 100 to 1000 hr, and a correlation is to be made between physical properties and the microstruc- tures. The relative merits of various preaging heat treatments of Hastelloy B are also being studied, and it is hoped that, as a result of this work, a procedure can be developed that will stabilize the microstructure sufficiently to reduce the sensitivity to high-temperature aging. The binary nickel-molybdenum alloys that have been studied have included 15 to 32% molybdenum, and the ternary alloys contained 20% molybdenum plus nickel and one of the following: 3 to 10% chro- mium, 2 to 10% columbium, 2% aluminum, 1% tita- nium, 2% vanadium, or 1% zirconium. The results of stress-rupture tests have shown low strengths and ductilities for most of these alloys and indi- cate that vacuum melting alone is not sufficient to obtain optimum properties. The possibility of increasing the strength and ductility through cerium additions appears to be promising, since the ad- dition of cerium has been shown to improve the physical properties of the alloy. It has become apparent that improved deoxidation practices are necessary in melting these alloys, and efforts are being directed toward the production of sounder ingots. Oxidation tests of nickel-molybdenum alloys containing 3 to 10% chromium have shown that about 10% chromium is necessary to form a non- spalling protective oxide on the alloy. However, the oxidation rate under static conditions can be reduced 50% by the addition of 3% chromium. Cyclic tests consisting of 190 air cools from 1500°F in 500 hr have now been completed for brazing alloys previously evalu‘otec{' in static oxidation resistance tests at 1500 and 1700°F. Most of the alloys tested showed good resistance v to oxidation under both static and cyclic con- . ditions. Cyclic tests at 1700°F are now under way on these brazing alloys, which include commercial alloys, experimental nickel-base alloys, and ex- perimental precious-metal base alloys. Physical property tests showed that the physical properties of Inconel are not impaired by the brazing process. The fabrication of several major experimental heat transfer test components was completed. The items fabricated included a 20-tube fuel-to-NaK < Inconel heat exchanger, a full-scale 200-tube fuel- to-NaK heat exchanger (intermediate heat ex- changer No. 2), two 500-kw NaK-to-air radiators, and a full-scale integral-fin liquid metal-to-air radiator designed by the Cornell Aeronautical Laboratory. The combination welding and brazing procedures used in the fabrication of 1rhese units “are described. : Three billets of vanadium were extruded at 2000°F. Tubing prepared from these blanks is to be clad on the outside with stainless steel, and the clad tubing will be used in corrosion studies. Four high-purity molybdenum billets containing 0.7% titanium were extruded into rod to be used in welding studies. Flow pattern studies of the extrusion of duplex and three-ply composites were continued. The three-ply materials are to be used - in the production of stainless-steel-clad seamless tubular fuel elements. Attempts were made to clad columbium with the following heat-resistant alloys: types 446 and 310 stainless steel, Inconel, and Hastelloy B. Columbium was found to be ade- quately protected by each of these alloys, and the effects on ductility were slight; however, interface @ reactions occurred in all combincfi'ons, and there ~was, as d result, separahon at the mterfaces Dn‘fusnon barrlers for assuring bonding at 'rhe _interface were fherefore studied. Preliminary evi- dence indicates that copper W|H be a sutfcble 7 barrier for ?he columbium. 1 __ sensitive; at 200°F the strength was a factor of 4 g™ : below that at room temperature. S fnveshgaflons of the fabrlccbxllfy and mecham_. e éu] Proper’flesl"""f_boron con’ralnlng shleld materials were confmued, ‘and an evoiuahon is bemg mcdef’;f" . of a mogneswm ||'rh|um clloy for use as a struc- - 'fural shleld muferlcxl The addmon of 20% ll‘l‘hlumb,::.l: . _,_‘ro mogneswm reduced 'rhe tensile s?rength and‘j—_-‘:, - T mcreaSed the ‘ductility of “the mcgnes:um.- Data ™ | " obtained at room temperature and at’ 200°F “ahaw the strength of the alloy to be very temperature PERIOD ENDING JUNE 10, 1955 Several type 310 stainless-steel-clad columbium thermal-convection loops were fabricated. A two- stage process was used in which the columbium and stainless steel were welded in separate oper- ations. A typical specimen of a spun Inconel configuration fabricated to simulate the ART core shell design was tested for dimensional stability after thermal cycling. No recordable diametrical, axial, or thickness instability was observed. A method was developed for producing cermet-to- Inconel joints with adequate ductility. Ceramics composed of rare-earth oxides, which combine the property of high absorption cross section for thermal neutrons and the usual ceramic properties of high density, strength, corrosion re- sistance, and high melting temperature, were pre- pared in the shapes required for testing in critical assemblies to determine their possibilities as control rod materials. The possibility of coating uo, particles with ZrO2 to protect the UO2 from reaction with molten silicon in an SiC-Si fuel element is being investigated. Also, graphite- hydrogen reactions are being studied. 7. Heat Transfer and Physical Properties The friction factor as a function of Reynolds ‘modulus was determined experimentally for the case of turbulently flowing NaF-ZrF,-UF, (53.5- 40-6.5 mole %) in Inconel tubes; the results are in agreement with conventional friction data. A full-scale ART fuel-to-NaK heat exchanger was studied as a water-to-water heat exchanger. Pre- liminary measurements indicate that the heat transfer coefficients on the fuel side of the ex- chcnger which has the controlling thermal resist- ances in the system, are about 1.6 times lower than would be obfomed by the conventional relation for turbulent flow in circular-pipe systems; also, the corresponding pressure drops were two times as "high as those for flow in smooth pipes. Some ~velocity profile data were obtained for the 18-in. ART '><‘::_o'|;'e for the case where the fluid enters the core with d_ rotational velocity component; the influence of turbulence-promoting screens at the core inlet on the flow was also studied. A study ‘was made of fhe Temperature ‘and ‘corresponding Tiensile stress flucfucmons in the Inconel walls of .k'the ART core for momentcry periods of flow ) stagnahon near the wall. The results of a theoreti- cal and an experimental study of a free-convection system containing a volume heat source are given. R T T terials were determined: namely, NaF-ZrF,-UF ANP PROJECT PROGRESS REPORT The 'én’rhdllpies and heat capacities of five ma- 4 (50-46-4 mole %), NaF- Z:'F‘1 (50-50 mole %), NaF- 7-)7'_‘ZrF -UF4 (56-39-5 mole %), NaF- LIF-ZrF UF - (20- 55 21-4 mole %), and lithium hydride. The hea’r _ccpacmes in the liquid state of the 17 fluoride ‘mixtures that have been studied to date are repre- ' séhfea by the simple equation where N is the average number of ions and M the average ‘molecular welgh’r The viscosities of seven ~ fluoride mixtures were determined: namely, NaF- LiF- ZrF UF -UF (20.9-38.4-35.7-4-1 mole %), _NGF ZrF -UF (50 46-4 mole %), NGF-UF (66.7- '33 3 mo[e %) NaF KF-UF, (46.5-26.0-27.5 mole%) NaF- LuF ZrF UF UF (20 55-21-3.6-0.4 mole %), -NGF LIF ZrF (22 55. 23 mole %), and NaF-LiF- BeF2 (56 16- 28 mole %). From the viscosity measurements that are now available for BeF - bearing fluoride mixtures, a relationship between BeF. concentration and viscosity was investigated. In general, the viscosity decreased as the BeF concentration decreased. Some preliminary thermal- conductivity data of a ZrF ;-bearing fluoride mix- ture in the liquid state were obtained with a new conductivity cell and are in agreement with previ- ous data obtained with a different type of cell. 8. Radiation Damage The program of MTR irradiations of Inconel cap- sules designed for comparing UF,- and UF ,-base fluoride fuels has continued. The results of exami- nations made thus far have shown no evidence of radiation damage in that there is no corrosion, no significant segregation of uranium, and no changes in the impurity content of the fluoride mixtures. " The fluoride-fuel loop that was operated in the 'LITR horlzom‘cl beam-hole has been disassembled, ; and parts of it have been examined metallographi- cally. Subsurface void attack of the Inconel tubing used for the fuel loop was limited to less than 1 mil in depth. The fission products Ru'®? and 'Nb?% were found to have plated out in two sections . of the loop, and thus partial substantiation of a - Vsumlldr occurrence in the ARE was obtained. The small “loop intended for operation in a verti- cal hole in fhe LITR has been charged with fuel and is now in the final stages of assembly for the in-pile test, Data which were obtained for three different sizes of charcoal traps to delay fission gases from the small loop in the event of rupture were positively correlated. Two charcoal traps have been incorporated in the cooling-air off-gas line, - A tube-burst stress-corrosion apparatus has been assembled and awaits filling prior to irradiation in the LITR, and an LITR-irradiated stress-cor- rosion rig is ready to be examined in the hot cells. The creep apparatus installed in the MTR has just completed six weeks of irradiation and is ready to be returned to ORNL for postirradiation measurements. An equation describing the behavior of the xenon poisoning in a fluid-fueled reactor was derived and was applied to the ART design. The calcu- lations indicate that the removal of xenon by sparging with helium will be a satisfactory means of controlling xenon poisoning in the ART. Also, no difficulties appear to exist in connection with shutdown poisoning at the sparging rates selected for ART operation. 9. Analytical Chemistry of Reactor Materials Apparatus was calibrated for the determination of vranium metal in fluoride salts by the method in which the metal is converted to UH, and subse- quently ignited in oxygen at 400°C to form water and UQ,. Samples of UF3 and KF-UF3 were then analyzed. The coefficient of variation was 7% for the range 6 to 60 mg of uranium. No interference from the presence of either fluoride salts or tri- valent uranium was encountered, and therefore the procedure should be applicable to all types of fluoride fuels. A comparison of the methylene-blue and hydro- gen-evofution methods for the determination of tri- valent uranium in LiF-Ber, NaF-LiF, and NaF- BeF, base fuels was made, Satisfactory agreement of the methods was observed for these materials; however, the results for trivalent uranium in a KF base obtained by the methylene-blue method showed negative bias when compared with those from the hydrogen-evolution method. Methylene- blue solutions which were 1.5 to 6 M with respect to HCl were shown to be reduced to methylene white at room temperature by finely divided metallic chromium, iron, nickel, and uranium-nickel alloy. Studies were continued on the simultaneous o 1) lron, chromlum, erCO i determination of frivalent uranium and total uranium in fluoride salts. The postulation of an interaction species of pentavalent uranium and methylene white was investigated. By using an anion-exchange resin in the hy- droxide form to retain zirconium, beryilium, uro- nium, and sulfate ions, the quantitative separation of alkali metal ions was rapidly effected. Determi- nation of the alkali metal concentration was made by titration of the free base which results from the anion resin exchange. When more than one alkali metal was present, the 2-ethyl-1-hexanol procedure was applied to the determination of lithium, and the tetraphenyl boron method was applied to the determination of potassium. Sodium was determined by difference methods. A rapid volumetric method for the determination of zirconium in fluoride salts was proposed. In this method, which is a modification of the method of Fritz and Johnson, a standard solution of disodium dihydrogen ethylenediaminetetraacetate is used to complex zirconium, and the excess re- agent is back-titrated with iron(lH), with disodium- 1,2-dihydroxybenzene-3,5- d:sulfonm‘e as the indi- cator. The bromination method for the defermin'afion, of oxygen as oxide was applied to samples of CrF, and Na,ZrF . Further tests were made on the electrolysis method for this determination. A modification of the Winkler method for the determi- nation of oxygen in water was applied to determine the oxygen in the off-gases from the electrolysis. The oxygen is absorbed in a solution of Mn(OH), and Kl, which, upon acidification, liberates a quantity of iodine equivalent to that of oxygen. The lodlne is determined spec’rrophotometrlcally by ex'rracflng it w:fh orthoxylene, - An mves’rlgcmon was made of the de’rermmahon ' of trace quantities of nickel in fluoride by the use -~ of the reogent sodium’ dlethyldlth:ocarbomafe. The " present experiments ‘were not successful because ‘Vr.lf“’_:O'F the interference of uranium and | iron. Absorbonce,_ ' specfro of the carbon tetrachloride extracts were determined for aqueous solutions containing sodium diethyldithiocarbamate cmd such cations as nlckel m, g: d urcxmum 10 | Rficovery and Reprocessmg‘of Reucfor Fuel“ - The feasniblllfy of ‘repeated use ¢ of a \nxlcke| S action vessel in the fluoride volatility—fused salt process for recovery of ARE-type fuel was demon- PERIOD ENDING JUNE 10, 1955 strated in 20 laboratory-scale runs. Nickel test coupons held in the nickel reaction vessel during the 20 fluorination runs showed corrosion of the solution type that was even over all surfaces, including welds, in contact with the molten salt, Severe local pitting was noted that varied in depth up to 19 mils on the fluorine gas inlet tube in the vapor zone above the molten salt. The attack on this tube in the liquid zone was more uniform, and varied from 4 to 7.5 mils in depth. The reaction vessel showed nonuniform attack of the solution type that varied from 5 to 9 mils in both the liquid and gas zones. At either 200 or 650°C, CaF, was much less efficient than NaF at 650°C in removing volatilized ruthenium from the 'UF6-F2 gas. However, results of runs made under various conditions indicated that the temperature, size, and conditioning of the NaF bed are very important, The engineering flowsheet for the ARE fuel recovery pilot plant is 85% complete. Design of 10 of the anticipated 29 process equipment pieces is complete, and the pieces have been ordered. PART IH. ‘SHIELDING RESEARCH 11. Shielding Analysis A semianalytical Monte Carlo calculation has been initiated to determine the history of all gamma radiation born within a circulating-fuel reactor. All shells of the core, reflector, and shield will be taken into account, but the calculation will be simplified by the assumptions that the reactor has spherically symmetric geometry and that all regions are homogeneous. The results will include de- terminations of the energy absorbed and of the energy spectrum and angular distribution of the . gamma rays penetrating the shield. The coding of a Monte Carlo calculation of the energy absorption resulting from gamma radiation ... incident on a multiregion shield with slab geometry _is nearing completion. The original code has been -.revised, and the calculation should now provide good statistics for gamma rays incident on a shield with a thickness of approximately seven mean free paths, ~ The codlng of a Monte Carlo calculation of the energy and angular distribution of air-scattered . .neutrons from a monoenergetic source is also nearing complehon Modifications in the original problem moke it possible to perform calculations for surface sources having angular variation in B g SR ANP PROJECT PROGRESS REPORT strength propomoncl to various powers of the cosine of the angle to the normal and then by a sun’rable combmcmon to duplicate the distribution ~ from the surface of a sphere such as a circulating- ~ fuel reactor. The analysis of the constant-velocity transport ‘equation has been extended with the aid of eigen- functlons of fhe various media. 12 le Tonk Shleldlng Facility The smtlc source tests in the second series of experlments “with mockups of a cireulating-fuel " reflector-moderated reactor and shield are in progress. The mockups consisted of the following regions: Inconel core shell, beryllium reflector, first boron curtain, heat exchanger, second boron curtain, pressure shell, and a lead-water shield. Gamma-ray dose rate measurements beyond mock- ups with variations of these regions have shown that (1) / -in. thickness of boral for the first - eurtain decreases the dose as much as a 1% -in.- thick boral first curtain would; (2) the dose is decreased with an increase in heat exchanger thickness; (3) there is appreciable gamma-ray production in and beyond large uniform thicknesses of lead in the mockup; (4) only an attenuation effect is observed when pressure shell material (that is, nickel) is added; and (5) the dose is increased 10% by the addition of 0.125-in.-thick Inconel cladding on the first boron curtain. A measurement of the distribution of the thermal- neutron flux within the 12-in.-thick beryllium re- flector showed that the flux peaked at about 41/4 in. of beryllium. Neutron measurements as a function of the lead region thickness revealed that ~an increase in lead thickness effectively only moved the thermal and epithermal flux outward from the source. A study of the activation within the heat 'exc}-'ucnger was also carried out, and, on the basns of this study, a calculation for a 300-Mw cnrplcne reactor indicated that a 2000-curie acti- vation is fo be expected in the NaK. For an un- shielded NaK-to-air radiator this would give 16 rem/hr at 60 fl'. 13. Bulk Shleldlng Facility A mockup of the swept-bock air-duct system for ’rhe GE ANP reactor was tested at the Bulk Shield- ing Fccuhfy (BSF). The mockup consisted of a “pair of annular ducts that-represented a segment of the duct system around the airplane reactor. A 10 long inlet duct and a shorter outlet duct were placed against, and on opposite sides of, the BSF reactor, which was modified with GE-type transition sections. Both ducts curved to the right, and the shorter one nested within the longer one so that they ran roughly paorallel. Measurements of the radiation ledking out the end of the large duct as the distance between the ends of the ducts was varied (that is, as the large duct was moved out) showed no evidence of streaming down the entire length of the large duct. Fast-neutron dose rate measurements beyond the shield and parallel to the last leg of the large duct were not affected by the position of the duct. Angular distribution measurements around a fixed point beyond the ducts showed no evidence of fast neutrons escaping from any portion of the large duct. 14. Tower Shielding Facility Differential-type shielding experiments have continued at the Tower Shielding Facility (TSF) with emphasis on measurements of the fast-neutron dose rate distribution within the detector tank and the GE-ANP crew compartment held at a separation distance of 64 ft from the reactor tank. This distribution was determined as a function of vari- ations in the reactor shield thickness, crew com- partment thickness, and angle of radiation emission from the reactor tank. The measurements, which were made at a height of 195 ft, are presented in this report. A method is presented for interpretation of the TSF differential shielding experiments in terms of the probability of fast neutrons scattering into the sides of a crew compartment of an aircraft divided shield. The effect of beam collimation on the experimental results, as indicated by single air- scattering calculations, is discussed. An indi- cation is obtained from the experiments of the effect of the neutron energy spectrum on the dose- scattering probability. The results and interpretation of the TSF dif- ferential shielding experiments were used to calculate the fast-neutron dose rate for a divided- shield mockup which had been measured previously at the TSF. The agreement between mockup experimental results and the calculations based on the differential experiments was good, in view of the various uncertainties existing at this time. In moking the comparison, procedures were de- veloped for using the experimental results in dose fay 3.4 i) v 2y predictions for aircraft divided shields. Some experimental information is presented on the re- lation between dose within a cavity and the source PERIOD ENDING JUNE 10, 1955 strength at the surface of the cavity, where the source strength is taken as the dose rate at the position on the surface with no cavity present. 11 P Part | REACTOR THEORY, COMPONENT DEVELOPMENT, AND CONSTRUCTION [ M] ¢y "'f_;'con stitute 1. REFLECTOR-MODERATED REACTOR E. S. Bettis W. K. Ergen A. P. Fraas A. M. Perry Aircraft Reactor Engineering Division AIRCRAFT REACTOR TEST DESIGN A, P, Fraas Aircraft Reactor Engineering Division The preliminary layout of the interior of the reac'r'or cell for the Aircraft Reactor Test (ART) which shows the major items of equipment and the recommended provisions for support was completed. A similar layout for the NaK piping and radiators was also completed. [t has been established that five bulkheads will be required in the tank and cell walls for instrument and control wiring, piping, and miscellaneous service lines. Consideration has been given to the design of the addition to Building 7503 required for the ART, the layout of the blower house, and the electrical power system, distribution, and auxiliary equipment. Mounting arrangements have been described for the radiators, for the fill and drain tanks, for the inner and outer reactor cells, and for structurally attaching the reactor assembly to the cell, dump lines, NaK piping, and other service and instrumentation lines. The type of equipment by which oxygen will be removed from the inner cell has been specified. The basic requirements of the heat dump system and the instrumentation and control thereof have been specified. The ART heat dump system is to provide heat dissipation capacity of 60 Mw of heat with o mean temperature level of ]300°F in the NaK system The most convenient, inexpensive, and compcct heat dump has been found to be a round-tube, p!ate-fm radiafor core. This basic type of heat transfer surface has been found fo be’ '_sufficnently rehable in hect exchanger test rigs. - Five separate systems will be used; four will »;mcnn heat dump system, w['ule fhe“ Cfifth will be the modera'ror heat dump system. Layouts have been prepared of the air duct and stack associated with the heat dump radiators, and air flow ra'res qnd pressure drop hcve been' o _"-:cc:lculoted o e o « .~ Design ; cdhs'ldefd}“io'hs hove been tem‘utlvely* ) -‘-;estcblrshed for the prevention ‘of failure of the “Inconel core and reflector shells durmg operahon": of the ART. The core shells are to be /8 in.-thick Inconel. They are to be cooled on one side with flowing sodium in order to maintain the shells everywhere below 1500°F, The reflector shell is to be maintained below 1300°F by surrounding it with a layer of boron-bearing material, a gas space to serve as a heat dam, and an Inconel cladding layer to separate it from the hot fuel in the heat exchanger. To prevent buckling of the outer core shell because of external radial pres- sure differences and to prevent excessive de- formation of the reflector shell because of ex- cessive pressure differences, the pressures in the sodium coolant system and in the fuel are to be adjusted to be approximately equal at a point halfway through the primary heat exchanger. In order to limit cyclic thermal stresses, the tempera- ture difference across the shells is not to exceed 300°F. Stress and heat transfer analyses are far from complete, but tentative pressures have been established for the fluid circuits. In case of a failure of apart of the reactor system, for instance a fuel pump, during operation, it may be necessary to maintain emergency con- ditions for short periods in which the pressure differences across the shells will be much greater than those existing during normal coeration. Calcu- lations indicate that these conditions can be maintained without sudden failure of the shelis by buckling or rupture. However, the total life of the reactor for normal operation will be dispro- portionately shortened. Details of the basic design of the reflector- moderator cooling system have been set forth. Final design awaits confirmation of the values - _used in the design by critical experiments, heat . exchanger optimization, core hydrodynamic tests, and beryllium thermal stress and corrosion tests now in progress (cf. Secs. 2, 3, and 7). The xenon-removal system has been designed, as described previously,! and the fuel and sodium pumps are being tested (cf. Sec. 2, “*Experimental - ~Reactor Engineering’’). New features that have been incorporated because of stress, fluid flow, ... ot fabricability considerations include an elliptical 1G. Samuels and W, Lowen, ANP Quar, Prog. Rep. Dec. 10, 1954, ORNL-1816, p 21. 15 ANP PROJECT PROGRESS REPORT fuel expansion tank, a rounded dome to enclose the top of the reactor, a newly designed sodium pump impeller, and other related items. A number of fuel-to-NaK heat exchanger designs have been calculated by utilizing the hecn‘ transfer data obtained experimentally (cf. Sec. 2, ““Experi- ‘mental Reactor Engineering,”’ and Sec. 7, *“‘Heat Transfer and_Physmal Properties’’). The most promising of these designs are described in Table 1.1. The values given in the table are intended 'fdrfd'éilifdte comparison of typical heat 'e:xlo‘:hahgers. The values listed for heat exchanger fhlckness and shield weight are only relative and were based on a mmtmum-fhlckness tube array svited to the present design layouts. Final selechon of fhe heat "exchanger must be con- tingent on a composnte evcluchon of stress limi- tations, geomefry, mass transfer and corrosion - effects, and heat transfer requirements. Conservchve colculc’rlons were made of the neufron ‘and gcmma ray dose rates at various locations in the ART facility. |t was determined that radiation levels in the building during power operation would everywhere be below laboratory tolerance. Further, with the exception of a few locations, the initial dose rate from a reactor catastrophe would not be more than a few roentgens per hour. These low dose rates are the result of the reactor being located 16 ft underground and shielded (in addition to the lead-water reactor shield) by 3 ft of water in the cell annulus and substantial thicknesses of concrete and dirt. The radiators are to be shielded so that, in the event that 10% of the fuel entered the radiators, the radiation dose would not exceed 1 r/hr at any point outside the main air duct or reactor cell. Research is under way that will provide much of the information needed as a basis for final design. In the experiments that are now being made with room-temperature critical assemblies of the reflector-moderated reactor, materials of interest in the design of the ART, such as beryl- lium, Inconel, and rare-earth oxides, are being evaluated. The information being obtained will be of benefit in making final decisions on reflector- moderator and island dimensions, configuration, and cooling requirements. Data on power distri- butions in the core will be used in determining the final configuration and size of the core and the end ducts, as well as the thickness of the Inconel shells and the cooling required. Measurements being made on rare-earth oxides are expected to provide data needed for design of the control rod for the ART. Intensive experimentation is in progress in at- tempts to find fuels superior to the ZrF ,-bearing fuel used in the ARE. As yet, no positively su- perior fuel mixtures have been obtained, but the physical properties of the ZrF,-bearing mixture have been improved to some extent by varying the TABLE 1.1. ART HEAT EXCHANGER DESIGNS Tube diameter, in. 3/16 Number of tubes per bundle 143 Tube spacing, in. 0.030 Tube wall thickness, in. : 0.025 Fuel temperature range, °F NaK temperature range, °F Tube length, ft 6.0 Fuel AP, psi 41 NaK AP, psi 39.6 Heat éxchcngér rhickhess, in. o 2.62 ' Reoctor shleld Welg['\‘l' water plus lead, Ib 72,800 "V:-;,‘-Tota| number of we!ds 6864 ’ P?wer, Mw ' 55 Relat.iv.e Nq-i-( ocfi.vofion ” 1 16 1 1250 to 1600 1070 to 1500 %6 ! ' 143 130 130 0.030 0.020 ' 0.020 0.035 0.025 0.035 | 1250 10 1600 1250 to 1600 1250 to 1600 ° 1070 to 1500 1070 to 1500 1070 10 1500 6.5 5.45 5.81 45 28.4 30.1 95 5.5 9.6 2.62 3.78 3.78 72,800 74,100 74,100 6864 6240 6240 55 55 55 1 2 2 v composition, and the corrosiveness has been lowered by the addition of even small amounts of trivalent uranium. The operation of forced-circulation, high-tempera- ture-differential, Inconel loops with fuel mixtures of interest has demonstrated that the velocity of the circulated fluid has little effect on corrosion and mass transfer and that the results of the numerous thermal-convection loop tests are appli- cable to the dynamic ART system. One in-pile forced-circulation Inconel loop that circulated a fuel mixture in the LITR has been examined, and radiation was found not to have a detrimental effect on corrosion and mass transfer; a second loop will be operated in the LITR during the next quarter. An in-pile loop that will more closely simulate the ART conditions is to be operated soon in the MTR. Experiments with sodium in Inconel loops have thus far indicated wunsatisfactorily high mass transfer, but it is thought that better purification and handling of the sodium will greatly decrease the mass transfer; alsc, the use of stainless steel in the cold portions of the system is being in- vestigated. Cermets for use as valves, bearings, and seals are undergoing intensive self-bonding and cor- rosion resistance tests, and procedures for brazing cermets to Inconel have been developed. Studies of expenmentcl nickel-molybdenum base alloys and commercial Hastelloy B, in particular, have indicated possibilities of finding a structural material with properties superior to those of Inconel for c:'ircula‘ring fuel reactor dpplicafion, PERIOD ENDING JUNE 10, 1955 Shielding Facility will provide much of the data needed for final design of the reactor and, in particular, the shielding. A mockup of the reactor and shield is being used to determine activations within the heat exchanger and the effect on dose rates at various locations of varying materials and thicknesses of materials. The activation of the NaK within the heat exchanger is of particular importance because of the effect it will have on dose rates at the NaK-to-air radiators, which are outside the reactor shield, Design of the engineering flow sheet of a fuel recovery pilot plant is 85% complete. It is hoped that the rapid reprocessing and recovery of fuel that would be reguired under service conditions can be demonstrated with this pilot plant. The many calibration and performance tests scheduled for the ART are summarizedin Table 1.2, Some of the tests can be run in the course of the endurance tests, which should consist first of 25 simulated flight cycles (16 hr at full power and 8 hr at from 1 to 10% power) and then of 100 hr of continuous operation at full power. This will give a total of 500 hr at full power and 200 hr at low power during the endurance test period. It is expected that the preliminary calibration and performance tests can be carried out during the first 400 hr of operation, with allowances for servicing and maintenance. In reviewing the design requirements and the tests planned for the ART, it is evident that many modifications could be made in the test conditions to vary the severity of the test. The key variables, together with the desired values for the test, are given in Table 1.3. but some of the probfems involved are of such , a mqgmtude “that it is ot thoughf 'rhaf such ma- terials can be made avmiable in time for use in the ART. The weldlng and brozmg matermls and techniques needed for fabrication of the ART have "«'_rffbeen developed qnd thoroughly‘ 'rested |n the con-“";&"w A study of The. behowo_ : nonp on ng n a fluid- fueled reactor has |nd|ca’red ’rhat fhe?:_: removul of _xenon by sporglng “with helium will 7 sfactory means for controlling xenon po s‘en g in 'rhe_"ART The sfudy alsdhmdncated\it " that no dlfflcul’ry wou[d exist in ‘connection with ~*“shutdown poisoning af the sparging rates selected for ART operation. Experiments now under way at the Lid Tank ART CONTROL J. M. Eastman Bendix Products Division El Ro Mdhn Ins’rrumentatlon and Confrols DIVISIon rJflm‘orma'uonfi"l:)lock dmg"“ms showmg the basic tfi-c'ontro! ac, |<'>hs'(de5tred (w:thouf reference to hard- ware or fechnlques) have’ been prepared for the ART They will be used as a basis for selection of ’rhe hardware types ‘and the control techniques to be used cnd for determmahon of areas of control ’rhuf w:ll reqmre new COmponent developmenf.” Information for reactor simulation has been supplied to Pratt & Whitney, and they have built a simulator, It will be used to check the adequacy of the basic 17 i ANP PROJECT PROGRESS REPORT TABLE 1.2. NUCLEAR EXPERIMENTS ON THE ART Zero-Power Experiments Criticdli?y ' Demonstrcte procedure for going critical o Determme crmcal mass Reochvn‘y Experlmem‘s Determine fuel temperature coefficient - Determine reflector and island temperature coefficient Defermme mass reachvnty coefficient Measure reachvn‘y as a function of flow rate ' Defermme delqyed neutron loss to xenon purge system Shle]dmg Survey surface of shield for radiation .. Determine NcK achvahon in heof exchanger High-Powe: Experiments Demonstrate That the Réa_qctor Is a Slave to the Load Demand - Simulate s'ubdrdén.d;ér'r‘achld for increased power by turbojets " Simulate failure of one turbojet Determine Reactivity Effects of Other Transients " Results of sudden flow stoppage Results of NaK flow stoppage Determine Effectiveness of Xenon Removal System Determine Compensation Required for Fission-Product Poisoning and Burnup Demonstrate Afterheat Removal upon Shutdown Obtain Heat Balances at Various Powers to Determine Extracted Powervs Nuclear Power for the Core, Heat Exchanger, Island, Reflector, Pressure Shell, Gamma Shield, and Neutron Shield control actions and to determine control component design data, Additional simulator work will also be done at ORNL. Control Principles The control system is to provide automatic cor- rective action for emergencies requiring action too rapid to permit operator deliberation. Automatic interlocks will prevent inadvertent dangerous oper- ation, with minimum operator limitation. Operation in the design power range will be independent of nuclear mstrumenfc’rlon and will have only limited dependence (for safety) on other instrumentation during' power transients. Three classes of emer- gencies will be provided for: 1. those requiring automatic rod insertion, load " removal, and fuel dump, ' ,'2. those requmng only automatic rod insertion and 18 load removal (followed by manual dump at oper- ator’s discretion), ' 3. those requiring only operator warning and manual action. One-half the wind tunnel blowers and one-half the pumps (of each type) w:ll be pOWered by com- mercial supply. The remaining blowers ‘and pumps will be powered by a deisel system. Duplicate fuel and sodium pumps have been provided for safety, but if any one of these four pumps failed, the output of each of the remaining fhree pumps would have to be decreased to prevent pump cavi- tation or pressure overstressing of the Inconel core shells in the current design, The basic control actions provide for this depression of output, but it is questionable whether pump speeds can be de- creased fast enough to prevent adverse transient conditions. For example, the Inconel may be over- ) TABLE 1.3. KEY ART PERFORMANCE VARIABLES Variable Desired Performancg Time at full power | 500 hr Total time critical 1100 hr Total time }hgrmany hot 1500 hr Power level 60 Mw Peak fuel temperature 1600° F Peok metal temperature 1540°F Peak NaK temperature 1500°F Temperature difference 100°F between fuel and NaK Number of power cycles 30 Rate of change 7°F/sec of mean fuel temperature Number of dump cycles 5 stressed momentarily, with a consequent increase in creep rate. To prove at least that the reactor can be safely scrammed after such a pump failure, it is planned to include cutting off the commercial power (or diesel) at design point (60 Mw) as a part of the test procedure. This will constitute a class ] emergency, as defined above, The control rod cannot be abruptly inserted in the case of an emergency, because to do so and yet prevent the fuel from freezing or getting too hot would require accurate transient matching of ab- sorbed power and flux power, which would be im- practical to accomplish. Therefore a scram (class 1 or class 2 emergency) will abruptly insert the rod “only enough to reduce Ak/k by 1%. This will be followed by automatic insertion at the rate of 1% Ak/k in 30 sec. This Speed ws“ be the ‘‘fast’’ 'normai rcn‘e of rod movement. The opemtor mcy” _-a!so choose ‘ slow” rod movemenf of 1% A]e/k in 5 min, Below 10-Mw power operahon, the opercn‘or will be automchca”y limited to the slower rate. The 1% scram will automatically cut off one-half ‘the blowers to reduce the load. The load will then be further reduced by automatic closing of the _radiator, shu’r’rer in response fo a 1050°F low limit signal for the NQK -to-air radiator outlet temperature. Since the de5|gn point occurs at the hlgh temper- ature limit (1600°F) for the fuel, accurate limiting will be required, |n addition to thermocouples in the fuel, the maximum fuel temperature will be PERIOD ENDING JUNE 10, 1953 indicated by continuous computing from NaK thermocouple signals by using heat exchanger calibration data, The dependability of fuel-temper- ature-sensing thermocouples is questionable be- cause of the poor conductivity of the fuel and the high gamma-ray heating. ' Operating Procedure The reactor is to be initially filled with barren (without uranium) fuel carrier to check out the system. This will be done with the system iso- thermal at 1200°F and all NaK and sodium pumps operating., Electric heaters will supply the heat needed beyond that produced by pump work dis- sipating, After the checkout of the system, part of the barren carrier will be removed and replaced with fuel-enriched material to provide 80% of the calculated critical amount of U235, Criticality will be checked, and the fuel will be further en- riched in steps until criticality is obtained. The control rod will be calibrated in the process, A flux servo will then be available for holding the reactor at very low power levels (under 10 kw) during low-power experimentation. At these levels the temperature coefficient will be inadequate for good control because of the high thermal capacity of the system. The minimum power level for control by the temperature coefficient is expected to be about 300 kw ~ the estimated heat removal by the radia- tors with the heat barrier doors open and no air flow. To take the reactor to the 300-kw level, it will manually be put on not less than a 20-sec positive period, and the barrier doors will be opened when the flux level reaches a set value equivalent to something less than 300 kw. |[f the doors were opened too soon or too late, excessive power surges might occur before the flux stabilized ‘and there might possibly be damaging thermal shocks, After stabilization at 300 kw, the control rod will be slowly adjusted to bring the mean fuel temperature to 1200°F (it will have gone somewhat " above this). To increase the power above 300 kw, the blowers will be turned on and the shutters opened as de- sired. The control rod will be used to adjust the "mean fuel temperature. Automatic limiters will close the shutters if the minimum NaK temperature drops below about T050°F (to prevent fuel freezing) and will open them if the minimum NaK temperature goes above 1300°F (to dissipate afterheat and pre- 19 ~ and femperature surges. ANP PROJECT PROGRESS REPORT vent loss of ZrF,). To take out more than 20 Mw will require that the mean fuel temperature be set above 1200°F (it must be set at 1425°F to take out 60 Mw). Below about 6 to 10 Mw, the heat dump will have to be adjusted slowly to avoid high flux ' Above this power the temperature coefficient is expected to permit reasonably fast (for jet engines) load changes without excessive transient surges. VREACTO R PHYSICS Probable Effect of Répldcihg Inconel by Columblum in fhe ART Core Shells Ww. K Ergen Aircraft Reactor Engineering Division The cold critical experiments for the ART have shown? that the critical mass decreases very rapidly as the neutron absorption in the core shell decreases. Hence, replacement of the highly ab- sorbing Inconel by a low-cross-section material, such as c:olumblum,3 would bring about a sub- stantial saving in vuranium investment and, inci- dentally, make the temperature coefficient more negative. For corrosion and fabrication reasons, only about 50% of the Inconel can be replaced by columbium. The effect of this 50% replacement has been calculated on the basis of recent data from the cold critical experiments and measurements of resonance capture integrals.? From the critical experiments it is concluded that the epicadmium flux is about the same for the core shell and for the fuel. From the capture ‘resonance integrals of Inconel and columbium, respectively, and the fission resonance integral of U235, it is computed, for the epicadmium range, that the columbium absorptions amount to 1.6% of the fissions, whereas the absorptions in the re- placed inconel amounted to 2.4%. For the ‘‘below-cadmium’’ neutrons the cold ~ critical experiments are consistent with the model that the shells absorb neutrons according to the thermal cross sections and that all neutrons pass through the shells perpendicularly. By using the 'l/u extrapolation to get the 1400°F obsorpfion A D. Callihan et al., ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 43. 3C B. Mills and H. Reece, Jr., Design Study of an ANP Circulating Fuel Reactor Nov. 30, 1954, WAD- 1930, p 44. —4R. L. Macklin and H. S. Pomerance, Resonance Capture Integrals (to be published). 20 cross sections, the columbium absorbs 0.5% of the neutrons, whereas the replaced |nconel absorbs 3%. The experimental result that the epicadmium neutron group and the subcadmium neutron group each contribute about 50% of the fissions then indicates that effectively 1,65% of the neutrons would be saved by the replacement, If (AE/R)/(AM/M) = 1/10 is assumed, the saving in critical mass would be about 16.5%, or 4 kg, corresponding to 0.7 mole % in uranium concentration of the fuel and 7 kg in total U235 investment. As a check, the above method can be used to compute the effect of completely eliminating, in the cold critical assembly, I/] in. of the core shell. The result would be a 41% saving in critical mass, compared with the experimental value of 45%., In another room-temperature critical experiment (cf, Sec. 3, ‘‘Critical Experiments’’), 127.4 g of the Inconel core shell was replaced by 98.5 g of columbium. The effect calculated by the above method would have been 2.5 cents, and the experi- mental result was 3.6 cents, ART Temperature Coefficient L. T. Anderson Aircraft Reactor Engineering Division A calculation of the ART temperature coefficient showed a positive contribution resulting from the decrease of the absorphon cross section of the Inconel core shells3 with increasing neutron energy. The smaller Inconel cross section allows more neutrons to enter the fuel annulus and thus increase the reactivity, It seems justified toassumeal/v cross section for the Inconel on the basis of the experimentally measured value of the resonance integral, and therefore, if an increase, AT, of the beryllium temperature causes a neutron energy increase proportional to AT, the fractional cross- section decrease of the shells is l/2 AT/T, and the number of fissions is increased by the factor 1/ (AT/T)Z,, x shell fh:ckness X cadmlum frachon if only the subcadmium neutrons are cn‘fected and the neutrons are incident normal to the shell, Actually, the thickness of the Inconel that the the neutrons go through is somewhat greater than that of the shell, on the average, but this is at least partially offset by the cross section of v ~cases. B '_'\.fculchon Inconel having a less-than —2KR, [T AT 7 R, (R, = Ry) _RZJ; where K is the coefflaent of Ilneqr ‘thermal ex- pansmn for fhe bery”:um and 7 is the fhermai age. By using K = 93x ]0 6/°F 7 =90 cm?, R, R, =25 ~cm, ‘and R2 56 cm, |'r is found thai Ak/k = —] 2 x ' ']O SAT . i The net of these ’rwo contrlbuhons is then 0. 5 >< 10-5/°F for. 1/8-m. Inconel core “shelis and Zo2x T 10‘5 for core_shells of /16""' [nconel and ]/]6-|n. - :;_f‘f'coltmelum. | f._'f,_:r_.;whlch is based ona conset "ctlv‘eya?ye “of 0.1 for'l o 'efflment of -1. 7 X ]0" /°F, S, VAnderson, Calculation of the Beryllium Contri- bution to the ART Temperature Coefficient of Reactivity, ORNL CF-55-5-76 (May 11, 1955). 6W. D. Manly, private communication, PERIOD ENDING JUNE 10, 1953 less, the results support the postulated negative over-all temperature coefficient of the ART, Reactivity Effect of a Heat Exchanger Leak A. M. Perry Electronuclear Research Division If the pressure in the NaK coolant circuit ex- ceeds that in the fuel circuit, a leak in the inter- mediate heat exchanger may first become apparent through its effect on the reactivity of the ART. The effect will be due to dilution of the fuel and to increased parasitic absorption. It has been suggested by the Advisory Committee on Reactor Safeguards that the reactivity effect can be greatly enhanced by adding lithium or Lié to the NaK, Preliminary experiments® indicate that perhaps 5 wt % lithium can be added to the NaK eutectic without raising the melting point above 25°C, The loss of reactivity due to parasitic absorption is calculated on the assumption that e M= iy T where / equals (neutrons absorbed in fuel)/(neu- trons absorbed in core), AX, is the change in poison cross section, and 2. is the fuel cross section. The loss of reachvn‘y due to fuel dilution is calculated on the assumption that — TET cmt—— — N —— i where V is the initial fuel volume, and AV the volume of NaK leaked into the fuel circuit, The factor 1/]0 is a conservativé estimate based on ~eritical experiment results. . The over-all reactivity loss, if AV is in cubic feet of coolant leak into the fuel circuit, is given by Ak T = ~2.4 AV(%) , | nc ‘rhe 'crorelic}lnt is Nak; by Ak S —(2.4 + 2.2 W) AV(%) ., the coolan;]s NCIK plus 'W wt % h'rhlum, and by Ak ”7;“- = (2.4 + 29.6 W) AV(%) , 2] o chonges ANP PROJECT PROGRESS REPORT lf 'rhe coolam‘ is NaK plus W wt % Lié (Of course, the opproxumahons made are valid only for small in reactivity and for small percentage “additions of lithium to the NaK.) The maximum volume available before an over. flow would occur will be 0.37 #3, This volume of NaK in the fuel circuit would produce a reactivity change of 0.9%. The same volume of NaK plus 0 5 wt % lithium Would produce a reactivity change of 5%. A leak of 0,007 ft* of NaK plus 5 wt % ' Ll would produce a reactivity loss of 1%, | Burnup and Gomma-Ray Heating of Control Rod - W. Fader Proh‘ & Whitney Alrcroff Burnup of Poison. ~ a slab of neutron absorber with thermal neutrons incident on one face are, for the case of an ab- sorber with n isotopes with high thermal capture ~ cross sections o, Z E + 2(;5 = 0 ’ d2; '—gg— + aiziqb =0, where ] = net neutron current normal to slab face, linear coordinate normal to the slab, H %, + 2, + ... + £ =macroscopic thermal neutron absorption cross section, flux, ' exposure time, I i x 2 é t By making use of the relationship | = ¢, where I is the average value of the cosine for the angular distribution of the neutron flux, the equations be- come (1 .Uax Z],— 0, azz. (2) I +o2] =0 . ot ~ In the region where most of the burnup takes place, 11 is constant if the macroscopic absorption cross - section is much greater than the macroscopic scat- | fermg cross section, lts value may be expected to range between unity at points deep in the slab ~to 0.5 at the surface, the latter value being the diffusion theory approximation to the boundary The equations of burnup in condition at the plane interface of a scattering medium and a strong absorber, Solutions for J(x,) and X(x,t) were found for the case of a single absorber and the results have been plotted in Fig. 1.1 for a slab of Sm,0,. For large values of o] /i, where J, = ](0 t), the curves of J(x,t) ond 2(x,t) assume rigid shapes that are propagated in the direction of increasing x with apparent velocity 0’]0 2 I 0 which, for the case of the curves in Fig. 1. ]. was v = 2,9 x 108 em/sec. This suggests the use of the linear formula (3) X() T 0.012 + (29 x 1022 [ 1) em for the thickness of a slab of Sm,0, that will transmit 10% of J neutrons/cm? osec mcudenf on its front surface after an exposure time of t sec, Equations 1 and 2 were integrated numerically for the case of gadolinium with two isotopes with high thermal absorption cross sections, and the results are shown in Fig, 1.2 for Gd ;03 The linear relohon between X and ¢ for large cr] ot/ for Gd O (4) X() = 0.0025 + (1.4 x 10=22] ) cm . Formulas 3 and 4 may also be used as an ap- proximation of the burnup of a cylindrical shell of absorber, provided that the inner radius of the shell is much larger than the thermal neutron mean free path of the absorber; for, in this case, Eq. 2 remains unchanged, while Eq. 1 becomes d (3) i 'éi = 3] - IT”‘!‘ r r for neutrons incident on the outer surface. For a shell of this kind, the ferm T]/7 in Eq. 5 is always negligible compared with %] in the region of burn- up. In the ART, a 20-in. control rod with a ]?’/B-in. outside diameter must absorb an average of at least 2 x 1014 neutrons/cm?:sec in order to be 5% effective in reactivity. Thus, if a cylindrical shell of Sm,0, absorbs 2 x 1014 neutrons/cm? sec for 1000 hr, the neutrons will penetrate a distance X(t) = 0,012 + (2.9 x 10=8) (7.2 x 106) = 0.22 cm. A shell of szo3 must have at least this thickness e it e e B . it i s . PERIOD ENDING JUNE 10, 1955 -CONPIERITAT ORNL-LR-DWG 7551 N ™ T~ % JF— 14eBOx 0y Vg _madyifp ] N ~ \ #=0 \,«f_=4,5 % 10° sec \v/f 9 x 10 sec Jo = 10'% nsem? - sec \ A=t \ a=Yy \ E=l 020 ¢ o=58x/ 0.6 Sy // -~ // Z 7 2(0)=200 cm™f FOR Sm,05; — 0.4 \ \\ \\ \\?<fi=_?uossec a1 . zO />\\ \ \ N\ L / L N AT I S N N \ 0 S~ T ~_ ~ //7/ 0.8 // ,// / /Lf 4.5 x10% sec / f=94“05 sec / H= i/2 // /‘—"z / Z{O 0.6 - / 5 / / - i e(chof “TOIx}/E ez«»x/fi s / R 04 A “ / /Q\f 9 x 10° sec / / p=lp 0.2 / pd / / 0 0.005 0.010 0.015 0.020 0.025 0.030 0.035 0.040 0.045 0.050 0.055 %, DISTANCE FROM FACE OF SLAB (cm) Fig. 1.1. Burnup of a Samarium Oxide Slab. to remain black to neutrons after 1000 hr with Jo =2 x 10%4, For Gd,0,, X(t) = 0. 0025 + 1 4>< 10-8(72x 106) 0 10 em . To obtain an eshmcn‘e of fhe error of the approxn- mation, note that for a shell with an outside dlam- ' eter of ]38 in, = 1.76 ¢m and a length of 20 in, = 50.8 cm, an absorption rate of 2 x 104 nev- trons/cm%:sec at the outer surface will burn up 4,05 x 1023 u'roms or 100 g of Sm'49, This corre- . sponds fo 868 g of Sm, 03. ‘the volume of burnedfi up material would be 116 em3. A shell with an outside diameter of 1.76 c¢m, a length of 50.8 cm,'jf‘“ and a voiume of 116 cm3 has a wall thlckness of been calculated, and the results are presented in Fig. 1.3. Gamma-Ray Heating. Two effects are expected to contribute to the internal heating of control rods of rare-earth oxides: absorption of gamma radiation emitted immediately after capture of ~ neufrons in the rod and absorption of gamma rays from the reactor fuel. ‘Preliminary calculations of the gamma-ray heating ‘in the control rod indicate that the average energy absorption will be at least 100 w/em3, More de- tailed calculations made by Pratt & Whitney Air- ¢raft indicate that the gamma energy absorption may be as h|gh as 170 w/cm3, For the higher value the maximum interior temperature rise above V'rhe surface temperature is 460°C for a 1%.in,-OD hollow rod of rare-earth oxide with a 14 -in, wall, . - if the thermal conduchvni‘y is assumed to be 0.0048 0.22 em. The burnup ‘of Eu 0 ‘slabs has olso”‘ “cal/sec:°Cocm. If a cermet of rare-earth oxide and iron is used in place of the oxide, the maximum temperature rise is about 45°C, if a thermal con- ductivity of one-third that of iron, or 0.050 cal/ sec:°Cecm, is assumed, 23 bt i ouoniNoh < e ,,,,,, ANP PROJECT PROGRESS REPORT BONFIRLNTTIL ORNL-LR-OWG 7552 ; _ JO=10'4 nyeme - sec ; N ) . ‘ \ \ \ o =7 10-2° cm? E . \ N +=6.5 2107 sec \ \ o =16 x 10 20¢m?2 4 =45 107 sec \ B=1 2 3 , p=1 % \ %,(0)=253 cm™* A 2,(0)= 620 cm™' / / / | J Z ‘06 i W N -0 t=4.5x10° sec E=lr > 1=9x 105 Sec A= I/, o D ] '—-/r-—-—-—- & O \ JEE L AN o \\\ \\\\\ \\ N / / \_ % | 0.8 / _fi' 0.6 Y § ' | —+t=4.5 xiossec r=9x105 se¢ 3 m=1 fi=1 %, /2,0 / » AL / _ b / S y : Y +=4.5x10° sec e £=8 % 10% sec o4 ~ a=1 a=1 p ~ 3 C 0.4 / _ 0.2 // 1 = — ; 0 0,002 0.004 0.0086 0.008 0.010 0.012 C.014 0.016 0.018 0.020 x, DISTANCE FROM FACE OF SLAB (cm) PTT Fig. 1.2, kB-urnup of a Gadolinium Oxide Slab. 24 PERIOD ENDING JUNE 10, 1955 SOMFITEN T ORNL-LR-DWG 7664 0.90 ' _ ) - L i g =1x 10" n/em? - sec o 0.80 2,(0) =120¢em ™ (Eu'®") | _ \ —7=360 hr 2, (0= 5.2 et (Eu'5?) F=500hr t=0hr t=250hr 0.60 N\ 0.50 | o4 AN L NN Jxh % /| j ) 0.0 o 0008 T Cgoio” o 0.015 ' ©0.020 S B "y, "DISTANCE FROM FACE OF SLAB (cm) ' ' 3. B urnup qf. a Europiumfi(l):‘cji"é‘lé Slub. | P 4 > 1 25 ~“ANP PROJECT PROGRESS REPORT 2. EXPERIMENTAL REACTOR ENGINEERING H. W. Savage E. S. Bettis Aircraft Reactor Engineering Division ~ Design work on the in-pile loop for operation in the MTR has been completed, and the final loop -is being fabricated and assembled. A bench test of a working mockup is under way. Twenty-two high-velocity forced-circulation large-temperature- - differential loops were operated for investigating the corrosion and mass transfer of Inconel by fluoride fuel mixtures under dynamic conditions. Six similar loops were operated to test mass trans- fer in Inconel and stainless steel loops in which sodium is circulated. A full-scale model of the ART fuel pump was tested with water, and performance characteristics were obtained. A test stand for high-temperature tests has been designed. A test stand for inter- mediate heat exchangers (100-tube bundle) is being assembled, and tests are under way with a stand designed for testing small-scale (20-tube bundle) heat exchangers, Flow patterns are being studied in a full-scale model of the proposed 21-in. reactor core and en- trance header. Several modifications are to be tried in an attempt to prevent flow separation in the core. - A fhermol-cyclmg test was made on a sodium- Inconel-beryllium system, and apparatus for « third test is being assembled. A small-scale gas- fired heat source was operated successfully at a power output of 100 kw, and minor modifications were planned that will increase the capacity. IN-PILE LOOP COMPONENT DEVELOPMENT D. B. Trauger Aircraft Reactor Engineering Division Flux Measurements D. M. Haines Pratt & Whitney Aircraft - Flux measurements, as previously described,’ were carried out in the HB-3 beam hole of the MTR, Cobalt foils installed in various assemblies that “simulated the loop were irradiated for 3 hr at about 5 Mw. Plans to make measurements at full power - 'wei'e aba'ndbned because an excessively long ir- : ]D M. Humes, ANP Quar., Prog. Rep. Mar. 10, 1955, ORNL 1864, p 32. T radiation was required to override the effects of the minimum period in which the reactor could be brought to full power. Gold foils irradiated at the low power and again at full power in a vertical hole adjacent to HB-3 provided a means for extra- polating the data to full power. This measurement made it possible to evaluate the effect of materials and geometry of the in-pile loop on the flux. Values obtained from foils inside the fuel tube were consistent with other data taken from foils irradiated in Inconel tubes by using a rabbit facility for irradiation in hole HB-3. Other data on the depression of flux in Inconel tubes with a fuel mockup present were also used to esti- mate the flux for the MTR in-pile loop. The flux seen by the fuel in the in-pile loop is now expected to be 30% of the unperturbed value. The heater loop, or nose section, originally de- signed for a depression of 50%, has been modified to obtain more power, The present nose section consists of a 2]/2-turn helical coil shaped, in out- line, somewhat as a truncated cone, It is mounted with its axis parallel to the beam hole center line and with the small end forward. This will permit the nose to be placed in the most forward position possible by utilizing the concial end of the water jacket. The developed length of the coil is slightly over 3 ft. A total power generation of 24 kw is required for the design conditions: Reynolds number, 5,000; temperature differential, 200°F, The average power density in the nose section, on this basis, will be 0.7 kw/cm?; however, this value may be conservative. Provision is being made to change the loop position in the beam hole during operation and thus utilize the flux gradient to adjust the power. A higher power density may be feasible for later loops. Fission-Gas Holdup D. W. Magnuson Aircraft Reactor Engineering Division The adsorption of krypton from helium by acti- vated carbon at liquid-nitrogen temperature is being tested in adsorption traps designed for use with the MTR in-pile loop. The traps contain 280 g of Columbia ACA activated carbon. A helium cylinder Sh e fadur giovGoob < fo s @ _ '_mcenner as fhey wn“ in fhe m-plle fes'r : h‘i_n 14 ft_of leng it containing 0.13% krypton is being used to supply five times the design flow rate through one trap. The same supply furnishes the design flow of 0.03 scth ’ro a second system confaining two carbon traps in series. The second trap wilt be isolated at the conclusion of the experiment, and the ad- - sorbed gas will be analyzed for total krypton. After 15 days of operation, the krypton concentration in the effluent stream was approximately 1 ppm in both systems, or the fractional breakthrough was less than 0.001. A temperature-sensing element filled with oxygen, with an automatic fill device, is being used to keep a constant liquid level in the metal Dewar which contains the traps. Bench Test L. P. Carpenter Aircraft Reactor Engmeermg Division A bench test for the |n-p||e loop Yas been oper- ated for more than 666 hr. The purpose of this bench test is to determine the feasibility of con- struction techniques, to tesi the suitability of the various materials of construction, to aid in estab- lishing control and operational procedures and in training operators for the in-pile test, and to deter- mine the endurance of supporting equipment. The design conditions as sut forth for the in-pile loop were cdhered to as closely as possible in the bench apparatus. Modifications were limited to incorporating a resistance-heater coil for power generation. The loop is enclosed in a plug that differs from the ccfucl in-pile loop plug only in that ’rhe forwcrd end is open to accommodate con- nections w:th 'rhe resnstcnce heater. AH servnce .. the alarm system. A leak in the drive pump suction . line has been found to be the cause of cavitation ~and_has been corrected. Tl in-pile power. unit lines to the foop run fhrough ‘the plug m the scrne TR st ing parts durmg construction, and the drawings have been revised in some instances. On-e"Ba q'i‘equre ent fof'cn:""'_‘: | of less ’rhan }’6 in, - - =y \iffl Jlfy was experlenced o ‘becc’”se of WQ"PI'"Q at fherm‘?,wbpie wells and poor fits. F:xtures dre being fabricated for use in align="" PERIOD ENDING JUNE 10, 1955 The commercial Kovar-glass seals used to bring power and thermocouple wires through bulkheads and to form leaktight closures have proved trouble- some by not being adequately leaktight. A seal has been developed at the Gaseous Diffusion Plant, K-25, which is quite satisfactory, except that it is vulnerable to breakage during loop ase - sembly. Fourteen thermocouple leads or eight power leads can be brought through a 1/ -in.~dia glass cup seal. Bench test experience showed that the use of a poured barytes concrete shield in the rear sec- tion of the shield plug required that the lead wires be protected from moisture condensation and the concrete. Glass-braid insulation was applied to individual wires, which were then bunched and encased in plastic tubing.. After the concrete was poured, the wires were found to be shorted. Heat- ing and pumping on the concrete made the con- nections usable, but other steps are being taken to prevent recurrence of this trouble for the in-pile loop. ‘ The operating conditions for the bench test are: temperature differential, 175°F; Reynolds number, 5100; power input, 21 kw; temperature differential for air, 250°F at 250 scfm. Control is achieved by regulation of the air flow through the heat ex- changer by an automatic controller that maintains a nearly constant temperature, +10°F, on the fuel tube at the pump. The electrical power input is manually controlled. o Operation of the loop is proceeding satisfactorily. Cavitation of the hydraulic power unit has caused momentary fluctuations in the pump speed that trip elng" operq'red and checked prior to ship- " ment 'ro_ NR.TS Horlzonfol Shuff Sump Pump o JUA. Conlln o Aircrcff Reacfor Engmeerlng DIV!SIOn | iously described,? was dlscssembled and lnspecfed after complehng ]OOO hr of opercmon. As can be 'seen in th. 2.1, there was no erosion ‘énd only d sllght mdlcahon of rubblng befween the pump 2}, Conlin, ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 32, 27 s ciout il g o e s ' 'ANP PROJECT PROGRESS REPORT UNCLASSIFIED PHOTO-23330 ‘ W Fig. 2.1. Ht;fi.ion'ful-SHin Sump Pump for In-Pile Loop After 1000 hr of Operufio.nuc'.ircul-afi.ng N;:F- Z(F,-UF, (53.5-40-6.5 mole %) ot 1400°F and 1 gpm in an Isothermal Loop. shaft extension and the pump housing. This rub- bing is known to have occurred during preheating of the pump and to have been caused by thermal stresses induced by an unduly rigid pump mounting. There was no evidence of zirconium fluoride vupor in the rear pump housing beyond a point / in, from the pump sump proper, and no salts were found in the vent line after the test. A continuous helium purge of this area of about 0.3 cth was used. There was no evidence of oil in the fuel portion of the pump, the measured oil seal leakage being 0.15 em? for the 1000 hr, However, the pump sump gas pressuie was higher than the bearing " housing pressure, which would tend to cause seal leakage to be toward the bearing housing. The ~ shaft seal faces were in good condition and ap- parently could have opera’red for an additional B 1000 hr, or more, The pump impeller that W|l| be used in the in- - pile experiment is identical with the one used in - -the test described above. The bearing housing - -and seal designs are different, however, in order to overcome the problems of radiation damage to , --;fhe oil and to prevent leakage of fission gases, " Instead of the oil cooling, the shaft is cooled by 30 copper spool on The shm‘t that operates in a he- 8 lium atmosphere with a close clearance to the water-cooled housing. The pump bearings and seals are drop-lubricated. To seal fission gases from the oil in the hydraulic drive motor, two shaft seals are employed, one between the pump sump and the bearing housing and one between the bearing hous- ing and the hydraulic motor. These are facetype seals with metallic bellows for the flexible member. The gas volumes in both the sump and the bearing housing are purged with helium to further reduce the possibility of oil contamination, The pump housing also serves as a leaktight bulkhead in the water jacket to seal the loop end from the bearing housing and motor section of the water jacket. This section between the pump bulkhead and the intermediate bulkhead will be used to accumilate the waste bearing- and seal-lubricating oil. An exploded view of this pump is shown in Fig. 2.2. Difficulty is still being expérienced with failure of the rotating seals. This trouble is principally associated with the bellows and with breakage of the carbon rings. Replacement of the brass bellows with stainless steel bellows and the ex- ercise of greater care in assembly seem to have improved the situation. The sliding surfaces have given little, or no, trouble. PERIOD ENDING JUNE 10, 1955 UNCLASSIFIED PHOTO~23679 Fig. 2.2. Exploded View of Pump to Be Used for MTR In-Pile Loop No. 1. lnsfrumenfoflon B ' ' ment, it has prov:ded a good test of fhe recording ‘R, G Affel ‘_ B A Gnodt _V:{M»L)C“n‘cxon'frOI SYSfems | Alrcraff ReGCfOr Englneer|ng D'VISIO“ o A A T _ lnsfrumento‘hon _design has been completed, and =~ W Cunmngham - ~ the panel for MTR operanon is necrly assemb!ed 4 Asrcraff Reactor Englneerlng Division Operatmg exper:enc’ wn‘h fhe bench test pqnel has L Mosf parts “for the flrst m-plle loop ‘have been - fabrlccn‘ed‘ and ussembly IS under way. A wooden / NRTS. Thi‘s,‘lf is hoped wnlruénsure a proper fl’r ond minimize on-the.site msfalicmon time. 29 — e et AR Lot e b i ANP PROJECT PROGRESS REPORT DEVELOPMEMT AND OPERATION OF FORCED- CIRCULATION CORROSION AND MASS TRANSFER TESTS Operation of Fused-Salt-inconel Loops W. B. McDonald C. P. Coughlen P. G. Smith Aircraft Reactor Engineering Division J. J. Milich ~ R. A, Dreisbach Pratt & Whitney Aircraft The operation of high-velocity forcedscirculation large-temperature-differential loops for investigat- ing the corrosion of Inconel by fluoride fuel mix- tures under dynamic conditions has become so routine that operation can be scheduled, and the schedules can be maintained for long periods with- out serious interruption. Twelve loops have been terminated following scheduled operation; five loops were terminated short of scheduled operating time because of various failures; and five loops were started and are continuing in operation. A summary of the typical operating conditions is given in Table 2,1, During the early part of the quarter, loop failures were encountered because heater lugs were welded TABLE 2.1. SUMMARY OF OPERATING CONDITIONS FOR 22 FUSED-SALT-INCONEL FORCED-CIRCULATION CORROSION AND MASS TRANSFER TESTS Maximum Fused Salt Temperature: 1500°F : : T Maximum o . Loop " Method of ype of Reynolds T?mperat?re Recorded Tube Fused Salt p;lzafmg Reason for No. Heating Heaf.ed Number lefeorenhol Wall Temperature Circulated me Termination . Section . (°F) (°F) (hr) 4950-1 Direct resistance Straight 5,000 200 1650 Na F-ZrF4-U F4a 1000 Scheduled 4950-2 Direct resistance Straight 5,000 200 1565 Na F-ZrF4-U F4‘z 1000 Scheduled 4950-3 Direct resistance Straight 10,000 200 1690 Na F-ZrF4-U F4b 1C00 Scheduled” 4950-4 Direct resisfunce.e Straight 10,000 100 1600 Na F-ZrF4-U F4‘z 1000 Scheduled 4950-5 Direct resistance Straight 10,000 200 1575 Na F-ZrF4-U F4a 1000 Scheduled 4950-6 Direct resistance Straight 8,000 300 16206 . Na F-ZrF4-U F4a 1600 Scheduled 4930-A Direct resistance Coiled 1,060 300 1695 Na F-ZrF4-U F4C 1000 Scheduled 4695-1 Direct resistance Coiled 10,000 300 1720 NaF-ZrF4~UF4b 385 Leak. 4695-2 Direct resistance Straight 15,000 200 1670 Na F—ZrF4-U de 887 P ump-bearing failure 4695-3 Direct resistance Straight 10,000 300 1640 NaF-KF-LiF¥ 630 Leak 4695-4A Direct resistance Straight 10,000 200 NoF-ZrF4-U F4a 0 Terminated after cleaning operation as first of a series for determination of effect of time 4695-5A Direct resistance Straight 10,000 200 Na F-ZrF4-U F4“ 10 Scheduled 4695-4B Direct resistance Straight 10,000 200 NaF-ZrF4-U I=4‘I 50 Scheduled 4695-4C Direct resistance Straight 10,000 200 NaF-ZrF -UF 100 Scheduled 4695-5B Direct resistance Straight 10,000 200 NaF~ZrF4-U F4a 241 Leak 4695-5C Direct resistance Straight 10,000 200 NaF-ZrF UF 500 Scheduled 4935-1 Gas-fired heater Coiled 1,000 300 1670 NaF-ZrF -UF,© 1000 Scheduled 4935-2 Gas-fired heater Coiled 5,000 200 1675 NaF-Zer-U F4“ 1000 Scheduled 4935-3 Gas-fired heater Coiled 1¢,000 100 1540 Na F-Zer-U F4a 1000 Scheduled 4935-4 Gas-fired heater Coiled 10,000 100 1690 Na F-ZrF4-U F4a 486 Leak 7 4935-5 Gas-fired heater Coiled 10,000 200 1645 NaF-ZcF -UF ,* 1000 Scheduled 4935-6 Gas-fired heater Coiled 8,000 200 1550 NaF-ZrF -UF,% 1000 Scheduled EComposition: 50-46-4 mole %. bComposifion: 50-46-4 mole % with 2 wt % of total uranium converted to U3, “Composition: 53.5-40-6.5 mole %. dCOmposifion: 11.7-59.1-29.2 mole %. 30 s col R i | o el SRR e .":‘; ) R - \,4.:9,_51‘-42' to the tube wall in such a manner as to cause high thermal stress concentrations and high electrical current density in the lugs. Three identical failures resulted before butt welds were sub- : stituted for axial bead welds to alleviate the con- dition that caused the failures. Two bearing failures resulted in an investigation of the thrust bearings used in the pumps (model LFB). This study indicated that the fits were too tight, and new bearings with looser fits between balls and races were obtained. No further bearing failures have occurred. - Metallurgical examination of early direct-re- sistance-heated loops in which the heated sections were coiled revealed excessive attack on the com- pression side of the bends. Temperature measure- ments showed the compression side of the wall to be approximately 100°F hotter than the tension side. Since the wall of the tubing on the tension side of the bend would have become thinner during bending and the wall on the compression side thicker, it was thought that the greater current density of the thicker wall on the compression side could have caused the overheating and the resultant excessive corrosion, An inyesfigation3 of flow revealed that poor fused-salt flow distri- bution at and past the bends resulted in poor heat transfer and thus also caused overheating. A new heated section was therefore designed that elimi. nated bends from the sections that carried high current and provided approximately 40 diameters 3H, W, Hoffman, L. D, Palmer, and N, D. Greene, Electrical Heating and Flow in Tube Bends, ORNL CF-55-2-148 (Fe,-l:vt 22, 1955). PERIOD ENDING JUNE 10, 1955 of straight section following each bend. In addition, the length of the heated section in the resistance- heated loops was increased from 12 ft to approxi- mately 17 ft in order to reduce the maximum tube wall temperature to a more tolerable level, These modifications appear to have corrected the over- heating and resulfant excessive corrosion, The gas-heated loops are being run to evaluate the effect of the heating method and the effect of tube wall temperature on corrosion and mass transe fer. Loops 4935-5 and 4935-6 are the first two of a series of three loops fo be run with wall temper- atures of 1550°F, 1650°F, and 1800°F. Sodium in Multimetal Loops C. P. Coughlen D. R. Ward Aircraft Reactor Engineering Division R. A. Dreisbach Pratt & Whitney Aircraft Four loops with sodium in Inconel and two loops with sodium in Inconel and type 316 stainless steel were operated and terminated. The important operating conditions for these loops are given in Table 2.2, Since appreciable deposits of mass- transferred material were found in the cold legs of these loops, a controlled series of experiments was initiated, The operating conditions for these loops, which are now operating, are given in Table 2.3. These loops will provide information on the effects of the oxygen content of the sodium, the use of a cold trap, the use of a lower maximum sodium temperature, and theuse of an all-stainless- steel system. ' emperature ' Operating -~ Ditterential Time 951 Inconel 4951-3 / Inconel >,.‘|5’b.06 S T R > 15,000 180 500 (°F) (hr) 7 oo B [ . 00 4sd ‘300 500 31 oy e IR ANP PROJECT PROGRESS REPORT TABLE 2.3. SUMMARY OF OPERATING CONDITIONS FOR LOOPS NOW CIRCULATING SODIUM Reynolds number: 15'00'0 Temperature differential: 300°F Scheduled operating time: 1000 hr Maximum Sodium L oop Loop Material Temperature Controlled Variation N°t o ("F) 4951-4 Inconel 1300 Maximum fluid temperature 4951-5 inconel 1500 0.15% 02 added 4951-6 Incone! 1500 Special high-purity sodium cold trap used 4951-7 ' Type 316 stainless steel 1500 Loop metal PUMP DEVELOPMENT A series of ten experiments has been performed, E. R. Dytko to date, and the test conditions and results are Pratt & Whitney Aircraft given in Table 2.4. The performance data from . « : . experiment No. 3 are plotted in Fig, 2.3, These A. G. Grindell G. D. Whitman Aircraft Reactor Engineering Division Water Performance Tests G. D. Whitman R. L. Brewster M. E. Lackey Aircraft Reactor Engineering Division A full-scale model of the ART fuel pump design designated MF-2 was built, and water performance data were obtained for several impeller designs and pump suction conditions for a given discharge- volute design. The discharge volute and the test impel lers were fabricated from brass, and a bearing housing used for the model MF-1 tests? was adapted for use in the rotary assembly. The unit was driven by a direct-coupled, 15-hp, variable-speed, d-¢c mofor. The pump was installed in a test loop built of é-in. pipe with head and flow measuring instrumentation and a throttling valve, The first experiment was performed by using a pump-suction configuration that simulated the re- actor design. A box containing a flat plate 1 in, below and parallel to the pump-suction flange was used. For the second test, the flat plate was removed and an 8-in. pipe was connected directly to the pump suction. Since pump performance was about the same in both tests, the remainder of the experiments were conducted with the 8-in. suction line. - 4A, G. Grindell and W. C. Snapp, ANP Quar. Prog. Rep, Muar, 10, 1955, ORNL-1864, p 34. 32 data are representative of the best operation ob- tained during the tests. The pump efficiencies are not considered to be accurate on an absolute basis, because the motor was not calibrated and the motor efficiencies were obtained from the manufacturer’s computed data. It is estimated that the efficiency of the pump, exclusive of seal and bearing losses, is approximately 70% at the design point. Data obtained by varying the running clearance between the lower impeller shroud and the volute indicated that the clearance could be in excess of 0.040 in. without loss in performance at design point. At low-flow high-head conditions, there was approximately a 10% loss in head because of re- circulation in the pump. A cavitation problem persisted throughout the experiments. At flows of over 400 gpm and speeds in excess of 2000 rpm, a slight noise was detected in the pump, and the intensity of this disturbance increased with increased flow or speed above the threshold values. However, no cavitation damage has been detected in the pump, and the data do not indicate a decrease in performance in the cavi- tation region. The performance of the pump was not altered by varying the suction pressure over a range of =5 to +15 psig; however, the intensity of the noise could be suppressed by increasing the system pressure. The suction conditions were altered by changing the radius of the inlet eye from 1/2 to 1 in., and the impeller nut was rede- signed to give better fluid guidance. Neither of - ) L) *y PERIOD ENDING JUNE 10, 1955 TABLE 2.4. CONDITIONS AND RESULTS OF WATER PERFORMANCE TESTS OF ART FUEL PUMP MODEL MF-2 Design point: 620 gpm, 35-ft head Experiment Impeller Design Suction Conditions Remarks and Results Number 1 Five vanes; blade tip Suction box simulating Design point met at approximately 2800 rpm angle, 22 deg reactor; ]é-in. radius on suction eye 2 Same as above 8-in. pipe with four No appreciable change in performance antiswirl vanes; ]/z-in. radius on suction eye 3 Five vanes; blade tip S‘qme as above Design peint met at approximately 2600 rpm; angle, 26.5 deg approximately a 10% increase in efficiency with respect to experiment 1, with the peak efficiency shifted toward higher flows 4 Five vanes; blade tip Same as above Decrease in pump performance with respect angle, 22 deg; leading to experiment edges cut back on an 80-deg cone angle 5 Five vanes; blade tip Same as above No appreciable change in pump performance - angle, 26.5 deg; in- at design point creased lower shroud clearance from 0.010 to 0,025 in, 6 Same as above with Same as above No appreciable change in pump performance radial clearance in- at design point; 10% loss in head with re- creased to 0.040 in, ' spect to experiment at flows below 300 gpm ' ' and speeds above 2000 rpm 7 Six vanes; blade tip Same as above Similar to experiment 1 angle, 22 deg Same 'as;i‘arb::vé _' " ‘ Similar to expérimehts 1and 7 ._'ane Yanes, blade hp _ Ie', 26 5 deg T Tongue cuf back on volufe, performance N S|m||ar fo thut |n éxper:ment 3 T Soma as above Same as above Tongue cut back on velute; mod'lfled :mpeller 'performqnce sumllar to Thqt m ex- per:meni 8 : Beurmg and Seal Tests these chdnges']hqd any’ detectable effect on the _ - ,'caw'rahon problem. It is believed, at present that T D.R. WF’F_‘.* " the cavitation may be a re\syfih of poor fluid guid-" _ W, C Tunnell ‘ J. W. Kingsley " ance at the leading edges of the impeller vanes.“; 7 Alrcraft Reactor Engineering Division A new vane design has been completed and will A test has been desibned for studying the func- be tested soon, ' tion of the interference fit between the journal 33 AP v s ANP PROJECT PROGRESS REPORT EORE " ORNL~LR~—DWG 7553 10 100 90 _ 4——SURGE BOUNDARY 80 / "4\‘ — — — PUMP EFFICIENCY - / / / \ 70 e 6\0/ .-.-"'--. O/ —_—————— st~ O T £e0g e 4\\ T =X $0 S T - 7 =3, - Ll I 40 A= 50 / 3 A& O)/ ” : T . / ' / // // ’ 30/ - \ // X DESIGN POINT /s /) < \fi ///g)j]o/ = L——— O 30 iz — \ _ ' -~ o ===l i |7\ °% . —em T N T\ v o N0 o, 10 | ~ A%, o 05 1000 o > 800 rpm ND\ o .hi'—.-..___.‘ - 0 100 200 300 400 500 600 700 800 900 CAPACITY (gpm) Fig. 2.3. Performance Characteristics of ART Fuel Pump Model MF-2 in Water Tésts. For test con- ditions see Table 2.4, bearing and the face seal in the model MF-2 pump at elevated shaft temperatures and under simulated bearing loadings. The test apparatus consists of an MF.2 rotary element to which a loading device is attached at the impeller location. The pump sheft rotates freely within the side-loading device. Heat is applied at the lower region of the pump to simulate gamma heating under actual operating conditions. One phase of the test will consist in an endurance run under simulated pump design conditions, and the other phase, for which a dupli- cate test assembly is to be used, will consist in short tests for studying the following variables: bearing loading, pump speed, lower shaft temper- ature, rate of coolant flow, and time. Mechanical Shakedown Test W. L. Snapp J. J. W. Simon Pratt & Whitney Aircraft A few additional mechanical shakedown tests were performed on the model ME-1 ART fuel pump. 34 The rotary element was operated continuously for a period of 300 hr at 3800 rpm without achieving a successful seal at the lower journal-bearing region. [t was found that the oil leakage rate was nearly constant at 1.8 in.% per day. Although this rate is not excessive, the goal of zero leakage was not achieved, probably because the face of the journal was not flat, No further testing of this model is planned. As part of the renovation of ARE-type sump pumps for use in heat exchanger and other tests, 100-hr cold mechanical shakedown tests of the rotary elements were conducted on five units. To assist in the development of the metal-to-metal seal for the model MF-2 pump, the floating Graph- itar ring at the lower seal assembly was removed and replaced, on two units, by a modified upper seal bellows assembly. This modification made a metal-to-metal lower seal assembly. One ele- ment failed in the cold shakedown test because of a faulty bellows convolution, but the other element D e e e i i ik o it s e, siva b i was found to be very satisfactory in that no leak- 3 age was detected, The latter element was then ' placed in operation at high temperature on a heat . exchanger test stand. After about 200 hr of hot - _ operation, a bearing hum developed and the unit e was replaced; however, up to that time, there had been no oil leakage detected at this lower metal- to-metal seal. Upon disassembly of the unit, the seal surfaces were found to be in excellent con- dition. As a result of this test, all additional ARE-type pumps will have metal-to-metal seals at the lower journal. Design work has been completed, and fabrication and assembly have been started on two mechanical shakedown test stands for testing model MF-2 rotary assemblies. ' ‘Short-Circuit Pump-Test Stand S. M. DeCamp Aircraft Reactor Engineering Division The study and design of « short-circuit loop for testing model MF-2 pumps were completed, and fabrication was started. Tests are to be made with this loop at operating temperatures in the range 1100 to 1500°F, The tests have been designed s ~ for determining the following: 1. proper operation of the lubricating and coolant system at operating temperatures, 2. leakage rates of the upper and lower seals during actual pump operation at elevated tem- perafures, ’ 3. temperature gradients along the pump support cylinder and in the rotary element, e 4, proper clecrcm(':'es' between the test impe”er ~and pump’ casmg at operating temperatures, 5. 'proper f:’mng of parts at operahng fempercfures cmd |ocds, N'pressure-breckdown Tabyrinth fpr leakage around the pump impeller. This labyrinth was designed to simulate pump-suction and discharge-volute con- .‘fh‘” thé PU“?P mG)'f; be operated at flow rates of Up to _600 gpm and a head of cpprox1mcn‘e|y 50 ft. / PERIOD ENDING JUNE 10, 1955 ditions as presently envisioned for the ART fuel pumps. The removal of pumping power, approximately 40 hp at the isothermal steady-state condition, is to be obtained by constructing a water wall around the pump tank. A movable furnace will be installed between the pump tank and the water wall, It should be possible to dissipate the pumping power at an isothermal operating condition by lowering the movable furnace and exposing a portion of the hot tank wall to the water wall. The fluid flow in this very short loop will be observed through ports during operation with water, HEAT EXCHANGER TESTS E. R, Dytko Pratt & Whitney Aircraft R, E. MacPherson Aircraft Reactor Engineering Division " Heat Exchanger Tube-Spacer Pressure-Drop Tests R. D. Peak Pratt & Whitney Aircraft J. W, Kingsley Aircraft Reactor Engineering Division In order to select the best spacer configuration for the ART heat exchanger tube bundle, a series of tests on various spacer arrangements has been conducted. The test apparatus consisted of a representative ART type of tube bundle composed of 25 aluminum rods 0.191 in. in diameter and 6 ft long contained in a square aluminum duct 1.12 in. inside, The tube spacers were formed from flat. tened copper wire 0,028 by 0.046 in. Tube-to-tube ’ ‘spacing was 0.028 in., and tube-to-wall spacing '"‘was'_" 0. 025 fo 0028 m.' 50 ered_ pn ‘opposite sides of the bundle into The spacer rods were . thick 'i:opper plafe ‘/ in. wide. Water wos me'rered through the qppamtus by a Rotameter, """""_"and total pressure drops were foken across the f\_tube bundmlrpby means of a monomefer. Six confugurahons were tested, and the results of each test are shown in Fig. 2, 4. The results "':ndlcate ‘that the pressure “drop is least when the spacers “afe’ mcllned at 45 deg 'to the dlre»chon of ”_'lflow. Vertlcal, m-lme Spocers, of . cour'.e, gave - :.’”'ifh \ lghest pressure drop Sfuggerlng the spccers inclining these staggered spacers prowded no improvement, as would be expected. 35 saiibainitu: ol il o T jEo dbaiEe G dbaa b ANP PROJECT PROGRESS REPORT SEGRET ORNL—LR—DWG 7554 50 /‘ STAGGERED INCLINED o O 00 OO A 80° 0° m 45° 45° A Q° 459 ® SPECIAL 450 0 74 /// // 17 : A O / 5 /L\‘——TUBES WITHOUT SPACERS wl W 50 - W 2 A/W / @ a 4 / = 2 o B / | / 0° STAGGER, ® 0° INCLiNE—y A 2.0 .l . N I A O AQ W e T .0 7/ < TT—-45° STAGGER, ~ 0 . 2 45° INCLINE >~ 0° STAGGER, 45° |NCLINE 4 / & / ~~——45° INCLINE , SPECIAL 05 l I | | [ . 500 1000 2000 5000 40,000 - F|924 L llé;as:ulfé of Pressure Drop Tests of Vari-d‘hs Tdbe;Spocer Arrungelfienfs for unARTHeot . Exchflngér Tube Bundle. REYNOLDS NUMBER e i ote (T LRSS i e ol Intermediate Heat Exchanger Tests . _ R. D. Peak . Pratt & Whitney Aircraft : Construction is about 70% completed on the test T stand described previously® and shown here in o Fig. 2.5. The two 1.3-Mw 100-tube heat exchanger e bundles have been completed,® as well as the two 500-kw high-conductivity-fin radiators. The NaK pump (DANA), fuel pump (DAC), and the radiator blower are salvaged equipment from the ARE. The 1-Mw gas-fired heater for this test loop is now being fabricated by the Struthers Wells Corp. Small Heat Exchanger Tests J. C. Amos M. M. Yarosh Aircraft Reactor Engineering Division R. I, Gray Pratt & Whitney Aircraft A test of a small fuel-to-NaK heat exchanger was started April 10, 1955. Pressure drop and heat transfer data have been taken through the fluoride mixture Reynolds number range 500 to ’ 6300. Preliminary analyses of the data indicate that the heat exchanger pressure drops are in good agreement with pressure-drop and friction-factor o information obtained from recent water-pressure- ) o S .:;", 5 ST ST ' o 1955, ‘ORNL- 1864, p 36, drop tests carried out on tube bundles with similar geometry. ' Basic heat transfer data are presented in Fig. 2,6, which compares the recent results with the theoretical relationship Nu/Pr0+4 = 0,23 Re0-8, The equivalent diameter used in calculating the Reynolds numbers was based on the total wetted perimeter of the tubes plus the s:de-wall crea,rr " while the equrvolen’r diameter used in compu’rmg the Nuseelt_number uhhzed only fhe ’rube perlmeter. h 6P, Patriarca et al., ANP Quar., Prog. Rep. Mar. 10, 1955, ORNL-1864, p 131. \Inconel “tubes 3/] 6 in. in outside dian'letfier\,-r wrth o The tubes, WhICh are qp',,\,_‘.. flows below this value. e fow separa‘flon is encountered at the outer core shell surface above the equator, turbulators will PERIOD ENDING JUNE 10, 1955 clearance., The fluoride mixture NaF-ZrF4-UF4 (50-46-4 mole %) is circulated outside the tubes, and NaK (56% Na-44% K) is circulated in the tubes, The assembly is now operating on an endurance run at a fluoride mixture Reynolds number of ap- proximately 3000, It is planned to take data over the Reynolds number range 500 to 6000 after 500 hr of operation and again after 1000 hr. CORE FLOW TEST G. D, Whitman R. L. Brewster Aircraft Reactor Engineering Division A full-scale model of the proposed 21-in. reactor core and an entrance header have been fabricated and checked for fluid flow reversal and/or stag- nation, The model was installed in a loop con- taining a 1000-gpm water pump, and two entrance lines were provided at the header to simulate the reactor design. Data were taken at flows that provided a Reynolds number correlation of 1:1 between water flow rates and design fuel flow, The core model was machined from cluminum castings and was instrumented at 72 static pressure and 18 probe points at nine elevations along the vertical axis of the core. Directionsfinding impact tubes were used to traverse the core-shell region at the probe points. There were two prcbe loca- tions at each of the nine elevations and in the entrance header. A traverse was made from the inner to the outer wall of the fluid passage to obtain the direction of flow and a total pressure profile. Static pressure measurements were taken at the walls at each elevation, and the fluid ve- locity was computed from the total and static pressure data, ~~The data were taken in the upper half of the core, dnd a region of flow reversal was encountered “around the inner core shell. " tended approximately one quarter of the distance This reversal ex- between the inner and outer shell surfecces and was not sensitive to flows down to one-half the design Reynoids number, Data were not taken at . The entrance geometry produced a large rotational . component of velocity in the fluid entering the core, and turning vanes are to be inserted to direct the xially through the core volume. If intolerable be attached to the surface in an attempt to reduce this trouble. 37 I “ ANP PROJECT PROGRESS REPORT . . o TWO 500-kw RADIATORS ORNL-LR-DWG 7555 I P Tl NakK PUMP - 4 mmfl" Qo o 5y ol i FUEL. PUMP STl 7 VENTURI FOR FUEL . PRESSURE MEASURING DEVICES - L I S t/ | ¥ - Is\:\ (\/:\.\\.‘:\ VENTUR| FOR NaK T “\é“ PRESSURE MEASURING DEVICES S FREEZE VALVE i—;jt FUEL SUMP TANK m__,fi “W‘n-fi—\—j-f—r( 7777 {-Mw GAS-FIRED HEATER ——— | : FREEZE VALVE NaK SUMP TANK o Flg 25 Isometric Drawing of Intermediate Heat Exchanger Test annd; o o 38 ool b o ._ Nu /PO 100 200 500 PERIOD ENDING JUNE 10, 1955 CTTRET ORNL—LR—DWG 7556 Nu/Pro% = 0023 Re®® 0 GROUP 1 (RUNS 1-19) & GROUP 1T (RUNS 21-38) o GROUP 11 (RUNS 40-53) 10060. 2000 5000 10,000 REYNOLDS NUMBER OF FLUORIDE MIXTURE Fig. 2.6. Heat Transfer Data for 20-Tube Fuel-to-NaK Heat Exchanger. THERMAL-CYCLING TEST OF SODIUM: | INCONEL-BERYLLIUM SYSTEM R. D. Peak M. H. Cooper F’ruh‘ & Whutney A:rcrcft ' - tabulated in Table 2.5. 71bid., p 134 “The ey- The beryllium piece was cycled 104 times be- tween full power and low power, with 20 cycles " having a cycle period of 20 min and 80 cycles “having a cycle period of 4 hr; four cycles were - required for instrument checks, power failures Opercmon of the second sodlum-berylhum—lnconel r P ’ _ compatibility testing apparotus, prewously de- e scribed,” has been completed The tes’r stand |s' | " shown in the isomefric drawin . lindrical berylllum tesf'plece was mounted between; s the secondary connections of a 250-kw transformer - and dlrectly heated by passmg electrical current o fhrough it. Hot sodium was pumped through the T berylhum 'tesf plece cmd fhen to a rodlqtor. The and startup. The time required for a change from ~ full power to low power or back was 2 min. Upon completion of the required cycles, the apparatus . was run at full power to achieve a total operating .. time of 1030 hr. The beryllium test piece is shown in Fig. 2.8 after the unit was disassembled, The beryllium was found to have grown from 0.0003 to 0.0041 in. on an average outside diameter of 1,125 in, There were three axial cracks approximately 1/16 in. long on the outer surface at the hot end of the beryllium. Inspection by the Dychek method revealed no other cracks on the outer surface. 39 ANP PROJECT PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 7557 SURGE TANK == = == FLOW METER RADIATOR ELECTROMAGNETIC PUMP COLD TRAP T SN o - BUS BAR TO TRANSFORMER AN e J BERYLLIUM TEST SECTION ‘l' BUS BAR TO TRANSFO@ ‘ FREEZE ValVE -§ I = i 1 N | Fig. 2.7. lsometric Drawing of Lodp for Thermal-Cycling Tests of a Sodium-lnconelaBeryliliu'r.h System, 40 e PERIOCD ENDING JUNE 10, 1955 TABLE 2.5, SUMMARY OF OPERATING CONDITIONS FOR THERMAL-CYCLING TEST OF SODIUM-INCONEL-BERYLLIUM SYSTEM il b i ke e WaaEE L a i M High Power Level Low Power Level Beryllium power, w/cm3 61 2.5 Current through test piece, amp ' 11,100 1,900 Sodium flow, gpm ' _ ‘ 4.0 4,0 Sfidium' inlet temperature, °F 1050 1050 Beryllium-sodium interface temperature at sodium inlet, °F 1095 Beryllium outside-diameter temperature ot sodium inlet, °F 1225 Sodium outlet temperature, °F 1200 1050 Beryllium-sodium interface temperature at sodium outlet, °F 1245 Beryllium outside-dianeter temperature at sodium outlet, °F 1375 UNCLASS BERYLLIUM TEST PIECE PHOI%(?-—Z.I’.»ZEI#D -s—— DIRECTION OF SODIUM FLOW Fig. 2.8. Beryllium Test Piece from Thermal-Cycling Test of Sodium-Inconel-Beryllium System, 41 ANP PROJECT PROGRESS REPORT ‘The beryllium pieces and sections of the loop are undergoing extensive metallographic examination, Apparatus is being assembled for a third test. - The test stand and test program will be similar to those used for the test described above, The same full-power density, 61 w/c_m3, will be em- ployed, but the average sodium temperature will be increased from 1125 to 1225°F. The test is to be completed during the next quarter. " GAS-FIRED HEAT SOURCE R E.'MacPherson Aircraft Reactor Engineering Division | R, Curry Pratt & Whitney Aircraft The small-scale gas-fired source, described previously,® has been tested at a power output level of 100 kw, The operating conditions for a ‘power output of 100 kw are given below, Sodium flow rate, gpm 7.9 Sodium temperature differential through 380 heater, 7 Furnace output, kw 103 Gas flow rate, scfm at 14,7 psia and 70°F 20.5 Chemical heat input rate (based on - 354 980 Btu/scfm), kw Furnace efficiency (furnace output vs 29 chemical heat input), % 42 Testing was not attempted at higher outputs be- cause a modification of the gas injection ports to reduce the pressure drop would have been neces- sary, Preheating of the heat exchanger section prior to sodium filling was accomplished by oper- ating the burner at low power output and adding cooling air downstream of the combustion chamber. It was possible to maintain heat exchanger tube temperatures in the 800 to 1200°F range by this means. o No difficulties were encountered except for the failure of the spark-plug igniter and several plastic thermocouple connectors because of the high radiant heat emission from the combustion chamber barrel. The gas fiow to the burner was momen- tarily interrupted several times in the course of the test of reliability of relighting. No problems arose prior to failure of the igniter mechanism, mentioned above, With a few minor repairs, the furnace can be put into operation as a vutility heat source. By en- larging the gas injection ports (a minor modifica- tion), the capacity of the furnace may be increased considerably for any future application. 8R, E. MacPherson and R. Curry, ANP Quar, Prog. Rep. Mar, 10, 1955, ORNL.-1864, p 37. e e Jo ‘Srn CrUdeIe J« W, Noaks ROOM-TEMPERATURE REFLECTOR-MODERATED-REACTOR CRITICAL EXPERIMENTS A series of experiments have been done on several critical assemblies of the circulating-fuel reflector-moderated reactor., These experiments embodied a number of different configurations of fuel and reflector, which were described in detail previously.!] The latest of these modifications consisted of a beryllium region surrounded by a fuel layer, which, in turn, was enclosed by the beryllium reflector. This configuration was ex- tended, with reduced dimensions, on opposite sides : of the central reactor region to form ‘‘end ducts®’ which simulate the entrance and exit fuel channels, Some results of a variety of experiments which were done with this assembly are presented here. e G ks ki . o s, Reactivity Meqsuremenis The fuel loading of the critical assembly was such that it contained about 3% excess reactivity, which was utilized for the evaluation of reactivity S|gn of the ART.’ on the crn‘:cal posmons of control rods. " the beryllium refle [ the ¢ ‘the assem_kn:.]y ‘was reduced'from}” remov ing fhé ou’rer (essenha!ly cyhnc?rlcci annulus, sund he‘cenfrol region of 'j-'rvcmnulus was coaxial with the fuel region. The o decre - 1A, D. Callihan et al., ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 41. chcnges incurred by u!ferlng 'rhe sfrucfure ond byi It was feasible to mcke 'rhese_;_'_" 7 “solid cylinders and on a cylindrical annulus were “made along the axis of the critical assembly in * the test volume described above, rements prlor to the reduction of the fuel\:if‘ , , ration “to thaf requ:red to determme the “eritical fuel loadmg. “The evaluations were based “iIn” oné’ e‘“'enmenf ‘the thickness of to 85 ine by .V\;hICh was 27 in. wide and’ 18% in. long. This y wcs ]65 cents. /_ o PERIOD ENDING JUNE 10, 1955 - 3. CRITICAL EXPERIMENTS - A, D, Callihan, Physics Division ’ Do Scott Wo Cn TUI’lne“ S Aircraft Reactor Engineering Division ' , R. M. Spencer, United States Air Force J. J. Lynn, Physics Division E. V. Sandin S. Snyder Pratt & Whitney Aircraft ART Control Rod Materials. To provide bases for the design of the control rod for the ART, measurements were made on samples of several possible components. A test volume was formed by removing beryllium from a region 1/16 by 17, e in. in cross section that extended 21.5 in. along the axis of the assembly from one end to a point 0.72 in. beyond the equatorial plane. A loss in reactivity of 19.1 cents resulted, Consideration was also given to the thimble in which the control rod would operate. Tubes of different dimensions and materials were placed within the test volume, aond the concomitant de- pressions in reactivity were determined. The results are given in Table 3.1. |t is indicated that, in the range of the measurements, the change in reactivity depends more upon the quantity of material present than upon its shape. A plot of the loss in reactivity incurred by a sample as a function of its linear density, that is, its mass per unit length, is given in Fig. 3.1. An evaluation was then made of the neutron ab- sorption properties of a mixture of some of the oxides of the rare-earth elements being considered for use as the poison in the control rod, Tests on The composition of the mixture of rare-earth “oxides tested was the following: . . Amount in Mixture Oxides in Mixture "M (wt %) Sm 203 63.8 G450, 2.3 Dy,0, 4.8 Nd203 0.9 Y1,0, (and others) 4,2 43 ANP PROJECT PROGRESS REPORT TABLE 3.1. REACTIVITY EFFECT OF CONTROL ROD TH!MBLE MATERIALS Dimensions _ ¢ Material (in.) Linear Density Reactivity L.oss - Qutside Wall (in.) {cents) ) Diameter Thickness Inconel _ 0.871 0.035 12.27 25.5 1.255 0.045 20.72 41.7 1.243 0.062 | 32.68 64.7 1.255 0.045 | } 32.99 63.3 0.871 0.035 o 1.250 0.G85 41.75 80.7 1,250 0.085 - | 54,02 97.5 0.871 0.035 ; Type 304 stainless steel 1.253 0.028 14,10 o229 ' Type 302 stainless steel 1240 0.060 27.91 41.7 EeRET The mixture was pressed into solid cylinders 0.450 100 ORNL-LR-DWG 7377 and 1.375 in. in diameter and into annuli 0.790 in. j / in outside diameter and 0,140 in. in width, Each . o / piece was about 3/4 in, in height, and the pieces were combined to give various-length samples, In the first set of experiments with the small- 80 / sized cylinders and the annuli, the Inconel thimble, / which was 1,250 in. in outside diameter and 0.085 70 in. in wall thickness and which decreased the - reactivity 80.7 cents, was in place in the axial § 60 void, The absorber samples were contained in an E INCONEL Inconel carrier tube which could be inserted into z the thimble. The outside diameter of the carrier 2 tube was 0,871 in., the wall thickness was 0.035 - }/ in,, and the end was closed by a 3/32-in.-fl1icl< plug. o 1O The small cylinders were centered in this carrier w S — STAINLESS STEEL tube by a third |nconel sleeve, with an outside . 30 / diameter of 0.500 in. and a wall thickness of 0.020 / in., which, in turn, was centered by aluminum 20 spacers. The effects of the carrier tube and of / assemblies of cylinders and annuli, both separately o / and together, are summarized in Table 3.2, The reference zero for the reactivity changes is taken with the thimble alone. |n all cases the test rod | 0 _ . . . . . i B 5 o 20 o 40 o o o was inserted 21.5 in, into the thimble; fhu.'r is, the . ] LINEAR DENSITY (g/in) | end extended 0,72 in. beyond the equatorial plane . , | of the assembly, _ Fig. 3.1. Reactivity Effects of Thimble Materials. In a second series of experiments a comparison i 1 44 i A 3 G i R S i i & - TABLE 3.2, EFFECT OF SAMPLE CONTROL RODS ON REACTIVITY e . o Reactivity | lL.ength Diameter Sample _ . Loss (in.) | {in.) (cents) Carrier tube only | 16.8 £0.5 Cylinders* 23.0 0.450 155.0 £ 15.0%* Annuli 2.0 0.790 230.0 +20.0** (outside) 0.510 (inside) Cylinders and 21,0 0.790 250.0 +20.0** annuli together *The innermost centering tube alone decreased the re- activity about 5 cents. **Obtained by exfropo!chon of a series of measurements with shorter samples, effect of carrier tube mcluded was made of the effecf on reactivity of the dlameter of the samples of oxide mixture. The diameters of the samples were 0.450, 0.790, and 1.375 in., the one of intermediate size being formed by plac- ing the small cylinders in the annuli described above. The samples were about 5.5 in. long, and each was tested at the same position in the as- sembly, The distance from the end of the sample to the midplane of the assembly was 2.0 in. |t was necessary to remove the 1.25-in.-dia thimble - for these measurements, and the absorbers had to be wrapped in aluminum foil. The changes in reactivity measured, referred to the array with the test section void (without the thimble), are given in Table 3.3. TABLE 3.3 EFFECT OF ABSORBER i _DIAMETER ON REACTIVITY - | Diameter 0 Lenath Loss. (m) - (_'nf) L .. (cents) PERIOD ENDING JUNE 10, 1955 a 3-inclong cylindrical sample, 0.450 in. in diameter, was measured at a point where the cadmium fraction measured by gold-foil activation was about 0.5. The measurement was then re- peated with the sample covered by cadmium 0.02 in. thick. The values were 55.9 and 61.7 cents, respectively. The cadmium alone reduced the re- duced the reactivity 37.7 cents, From these data it appears that about 43% of the reactivity decrease is due to neutrons which penetrate the cadmium. Inconel. A measurement, similar to that just described, was made with an Inconel sample in order to estimate the dependence of the reduction in reactlvny upon neutfron energy. The sample was 11/2 43/8 X /32 in.; it weighed 84.14 g; and it could be provided with a 0.02-in.-thick cadmium cover. The effects, as poisons, of the [nconel and the cadmium were measured singly and together at the center of the beryllium island and in the fuel region adjacent to the fuel reflector interface. (The I/8-in. Incone! core shell was reduced to % in. in thick- ness in the region of this measurement.) The gold- activation cadmium fractions at the two positions were 0,57 and 0,33, respectively. The reactivity changes are given in Table 3.4, The fractions of the total change that are caused by neutrons penetrating the cadmium are also tabulated, TABLE 3.4, REACTIVITY CHANGES FROM BARE AND CADMIUM-COVERED INCONEL An estimate of the ‘spectral distribution of the neutron absorption by the rare-earth elements was also made. The change in reactivity produced by At In Fuel-Reflector Island Interface “Reo.‘o‘fivi'ij loss (cents) - Ificonél ‘ 18.0 4.0 lnconel cadmlum covered 17.4 26.8 ' Cadmlum cover 115.2 26.5 Fractional chonge inre- 12,2 10,6 7.5 £2.6 _dctivity caused by epi- cudmwm neutrons (%)_ Gold-achvohon codmlum 0.57 0.33 fractlon o BB |, it is also bemg consndered as a reacfor {5\ )..\?‘r"‘.a., mafenals was therefore mcde in the crmcal as- sembly. A sample of columbium 4% 7, % 3 % 0.05 in. 45 (o o33 Rkt ANP PROJECT PROGRESS REPORT .\.f‘ ‘weighed 98.5 g and was located in the fuel region adjacent to the ¥ .-in. Inconel separator at i+ the fuel-reflector interface reduced the reactivify 3.0 cents. A sample of Ifiéonel 43/4 x 3/ X / in. < that welghed 127.4 g and was located q’r the Sc:me “position reduced the reactivity 6.6 cents. For . value must be reduced to 5.3 cents, if it is assumed © that the effective absorption varies linearly with i the scmple thickness. The layer of {aconel adja- ‘ cént to the columbium simulated the cladding ”'whlch ‘would probably be required for corrosion e&:tlon. Some reduction in the critical uranium ":_‘frahon would be effected by the use of columblum. " Beryllivm in the Fuel Region. Since the pres- ence of beryllium in the circulating fuel stream would be another possnble way of reducing the :crmcal urcnlum concenfrahon, an experlment was devised for. evaluahng the effect of beryllium in ~ the fuel | region of the critical assembly, One of S the cenfrcl fuel subassemblies that consisted of 27 uranium metal sheets (each 0. 004, in. thick) and . 55 Teflon sheets (each /]4 in. thick) was modified ' -_.‘_m the: followmg manner. Seven uranium sheets C o and 16 Teflon sheets were removed from the center o effecf a decrease in reacflwty of 22.7 cents. . The subsnfuhon of a somple of beryllium, 7/ X 27/ x 1 in., for this uranium and Teflon resulfed ~ in c net reactivity gain of 8,5 cents over the un- “perturbed value. The addition of a % -in.-thick f ;'}_’ layer of lnconei cempletely around the beryllium reduced this net gain to 4.0 cents. An estimate of _i"i' T‘fhe effect of the |ncone| clad beryllium on the "concentrahon IS glven ‘by the observation that the “removal of two of the remaining 20 uranium sheets " further reduced the reactivity by 4.5 cents, that is, - “to a value slightly below that of the original array. - The two sheets removed last were adjacent to the beryllium sample and were of greater than average lmporfunce. ~Fuel in the End Ducts. One of the safety features " of the high-temperature critical experiment presently 5 ""'bemg designed, which is described below, is that . the liquid fuel can be removed by draining under HY‘J_,_grav:fy. The degree of safety depen_ds upon the o 5_rcte of removal cnd the sensmv‘f LN e k{',.{ '/:‘4'1"':,upper end duct. A measure of "*obfclned m the room-tempera 1 D T e 0 O"’ B :J- ¢ e :3"" (D U- S0 m (.h ..., 0 I e o S0 O_ W (fl (Q - 0 gy ¢ — | CD T -0 - , gy C o "j-f':f{_ drmn system. a6 ‘comparison with the columbium value, the Inconel - S o the “fuel Ievel pqrhculorly the’ “Ievel in the hi: sensmvaty was' ) Three annular rings of fuel were successively removed in 2%-in. increments from one end duct, and the resu?’rmg decreases in reactivity were noted, Each ring contained 825 g of U?33, and thus a total of 2,475 kg of U235 was removed. The losses in reactivity accompanying the removal of the three rings are shown in Fig. 3.2, which also gives the loss in reactivity per unit displacement, or fuel sensitivity, averaged over each of the fuel annuli removed. SECREL ORNL—LR—DWG 7378 200 180 160 » c 8 140 > E > 120 '_ Q W W 400 2 —_ p 80 g Q & - o 60 8 > = 40 > = o 20 & [#2] o 0 1 2 3 4 5 6 7 8 9 FUEL REMOVED (in} Fig. 3.2. Effect of Fuel in End Duct on Reac- 'HV“'YG Power Distributions The relative fission rate distribution across the fuel annulus was determined from the fission- product activities collected on aluminum foils in contact with the uranium. Exposures were also made with the foils and the uranium enclosed in 0.02-in.-thick cadmium in order to obtain a measure_ of the energy distribution of the neutrons causing . fission. The locations of foil traverses within the fuel section of the assembly are shown at the top of Fig. 3.3. As may be recalled, the fuel consists of laminae of vranium (0,004 in. thick) and Teflon (0.142 in. fhlck), the measurements were made at " selected positions on lines perpendlcular to these laminae, The fission rate distribution ocross each traverse is plotted in the lower part of Fig. 3.3 as a function of the number of uranium sheets between _ Ly RELATIVE ACTIVITY bl - B RIS b B B BBy By Bogiciail, 40 30 20 Cm B e SR sl ORNL-LR-DWG 7379 - BERYLLIUM REFLECTOR FOILS |—-———— END DUCT N T b T [T Nt I 1 - N ) L 4 4 i D E F BERYLLIUM ISLAND | o l ® BARE CATCHER FOIL O CADMIUM-COVERED ’ ¢ CATCHER FOIL : CADMIUM FRACTION' i B C D ' ® F 9 NV Q o) l{ )l " \/ 27 0 27 0 27 0 8 O 8 o 8 LAYERS OF FUEL FROM ISLAND {1 LAYER =0.004in. OF U AND Q.14 in. OF TEFLON}) Fig. 3.3. Power Distribution Across Fuel Annulus. c.8 0.6 0.4 0.2 CADMIUM FRACTION §S61 ‘0L ANNr ONIANI QOl¥3d e i PR the datum point and the island of the assembly. ~ Values of the cadmium fraction (the fraction of all ~ fissions produced by neutrons having energies less than ~ 0.5 ev) are also plotted, At the time these measurements were made, the reflector around the . center sécfidn ‘of the assembly was only 85/8 in. 'ASSEMBLY FOR HIGH-TEMPERATURE CRITICAL EXPERIMENT " QOperation at high temperature of a critical as- -~ sembly mockup of the ART is scheduled for late ' _sdmmei', 1955. |n addition to a determination of “the critical uranium concentration at the elevated " temperature, the purposes of the experiment are to “evaluate the temperature coefficient of reactivity, to investigate the effectiveness of control rod materials, and to measure the contribution of the fuel in the upper end duct to the over-all reactivity. The experiment will be performed at zero nuclear power and at about 1200°F, Heat will be supplied by electrical heaters external to the reflector. The . assembly is to include a beryllium island and a - beryllium reflector, essentially 12 in, thick, which ~will be in a helium or argon atmosphere and will contain no sodium. The ART core configuration, as presently envi- sioned, will be exactly mocked up in the experiment between points 18 in. above and below the equatorial plane of the core. Some minor deviations from the s ART design are being made beyond these points to simplify filling and draining. The fuel will not be circulated. The core will be filled with molten fuel from a sump tank by using helium to force the molten salt into the assembly, The system will be tilled initially for cleaning and testing with a 50-50 mole % mixture of NaF and ZrF . [ncrements of Na,UF (with the uranium enriched with 93% U233) will subsequently be added to the NaF-ZrF, mix- ture in the sump tank. After each addition of Na,UF the mixture will be pressurized into the core and then drained until the critical vranium concentration is determined, | A single rod will be located within a 1,50-in.-1D Inconel thimble along the vertical axis of the beryllium island. |t will extend from the top of the reactor tank to 10 in, below the equatorial plane of the core. The control rod will be a cylindrical annulus of the mixture of rare-earth elements described above, enclosed in an |nconel shell. Annuli of two widths, 1/8 and ]/4 in., will be provided for comparison. ' The system is being designed to operate iso- thermally at 1200°F normally, with provision made for short periods of operation at temperatures up to 1300°F to enable reactivity temperature coefficient measurements to be made. An attempt will also be made to measure the fuel temperature coefficient by inserting the fuel into the reactor assembly at a temperature different from that of the beryllium. .y o, Part Il MATERIALS RESEARCH o ¥ o W : _accelerated engmeermg test progrom - 4. CHEMISTRY OF REACTOR MATERIALS W. R, Grimes Materials Chemistry Division -~ Phase equilibrium studies were made of the systems LiF-ZrF,, NaF-LiF-ZrF, and NaF-LiF- ZrF -UF,. Two BeF ,-beafing systems - NaF- LiF-Berund NaF-LiF-BeF,-UF , — were studied, and the solubility of UF, in BeF ,-bearing systems was investigated., A method for zone melting of fused salts was devised as an aid in phase equnllbrium “studies. Additional work was done in mvestlgctlng the equilibrium reduction of FeF by hydrogen in NaZrF, the reduction of U'F by structural metals, and fhe stability of chromlum and iron fluorides in molten fluorides. Vapor pressure measurements were made on UF in the temperature range 1270 to 1390°C. ‘The investigation of the variables affecting the reduction with metallic uranium of UF, dissolved in alkali fluorides was continued. A study of the effect of 'rhe nickel filters used in expenments for determlnmg the stability of UF3 in the NaF-KF- LiF system showed that disproportionation of the UF, occurred because of the filter, Fuel purification and preparation research included experimental use of electrolysis under a hydrogen atmosphere fo remove oxides and the use of metallic zirconium to replace most of the hydrogen in the stripping operation. In addition, attempts were made to prepare mixtures containing very high UF,/UF, ratios., A study of the con- ditions for the preparahon of BeF ,-bearing melts was conhnued and producnon operuhons were resumed to provnde fes? moterlai for the_greatly - (‘J’\ nrfi e oM Bratcher, V. S. Coleman, and C. J, Barfon, ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p 44. PHASE EQUILIBRIUM STUDIES C. J. Barton F. F. Blankenship Materials Chemistry Division H. Insley, Consultant The very considerable interest in obtaining fuel mixtures with physical properties more favorable than those available in the NaF-ZrF ,-UF system has led to evaluation of the NaF-LiF-ZrF, ternary and the NaF-LiF-ZrF ,-UF, quaternary systems. Largely as a consequence of studies pursued during the past quarter, the former system has been reasonably well defined; the latter requires considerably more effort. While phase equilibrium data show that quite low melting points are available in this system and that adequate melting points are available at ZrF, concentrations as low as 21 mole %, no compo- sition with physical properties better than those available in the NaF-ZrF -UF, system has been demonstrated. Study of the analogous NaF-LiF-BeF, system has been continued. Adequate melting points are available over wide areas in this system, Whether the physical properties of proper compositions in this system can show significant advantages over those in the NaF-ZrF -UF , system cannot yet be answered with certainty. The Binary System LiF-ZrF, R. E. Moore R. E. Thoma Materials Chemistry Division D L __StocktonMMerck & Company A ?en’raflve | ~diagram f “the LiF- ZrF bmary system, based prlmorlly on thermal anulys:s data, _was published prewousiy. A re-examination of ... this system by pefrograph[c and x-ray diffraction studies of quenched and slowly cooled samples and by dlfferenhal thermal anclysns wos started A ___durmg the past qucrter becouse of currenf mferest _in the NaF-LiF- ZrF, system. o ) Alfhough quenchmg of LiF- ZrF4 mixtures is s : r\oir comp|ete|y sahsfoctory because of erld phase re!ohonshlps in fuel” systems “of interest. crystallization of ‘many “mixtures in the system, some conclusions may be drawn from the results. A eutectic between LiF and Li;ZrF, at about 51 T Y ANP PROJECT PROGRESS REPORT 21 mole % ZrF was confirmed by quenching "expenmenfs on both sides of this composition. The eutectic temperature found by quenching is ~ about 600°C instead of the 590°C indicated by - thermal analysis. The melting point of
  • 7rF + 4e= , o = Zr° + 4F(, | stan) while at the cathode the reaction is o+ 4e — 7Zr° + 4F" . (“]) (az) + 4F= (@)) ZF v arg — S ZrF 4(61}) (d ) 4(a ) ‘However, ‘hb'i"g-;"t;c;i"i'dh through the bridge proceeds according to 4 ' (a)) 0, s (ay) - z)F(a), where ¢ is the transference number of the sodium 77 "ANP PROJECT PROGRESS REPORT TABLE 4.19. MEASURED POTENTIALS FOR Z¢F , CONCENTRATION CELLS ) () 2) _ ":Te_mp'é.rat'ure EMF (v) ;: S Cell No, 1* Cell No. 2% Cell No. 3* . Ey +E, . o700 0.141 0.120 0.264 o026l : 650 0.129 0.113 0.252 0.247 } 600 0.114 0.113 0.229 o 27 i 550 0.092 0.108 0200 ) "0,200 B o *""'.‘C'ellAN-d 1‘ C..l = 41.8 mole % ZrF4; C:: = 50.0 mole % ZrF“1 : Cell No. 2: cy = 36.0 mole%ZrF4;c; = 41.8 mole % ZrF4 Cell No. 3. .c.l = 36.0 mole % ZrF4; c.} = 50.0 mole % ZrF4 lon. The 'ro'rcl cell recctlon is TABLE 4.20. ACTIVITY RATIOS FOR NuF - o IN NeF-ZrF MIXTURES AT 650°C ' ZrF 4 41 NOF —é ZrF + 4t NaF . 4 4(ay) (al . : F NaF vy NaF The elecfromoflve force of the cell may, therefore, (cg) (29) (ag) be expressed as NaF NaF | yNoF | 4¢ RT ZrF4(a ) NaF(al) E = oI 2 CeliNo, 2 64.0/58.2 2.4 2.1 4F Z'F4(a:) NaF(a,) CellNo.1 58.2/50 3.4 3.0 Cell No, 4* 50/44.8 1.1 1.0 It is reasonable to assume that the fluoride ion | in such a system is rather completely complexed *Cell No. 4: ¢y = 50.0 mole % ZrF ,; c, c1 =55.2 mole % by the ZrF, to form complex ions of the type ZrFi Eg5q0c =0.010 v. ZrF.~, ZrF '”', etc. Since current transported by such Icrge ions should be very small compared with that carried by small simple ions, it seems likely that the transport number of Na' is near unity, The vapor pressure of ZrF, over NaF-ZrF 4 melts has been measured with considerable precision. |f the ratios of vapor pressures for ZrF, at the concentrations shown above are cssumed to represent the activity ratios of ZrF and if ¢ is assumed to be unity, the activity rahos for NoF in the various combinations can be ob- tained directly from the emf equation. The results of such calculations are shown in Table 4.20. From the calculations, it appears that the rapid and uniform addition of ZrF, to molten Na,ZrF - reduces the c:chwty of the NaF until the 50-50 "mole % composmon is reached; further additions of ZrF , do not alter the NaZrF . complex, Cells with Structural Metal Fluyorides. Cells consisting of metallic nickel electrodes immersed in solutions of NiF, in molten NaF-ZrF , mixtures showed potentials of 1 to 3 mv when the NiF, concentrations in the half cells were varied; the Nil'-'2 concentrations were, in all cases, suf- ficiently high to afford a saturated solution. Similar Fe/FeF, cells in which the half celis contained equal concentrations of FeF, showed potentials of 1 to 10 mv. When the half cells contained differing FeF, concentrations but more FeF, than that required for saturation, potentials were obtained which varied from about 40 mv at 550°C to 10 mv at 700°C. Exact voltages varied from cell to cell, but the decrease in emf with temperature seemed quite reproducible. © A cell of the type PERIOD ENDING JUNE 10, 1955 FeF,(c,) 7:0, FeF3(c:) +» NaF-ZrF, NaF-ZrF, , Fe , FeF (c,) NaF-ZrF, Fer(c;) where o . where ¢y = cyand ¢, = Cg + ¢, = 7.0t07.3wt%, resulted in emf’s of 1 mv. It was believed that all the FeF, would be reduced to FeF, by the iron electrodes and that as a result a pure Fe/FeF cell with etched Fe electrodes would be obtained. However, analysis of the final melt proved that most of the FeF, was still present; it is not likely that equilibrium was established, Cells of the type ‘ ZrO2 - Zr ZrF NaF 4leqg) ! NaF-ZtF, have yielded moderately reproducible potentials at various temperatures, |In these cells, ¢, = 47 mole % and ¢, = 5.0 to 5.6 wt %. From the data obtained, E° is estimated to be 1.4]1 to 1.42 v by assuming that the saturating phase is the pure fluoride and that the activity of the ZrF, can be established, as before, from vapor pressure data. From this E°value, it appears that AF® = <131 kecal for the reaction Zr® + NiF, —> ZrF, + 2Ni° This is in excellent agreement with AF® = =127 - kecal based on the thermochemical estimates of'_;___ ‘ Brewer 25 Na F--ZrF4 it c, = 5. 1to54wt%, have yielded the data shown in Table 4.21. The thermochemical estimates of Brewer suggest that E° for this cell should have a nearly constant value of 0.25 v over the 550 to 700°C temperature interval. Again, examination of the solidified melts showed that isomorphous complex compounds NaF-ZrF, , NiF, Ni €2 sat’d of NiF2 and FeF2 with the solvent occurred. If the junction potential of this cell can be assumed to be negligible, then at 650°C, » RT aFeF2 042v = 0,25V = — In —— 2F aniF 2 and AFeF 2 —_ = ~10-2 , anie Nle " Since the saturation solubility of FeF, is nearly There is some evndence, “however, from pefl.o,"""“'I"'""fw'i'c"e that of NiF, ot this temperature and since - graphic and x-ray ‘examination of the cooled melfs' ,__whlch mdlca'res that ’rhe soturctmg phdse |s no'r"' ) the fluoride ion activity should be similar in the "‘be""s'dlu'rions, it appears that the activity coef- ' f|c1ent for NI is hlgher by '200-1"old than thot for f‘f_'j;fluonde._ The da’ra presenfed WOuld seem to . indicate thot thls complex fluorlde is qun‘ew ‘ "Vii-funstable. | C : : ~ Experimental mecsuremenfs of cel!s of the type ' 25L Brewer et al., p 107 and '110 in Cbemzstry and — Metallurgy of Miscellaneous Materials; Thermodynamics (ed. by L. L. Quill}, McGraw-Hill, New York, 1950, 2 - . NaF-ZrF,, N:F2< c, ) Ni sat’d 79 s i e Mk B dian e 'ANP PROJECT PROGRESS REPORT TABLE 4.21. POTENTIALS OF CELLS Fe/Fer/Nin/Ni IN NuF-ZrF4 SOLVENT EMF (v) Temperroiure | (°C) Trial No. 1 Trial No. 2 700 0.425 0.423 650 0.2 0.424 600 0.414 0.416 . 550 0.407 0.409 Fett, H appecrs also that the solvent-Fe ** complex is much more stable (about 8 kcal/mole) thon is ihe comp(ex mvolvmg Ni* Vapor Pressures of LIF-ZrF Mixtures | R E. Moore Materials Chemistry Division ‘The determination of the vapor pressures of a series of LiF-ZrF, mixtures by the method of Rodebush and Dixon2% and Fiock and Rodebush27 26W H. Rodebush and A. L. Dixen, Phys. Rev. 26, 851 (1925) E. F. Fiock and W. H. Rodebush, J. Am. Chem. Soc. 48, 2522 (1926). was started during the past quarter because the NaF-LiF-ZrF, system is being considered as « possible fuel carrier. The vapor pressure work on mixtures containing 33.3 mole % ZsF, (LiéZrFé) and 50 mole % ZrF, was completed. The data for these two mixtures are given in Table 4.22. ~ The vapor pressure equations, which were ob- tained from the best straight lines on a log pressure vs reciprocal temperature plot, are, for the 33.3 mole % ZrF, mixture, log P {mm Hg) = —(8333/T) + 7 967 and, for the 50 mole % ZrF mixture, log P (mm Hg) = —-(8848/7‘) £ 9397, where T is in °K, The heats of vaporization are 38 kcal/mole for the 33.3 mole % ZrF, mixture and 41 keal/mole for the 50 mole % ZrF, mxxture. The vapor pressures of the LiF- ZrF4 mixtures are considerably higher than those of the corre- sponding NaF-ZrF, mixtures. For example, the vapor pressures of the 50% LiF mixture are 50 to 100% higher than those of the 50% NaF mixture. It might have been expected that the small [ithium ion would produce compounds in the fused state which were more stable than the sodium com- pounds. One possible explanation for the higher vapor pressures is that LiF,~ ions may exist in the melt, If the lithium ion has a marked tendency TABLE 4.22. THE VAPOR PRESSURE OF TWO LiF-ZrF4 MIXTURES Temperature Calculated Pressure Observed Pressure (°C) (mm Hg) (mm Hg) LiF-ZtF , (66.7-33.3 mole %) 944 13 13 992 22 24 1014 30 31 1059 54 52 1112 89 89 1178 166 168 LiF-Z¢F , (50-50 mole %) 809 16 16 812 17 17 '330 23 24 849 33 32 897 67 68 941 125 130 208 977 209 80 Gl i Ei & iy to attract fluoride ions to produce such complexes, fewer fluoride ions would be available for the formation of complex ions with ZrF,. Solubility of Xenon in Fusé.dwlgalts R. F. Newton Research Director’s Department The previously described procedures?® for de- termining the solubility of xenon in fused salts were modified so that the source of the spread of values obtained could be ascertained. Long ex- posure without stirring gave essentially the same results and thus indicated that the supposition of the production of fine gas bubbles by stirring, which bubbles were then transferred to the stripper along with the liquid, was unfounded. However, fong continued cycling of helium through the melt, without recent exposure to xenon, gave material which was caught in the liquid nitrogen trap and was read on the McLeod gage as xenon, This material may have been SiF, from the reaction of the glass with the HF liberated from the NaF- KF-LiF (11.5-42-46.5 mole %). Means for elimi- "nating this material or correcting for it are now under s’rudy. X.Ray lefrochon Study of quuxds P. C. Sharrah M. D. Danford H. A. Levy P. Agron R. D, Ellison M. A. Bredig Chemistry Division The construction of an x-ray diffractometer de- signed specifically for studies on the structure of liquids was completed recently. The diffraction pattern from the horizontal surface of the liquid sample is obtained with a divergent beam tech- nigue similar to the Bragg -Brentano flat- sample ... 'system frequenfly used on powder somples. The “‘instrument prowdes for simultaneous angular ~: ~surface. Pl fdrnoce for work wnth o in the de5|gn. “sential to good work “with liquids, is obtained "*”_}hrough the use of c benf crysfol monochromcn‘or 28R F. Newton, ANP Quar. Prog. Rep. Dec. 10, 1954, ORNL-1816, p 72. PERIOD ENDING JUNE 10, 1955 mounted on the arm with the de’rector.' A scintil- lation detector and a molybdenum target x-ray tube are in use in the system now undergoing final testing. High-Temperature X-Ray Spectrometer Studies G. D. White Metallurgy Division T. N. McVay, Consultant A high-temperature attachment for an x-ray spectrometer is being used to help clarify some ~of the phase relationships in fuel systems of interest, The apparatus consists of a water- cooled aluminum cylinder with a vacuum-tight lid which fits the open end of the cylinder and o projecting rod on the closed end by which the attachment is fitted into the goniometer. The sample holder, tantalum heating element, reflector, and thermocouple leads are all mounted on the lid, which also contains an aperture for pumping a vacuum. Inserted around the center of the cylinder is a beryllium window through which the x rays pass., When assembled, the attachment can be pumped down to a vacuum of less than 1 p by using a Welch pump and a small diffusion pump. Initially, the attachment was aligned by using a Th02 sample, The bracket which holds the sample can be rotated and can be shifted vertically or horizontally. It is held in position by three radial set screws, By manipulating these set screws, the ThO sample was put in a position where one of its sfronger x-ray peaks was at the proper angle and had maximum intensity. Thus far the alignment has remained true, although samples have been x-rayed almost dmiy for the past three months, In order for the attachment to be used most " effectively, the sample to be studied must give an” x-ray diffraction pattern with fairly intense - motion of the x-ray ‘tube and the defector on arms _"_;l'predksr the diffraction patterns of the polymorphs | mévmg oboui a horrzontal axis lying in “the hqu:dw _7 TI‘HS ‘makes it possible to work with liquids ‘with a free “surface where sample holder “absorption and sccn‘terlng are ellmlna’red A small" o ' { " and loaded into a nickel sample holder with space Monochromqflc X mdjgf,on,' es':’"":;"'m'll eft for expansion. After the sample holder has “must be considerably different, and the inversions "ot phase changes must be fairly rapid, especially "if the sample being studied oxidizes readily, The sample to be x-rayed is ground to a powder been mounted in the bracket and fhethermocoup|e bead has been placed on the surface of the ~~"powder, the lid is fastened to the cylinder. The system is pumped down overnight, and the x-ray diffraction patterns are obtained the next day. 81 ANP PROJECT PROGRESS REPORT Roc_')m-temperafure patterns are always obtained before and after heating to check on the alignment and flatness of the surface of the sample. - Tol_d‘q'fe, work has been done on compositions in the systems: NaF-ZrF_, LiF-ZrF,, and NaF- BeF,. Only the study of the composition ‘2NaF.BeF, is discussed here. Two samples of Na2BeF were found to be different optically, One of :‘1e samples contained twinned monoclinic crystals, whereas the parallel extinction of crystals in the other indicated an orthorhombic structure. Room-temperature x-ray ~diffraction patterns of the samples were very similar, the only difference being in the intensities of the fieaks; - ngh-temperafure X-fcy diffraction patterns were also obtained on the samples, When heated, both ": . "$“c‘l'mp-le_s' showed inversions at 236 and 336°C, “and when cooled, showed inversions at 310 and below 185°C, The temperature could not be maintained kelow 185°C, and therefore the lower inversion temperature could not be obtained. The intensity and 4 values at room temperature, 300, and 340°C are 82 At Room At 300°C At 340°C Temperature - d I - d I d i - Sdrh#le 1 4.18 14 2.71 35 2.78 7 3.90 6 2.65 33 2.65 27_ 3.67 10 2.34 30 2.18 27 2.92 35 2.25 7 2.11 8 2.68 10 2.21 22 Sample 2 2.62 20 2.17 22 2.9 12 243 26 2.1 13 2.78 7 2.37 32 2.02 15 2.73 9 2.23 9 ' 2.65 27 2.11 17 2.18 27 1.84 18 a2 8 These data suggest that vth-e.'f samples are actually a mixture of two phases: g room-temper- ature phase and the 300°C form. There are several 4 values at room temperature which are present at 300°C if allowance is made for ex- pansion of d values as the temperature is increased. Therefore, the difference in intensity values of the two samples at room temperature could be due to a difference in ratios of the two phases in the samples. " . PERIOD ENDING JUNE 10, 1955 5. CORROSION RESEARCH W. D. Manly G. M. Adamson Metallurgy Division W. R. Grimes F. Kertesz Materials Chemistry Division Several Inconel forced-circulation loops that were “operated with fluoride mixtures and with sodium as the circulated fluids were examined. Results of corrosion and mass transfer under dynamic conditions were obtained. Data were obtained on an alkali-metal base fluoride mixture containing UF, and on ZrF ;-base mixtures with UF ~and with combinations of UF, and UF, circulated at low (1,000) and high (15,000) Reynolds numbers at a maximum fluoride mixture temperature of 1500°F. Further thermal-convection loop studies were made of alkali-metal base fluoride mixtures with UF, and UF, and ZrF,-base mixtures with UF, and UF | in Inconel. The effects of temperature on mass transfer and of the size and shape of the loop on corrosion were investigated, and an evaluation of the erratic results recently obtained with control loops was made. Several brazing alloys on type 310 stainless steel and ''A’" nickel were tested in sodium and in a ZrF ,-base fuel mixture in an attempt to find a brazing alloy that has good cor- rosion resistance fo both mediums. Cermets that do not form solid-phase bonds were screened for suitability for use in valves, bearings, and seals exposed to liquid metals and fused fluorides. Mass transfer of sodium in an Inconel thermal- convection loop ‘and of lithium in a type 347 stain- less steel loop was studied, as well as the cleanlng *of berylhum-lnconel systems with Versene. In the fundamentol corrosion research, further ~ work was done on the mveshgahon “of film for- ‘me’rlon ‘on metals, |ncludmg tests of oxidation “theories, and oddlfional experlments were conducted in the lnveshgahon ‘of corrosion and mass transfer L f"of mefuls by fused sodium hydroxude. o The chemlcal sfudses of corrosion include in- .:-.LIF BeF _ and by LiF-BeF, UF4, ‘the effect of '-UF UF mixtures on corrosnon of Inconel by o fvarious 'solvents and studles of the sodium :;'fhydromde-—mcke[ reachon o 3 ‘\,;..’_3_,.'_;;1 ]L. A. Mcnn, w. B. McDonald and W. C, Tunnell, ANP Quar. Prog. Rep. Dec, 10. 1954, ORNL- ]816 Fige 3.4, p 45. found in the first heated section. < IC > _ ___study of this loop showed that the maximum wall eshgahons ' of “the corrosmn of Inconel by “in this area. 'FORCED-CIRCULATION STUDIES G. M. Adamson R. S. Crouse Metallurgy Division Fluoride Mixtures in Inconel Several forced-circulation loops that were oper- ated by the Experimental Engineering Department at the request of the Metallurgy Division were disassembled and examined. The conditions of operation of these Inconel loops in which fluoride mixtures were circulated are summarized in Table 5.1. Two of the loops examined (4695-1 and -2} had circulated NcF-ZrF4-(UF4 + UF3) (50-46-4 mole %) in which about 2 wt % of the uranium had been converted to U3Y. The design of loop 4695-1 was similar to that used previous[y.] Operation of this loop was terminated after 385 hr because of a leak at a heating terminal. Examination showed the maxi- mum attack, which was found in a bend in the first heating leg, to be 12 mils (Fig. 5.1). In a straight portion of the heated section, the maximum attack was to a depth of @ mils. The depths of attack were thus only slightly less than those found in loops which circulated a similar mixture containing no trivalent uranium; however, the number of voids was reduced by about one-half. The second Iobp examined (4695-2) had been ~ fabricated with two straight heating sections rconnected with a U bend (Fig. 5.2), and it had _ operated for 887 hr before belng terminated because ~ of a pump bearing failure. The maximum attack was to a depth of 8 mils (Fig. 5.3), and it was A temperature temperature occurred in the first heated section but that the maximum fluoride mixture temperature occurred in the second heated section. The loop was assembled so that no heating occurred in the bend, and upon exammcmon, no attack was found ‘A thin, as yet unidentified, deposit was found in the final portion of the cold leg. For comparative purposes, a portion of this same batch 83 TABLE 5.1. CONDITIONS OF OPERATION OF FORCED-C}I.RC.ULATION_ INCONEL_ LOOPS‘.‘ Loop Number 4695-1 ' Operating Conditions 4930-A 4935-1 4695-2 4695-3 Fiuvoride mixture circu- NuF-ZrF4-UF4 NcF-Zer-UF4 NaF-ZrF4—~(UF4+UF3) NoF-ZrF4—(UF4+UF3) NaF-KF-LiF+UF4 (mils) lated (53.5-40-6.5 mole %) (53.5-40-6.5 mole %) (50-46-4 mole %, (50-46-4 mole %, (11.5-42-46,5 mole % plus 2.2 wt % of U as U3+) 1.7 wt % of U as U3+) 12.3 wt % UF4) Operating time (hr) 1000 1000 385 887 630 Preliminary operating 24 50 71 43 period at isothermal temperature (hr) Maximum fluoride mix- 1500 1500 1500 1500 1500 ture temperature (°F) Approximate maximum >1710 1680 >1735 > 1670 >1640 tube wall temperature (°F) Temperature gradient of 300 300 300 200 300 fluoride mixture (°F) Reynolds number 1000 1000 10,600 15,000 10,000 Velocity {fps) 0.6 0.7 6.7 10.2 3.3 Length of heated tube (ft) 23.5 First section 3.5 5 5 -5 Second section 3.5 7 7 7 Total length of loop (ft) 14 48 50 50.5 | 50.5 Method of heating Electrical resistance Gas Electrical resistance Electrical resistance Electrical resistance Shape of heated section Coiled Coiled Coiled Straight Straight Ratio of hot-leg surface 1.6 3.1 1.5 1.5 1.5 to loop volume (in.2/in.3) Cause of termination Scheduled Scheduled |_eak Pump bearing failure Leak - Maximum depth of attack 25 11 12 8 35 LY0dIY SSIHD0¥d LDIr0¥d dNV . «} ‘ ”concenfr UNCLASSI‘FIED 2 ; s T Fig. 5.1. Maximum Attack Found in Inconel Loop 4695-1 After Circulating NaF-ZrF,-(UF, + UF,) (50-46-4 mole %, 2.2 wt % U as U3?) for 385 hr at a Maximum Fluoride Mixture Temperature of 1500°F and a Reynolds Number of 10,000, 100X. Reduced 16%. of fluoride mixture was circulated for 1000 hr in an Inconel thermal-convection loop. The hot leg of the thermal-convection loop was attacked to « depth of 10 mils, whereas a depth of 13 to 15 mils was normally found in control thermal-convection foops in which the fluoride mixture contained UF, but no trivalent uranium. However, thermal- convechonr 100ps opemfed prev:ously in wh:ch trwaent uranium was obfumed by adding zirconium hydride were attacked to a depfh of only 3 mils. . The deepest attack found in any forced-circulation ~loop was that found in |06§1"4695‘3 ~ This loop ""rc:rculated the alkoh-metol base mixture NaF KF- LiF ; i‘(” 5.42- 46 5 moie %) wfih 12.3 W’r % Urdmum addedw - S as UF4. “The ‘maximum attack in this loop (de- "signed as shown in Flg 5.2) occurred in the :"\‘second heated sechén as a very heavy mtergranular RS *"fgice voids to a depth of 35 mlls (Flg 5'_4) The 'reinperature pattern of this foop showed 'rhcx'r ‘because of the better heat trcmsfer properhes of fhe ulkah me’ral base mlx-‘:_ ,'_'fures both ‘rhe maximom wall cnd fluorlde mixture temperatures ~ occurred in the second heated section. A deposit that was up to 65 mils thick was found on the lower inside surface of one area PERIOD ENDING JUNE 10, 1955 in the cold portion of the loop. In other areas, similar deposits that were up to 5 mils thick were found. A spectrographic analysis of the deposit showed the following: >5 wt % Cr, 0.5 wt % Fe, >10 wt % K, 3wt % Li, 2wt % Na, 0.5 wt % Ni, > 10 wt % U (balance unidentified}). Another portion of this batch of fluoride mixture was circulated for 500 hr in an Inconel thermal-convection loop. The maximum aftack was to a depth of 42 mils, and was unusual in that it was found only around one third of the circumference of a sample of the hot leg. In the remainder of the hot leg the attack was to a depth of 5 mils. A cold-leg deposit that was 0.3 mil thick was found. Two other forced-circulation loops that had completed the scheduled 1000 hr of operation were also examined. These loops had circulated NaF- ZrF4-UF4 (53.5-40-6.5 mole %). One of these loops (4930-A) was heated by the electrical resistance of a coil,] and the other (4935-1) was heated in a gas furnace. These loops were similar except that the gas-fired loop had a heated length of 23.5 #t, and the heated length of the other loop was only 7 ft. To transfer the same amount of power, it was obviously necessary to use a much higher wall temperature for the short loop. The maximum attack in the short loop (4930-A) appeared as a heavy intergranular concentration of sub- surface voids to a depth of 25 mils and was found in a bend in the second leg of the heated coil. The maximum attack in a straight section was 21 mils. The attack was both deeper and heavier than that found in loops operated previously with this fluoride mixture, even when much higher velocities and Reynolds numbers were used. Examination of loop 4935-1 showed the maximum attack to be to a depth of 11 mils; however, this attack was moderate in intensity and was found for only a short length at the hottest end of the coil. In this loop no ~differences were found on opposite sides of the bends. The original purpose of these loops was to compare gas and electric heating; however, because of the differences in heater length, it will be necessary fo repeat the tests with identical loops -~ -operated under similar conditions. From the results obtained with these loops and from those reported previou’sly,:z it is apparent that ‘corrosion and mass transfer in these systems are not so serious as had been feared. It also appears 2G. M. Adamson and R. S. Crouse, ANP Quar. Prog. Rep. Maf. 10) 1955‘ ORNL']864' P 75. 85 86 ANP PROJECT PROGRESS REPORT HEATED SECTIONS UNCLASSIFIED ORNL-LR-DWG 7414 \m COLD SECTION HEATER CONNECTING LUGS FILL TANK Fig. 5.2. Schematic Diagram of Forced-Circulation Loop with Straight Heated Sections. that reducing the tube wall temperature in these experiments would result in a reduction in depth of attack. The cold-leg layer found in loop 4695-2 must be investigated further. The depths of attack found in the forced- circulation loops are not much, if any, greater than those found in the low-velocity thermal- _ convection loops Thus velocity and Reynolds number appear to be only minor variables. This conclusion is confirmed by the deep attack found in the low-velocity forced-circulation loop 4930-A. The data obtained from the thermal-convection loops should fherefore be applicable to forced- circulation systems The most important variable is the maximum tube wall temperature or, more exactly, the temperature of the reaction interface, This is shown by the ~ deepest attack occurring in areas of maximum wall temperature rather than where the maximum bulk fluoride mixture temperature occurred. Additional evidence is afforded by the deepest attack in a loop having occurred in the hot spots on the inside of bends.? The addition of UF, to the ZrF -base mixtures reduces the attack, and even though the low UF, concentrations used in these forced-circulation loops did not reduce the attack as much as had been hoped, it appears that higher concentrations would result in a greater decrease in attack. Unless the addition of UF; can also reduce the attack and mass transfer found with alkali-metal base mixtures without disproportionation of the UF, and the formation of deposits, it appears that such mixtures will not be useful in large, Inconel dynamic systems. Sodium in Inconel An Inconel forced-circulation loop (4689-4) was examined in which sodium was circulated at a maximum temperature of 1500°F for 1000 hr, with W ”’:“:for 480 he wi UNCLASSIFIED T7643 ! i a - Sl T £ . e . wrF : G - LA i vt ¢ S J = T —t St e s ¥ LW - . . & . o . © . . L ? " ¥ w . . & T - Y, Lk . s : g . L ' . LS . bdl £ - ., ML . & P [ ) v . i cal : + . < i . - - v L v A . i o - i ; - - — L w’ ; - i . 2 . - # . LR - & ¥ .- { * Fig. 5.3. Maximum Attack Found in Inconel Loop 4695-2 After Circulating NaF-ZrF (UF + UF ) (50-46-4 mole %, 1.7 wt % U as US™) for 887 b at ¢ Maximum Fluoride Mixture Temperature of 1500°F and a Reynolds Number of 15,000. 100X. Reduced 11%. a 300°F temperature drop and a Reynolds number ~of greater than 50,000. The loop had the configu- ration shown in Fig. 5.5. A heavy layer of dendritic metal crystals was found in all sections of the economizer and to a lesser extent in the cold loop, as shown in Fig. 5.6. The crystals were shown by chemical analysis to be 89.6% Ni and 8.6% Cr and to contain '50'ppn1 Fe. The layer found in the cold “end of the economlzer was shown by metollogrqphlci 'exommohon to be 26 mils '|'|"IIC|< The hot leg ‘showed mfergranulor attack to a dep'rh of 2.5 mlls.w | - Efforts were made to obtain oxygen onolyses on ' ‘_'both the orlgmol ‘and the dmlned sodlum, but bothf '--f=va|ues are’ quesnonuble. From other work, it : ‘-..appeors fhat fl1e orlgmof sod:um contolned obouf _' | --7-'.‘,:50 PPM oxygen. SR A second snmllar' The maximum thickness of the deposn found metal- jrl"o‘a'p 'k(}195'l l)mwus opercted_fl_;: ith a _200° F temperqture drop, and the':f Cresults conf:rmed 1hose reported above. The moss: ~“fransfer found in this loop is shown in Flg 57. PERIOD ENDING JUNE _10, 1955 UNCLASSIFIED T-7727 | . e : v’ + . o N . T - e : “ . ' * o M - . ‘,. i 3 e ..'. - ., - “\, ,, ; eyl . . Ly LT P : - ..M O r’; + . [ _ - O . e e M > [N 7 o S . . AN L p 3 e T O > - B3 . e e, T . cw LA . St ~ Y w e ¢ ; ‘s i [Py ' . . , Fig. 5.4. Maximum Attack Found in Inconel Loop 4695-3 After Circulating NaF-KF-LiF + UF, (11.5- 42-46.5 mole % plus 12.3 wt % UF 4) for 630 hr at a Maximum Fluoride Mixture Temperai‘ure of 1500°F and ¢ Reynolds Number of 10,000. 100X. Re- duced 20%. fographically was 11 mils. The hot leg in this loop also showed intergranular attack to a depth of 2.5 mils. Two additional loops were operated with portions of the cold legs constructed from type 316 stain- less steel. Loop 4689-6 had a type 316 stainless steel economizer and cold leg, .while loop 4689-5 had only a type 316 stainless steel cold leg. Both these loops also showed mass transfer, but not so much as was found in loop 4689-4. The maximum deposit thickness in loop 4689-5 was 15 mils and —in loop 4689-6 it was 12 mils. The hot legs showed similar intergranular attack to a depth of 2.5 mils in loop 4689-5 and 1.5 mils in loop 4689-6. - These data indicate that a type 316 stainless steel cold leg reduces the mass transfer slightly, but further study of the variables in the process is "j__nnecessory Additional IOOps are bemg operated 7 vunder com‘rolied condmons in which' the effects of oxygen concen’rrohon of 'rhe sodlum the use of a cold frop, the use of c: Iowe-r maximum sodium temperature, and ‘the use of an all-stainless-steel system are being investigated. 87 | ANP PROJECT PROGRESS REPORT . UNCLASSIFIED _ ORNL-LR-DWG 7415 el r SURGE TANK e wlll” Y n Y N 2 PUMP ELECTROMAGNETIC FLOWMETER Fig. 5.5. Schematic Diagram of Loop Designed for Forced-Circulation of Sodium in Inconel Tubing. "~ THERMAL-CONVECTION STUDIES G. M. Adamson . Metallurgy Division - T. C. Price V. P. Treciokas ' - Pratt & Whitney Aircraft 'Alkqlri-Mefal Base Mixtures with UF, and UF in Inconel The results of metallographic examinations have been received for the remaining Inconel thermal- 88 convection loops in which alkali-metal base fluorides with varying proportions of UF,/UF, had been circulated. The results for the first half of this series of tests were reported previ-- ously.® The metallographic data are presented in Table 5.2. The results for the loops with high UF,; content confirm those reported previously. No attack was found in any section, but deposits 36. M. Adamson and V. P. Treciokas, ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 70. PERIOD ENDING JUNE 10, 1955 S ' S UNCLASSIFIED ' ' UNCLASSIFIED . S - o T-7761 - T-7763 Fig.5.6. Sections from Cold Leg and Economizer Fig. 5.7. Sections from Cold Portions of Forced- . of Forced-Circulation Inconel Loop 4689-4 After Circulation Inconel Loop 4951-1 After Circulating Circulating Sodium for 1000 hr at a Maximum Sodium Sodium for 480 hr ot o Maximum Sodium Temper- Temperature of 1500°F and a Temperature Drop ature of 1500°F and a Temperature Drop of 200°F. - of 300°F. TABLE 5.2, RESULTS OF METALLOGRAPH!C EXAMlNATlON OF INCONEL THERMAL.- CONVECTION LOOPS AFTER CIRCULATING NGF KF- lF (” 5-42-46.5 moie %) CONTAINlNG UF AND UF | T oo o Time of , i I,q U3 Content Maximum L_oop . Uranium { %) . : . Operation wt % Attack Metallographic Notes No. h Content : . i1s) : (l’) L wmw F'”, Fmal‘ (mils 51 s00 ns 17 o 2 Thin cél.d-iég doposit 626 ) 500 .2 _ 1.5. 08 R 1.5 Cold- Ieg deposn 0.3 mll thick 594 1000 126 20 06107 2 Cold- ieg depos:f o s mil thick | 625 s46 12 30 09 0 Layerto lmil thick inall sections 593 _ ]000 - 138 - 56 - f_ 0.9 ."0 o Layer to 1 rn||'rhlck in all sections < 592 - 'I-OOOV 109 5.2 - 0.9 -‘ 0 Layer-‘fo i n.}i'! *hick in all sections P 89 ANP PROJECT PROGRESS REPORT were visible in both the hot and cold tegs. The trivalent uranium content had been almost com- pletely lost by disproportionation in all loops. The results for the three loops with low UF, contents are encouraging. In loop 594, after 1000 hr, the hot-leg attack was only to a depth of 2 mils, and there was no evidence of a hot-leg layer (Fig. 5.8). The cold legs of these loops showed thin, as yet unidentified, layers. Even in these loops it seems likely that some dispro- fe———————you1 0400 ‘,21 ’1' :‘,’;: "?/? ‘/, Fig. 5.8. Hot-Leg Attack of lnconel Thermal- Convection Loop After Circulating NaF-KF-LiF (11.5-42-46.5 mole %) Containing UF; ond UF, for 1000 hr at a Hot-L.eg Temperature of 1500°F. 250X. Reduced 21%. UNCLASSIFIED T.7544 portionation occurred, because one-half the origi- nal U3Y content was lost, a Chemical analyses of the fluoride mixtures before and after operation of the loops are pre- sented in Table 5.3. The data show that the impurities react differently in the alkali-metal base mixtures than they do in the ZrF,-base mixtures. The chromium content appears to de- crease slightly, whereas the nickel content shows ~a small increase. Zirconium Fluoride Base Mixtures with UF, and UF, in Inconel A group of Inconel thermal-convection loops was operated at a hot-leg temperature of 1500°F with the ZrF,-base mixture NaF-ZrF4~UF4 (50-46-4 mole %) containing varying amounts of the UF, converted to UF,. Metallographic results have been received on only the first half of this series of tests and are reported in Table 5.4, The chemi- cal results for these tests are given in Table 5.5, While the depths of attack found were slightly lower than the 11 mils now being found after 500 hr in contro! loops, the results are disappointing in that, in previous thermal-convection loops operated with a mixture containing UF, obtained by the reduction of UF, with ZrH,, attacks as low as 3 mils were found. As in the forced-circulation loops, the presence of UF, reduces the amount and intensity of the attack more than it does the depth. In the loops operated previously, the UF, content was not known and may have been slightly higher. 1t should be noted that the chromium content was TABLE 53. RESULTS OF CHEMICAL ANALYSES OF ALKALI-METAL BASE MIXTURES BEFORE AND AFTER CIRCULATION IN INCONEL LOOPS Locop Uro;-aium (wt %) Nicke! (ppm) Chromium (ppm) lron (ppm) Ne. Before After Before After Before After Before _ Aftgr 591 15 1.5 65 100 &0 30 140 75 626 1.2 2 10 110 4 15 to 50 120 85 | | 594 12.6 | 10.7 60 110 115 15t0 ? 70 " '75 B / | 625 13.2 12.5 60 40 40 140 ) | . i 593 s 10.0 45 85 75 15 o 0 | | ' 592 109 10.5 110 95 55 ? 0108 165 0 90 " 0" TABLE 5.4, PERIOD ENDING JUNE 10, 1955 RESULTS OF METALLOGRAPHIC EXAMINATION OF INCONEL THERMAL-~CONVECTION LOOPS AFTER CIRCULATING ZrF ~BASE MIXTURES CONTAINING UF, AND'UF4 L.oop Operating Maximum Metaliographic Notes Time Attack Ne. (hr) (mils) Hot-Leg Appearance Cold-1_eg Appearance 633 500 10 Moderate to heavy general attack and No attack or deposit o intergranular voids 634 500 8 Moderate to heavy intergranular Few voids to a depth of 1 mil with deposit voids covering voids 683 500 7 Moderate intergranular voids Few voids to a depth of 1 mil 635 1000 7 Moderate general attack and inter- Few voids to a depth of 0.5 mil and some _stlll in operatlon. i‘appears to be rea o granular voids thin deposit TABLE 5.5, RESULTS OF CHEMICAL ANALYSES OF ZrF -BASE MIXTURES CONTAINING UF ;, AND UF, BEFORE AND AFTER CIRCULATING IN INCONEL THERMAL-CONVECTION LOOPS Loop U3t (wt %) Total U (wt %) Nicke! (ppm) Chromium (ppm) Iron (ppm) Ne. Before After* B—etore After Before After Before After Before After 633 1.37 0.7 1.8 1.4 25 15 100 570 65 50 634 1.28 0.9 8.8 8.9 35 20 65 300 40 70 683 1.43 0.8 9.0 8.5 25 25 140 230 25 30 60 635 1.98 0.7 8.9 9.7 10 100 225 80 40 *These values showed considerable variation and it was dlfflcult to estimate an average; the hot-leg value was con- sistently lower than the cold-leg value, lower than normal in three of these loops after circulation. Addltional |oops in this series are data obtamed in these experlments are tabulated in Table 5 6. |t is not completely understood why“ ' :attack is deeper “at IQOOOF than at 1350°F but th’lS effect has been noted m prev:ous tests and : eal. The'r remammg lToops show a - definite incredse in depth of attack with increasing :hot Ieg temperaturew In |oops operated previously under similar conditions for 500 hr, no effect of temperature on the depth of attack was found. Since the mass transfer effect is masked by fhe effect of impurities and nonequilibrium conditions during the first 500 hr of operation, it may be " considered that the increased depths of attack with L3 o | o7l increased te tures, in the | ted f - Effect of Temperature on Muss Transfer _, {;00 ?1r weremcf::;osefl by n;:ss :q::fei operated for L A group Of |nconel ‘°°P5 ‘was operated mthsww " NaF- ZrF -UF (53 5-40-6.5 mole %) for 1500 hr at f___,-,‘;,_.;"'varlous hot-leg temperatures m order to determine - the effect of temperature on mass tronsfer The Effect of Loop Size and Shape on Corrosion A series of Inconel loops was operated with - NciF-Zer-UF4 (50-46-4 mole %) to determine the " effect on depth of attack of varying the loop size ‘and shape. These loops all operated for 1000 hr with hot-leg temperatures of 1500°F. The data ‘obtained from these experlments are presented in “"Table 5.7. Within the accuracy of the data, it does not appear that varying the length of the horizontal leg is any more effective than varying the vertical leg. For special materials in limited supply, it appears from the data that the loop size may be N i e e ANP PROJECT PROGRESS REPORT 'TABLE 5.6. EFFECT OF HOT-LEG TEMPERATURE ON MASS TRANSFER IN INCONEL THERMAL-CONVECTION LOOPS CIRCULATING NoF-ZF .UF , (53.5-40-6.5 mole %) Operating time: 1500 hr Meiu!logruphic'Nofe.s. o L 'Hrcz>‘t-'Leg' - AT Final Chromium Maximum o NODP Temperature °F) Concentration Attack Hot-Lea Attack Cold-Leg e (VOF) (ppm) (mils) otr--eg Aftac Appearance n 578 © 1200 _ 165 610 to 640 8 Heavy witH small voids A -N(‘)'vi‘s ible defiositl 579 1350 ‘ | 190 690 to 815 5 Moderate to heavy No visible deposit o R : intergranular voids - _ » 583 ' ]350'- | 195 675 to 775 5 Moderate to heavy | " No visible deposit e T o ' intergranular voids ' 580 1500 ~210 615 to 830 12 Heavy intergranlar No visible deposit o ' : voids o | 584 ~ 1500 ' _ - 210 955 to 980 15 Heavy intergranular Very Iigh.f'dépc':s it e voids 585 - 1600 7 240 540 to 810 18 Heavy infergranu‘ar " Metal i'n cold tr..ap ‘ , o voids TABLE 5.7. EFFECT ON CORROSION DEPTH OF VARYING THE SIZE AND SHAPE OF INCONEL THERMAL-CONVECTION LOOPS ' Circulated fluid: Nc:F-ZrFA-UF:4 (50-46-4 mole %) Operating time: 1000 hr Hot-leg temperature: 1500°F . _ Length of Length of Maximum Final Loop Vertical Horizontal AT Depth of Chromium Hot-Leg “ No. Leg Leg (°F) Attack Content Attack {in.) (in.) {mils) (ppm) 601 8 8 143 10 620 to 680 Heavy 606 8 8 145 10 530 to 675 Heavy 598 15 8 172 11 1030 to 1075 Moderate to .heavy 603 15 8 188 10* 740 to 965 Heavy 600 26 8 220 16.5 980 to 1040 Heavy 605 26 8 218 16 890 to 9215 Ht_ea\fy . 597 s 15 208 10 895 to 1040 Moderate to heavy o602 15 15 195 13 890 to 960 Heavy 599 15 26 195 15 710 to 1030 Moderate to heavy 614%* 26 17 245 15 835 to 890 Heavy ) . : *Inr‘t‘h'i's loop:, uryery few boundaries were attacked to a depth of 14 mils. E **Cofifiéh standard size, shape, and temperature differential. 7 . . 2 92 1 n reduced to 15 by 15 in., at least for screening experiments. A series of loops of this size will be operated to get a better statistical picture for determining whether all the loops should be reduced to thls size. ' Evoluanon of Control Loop Results A gradua! increase in depth of attack of the ~standard Inconel thermal-convection loops operated ~with ZrF ,-base mixtures has been noted recently, and difficulty has been experienced in reproducing results. As part of a study of this difficulty, three Inconel loops were filled the same day from the same batch of fluoride mixture and operated for 500 hr in as nearly an identical manner as possible. The data obtained from these experiments are given in Table 5.8. The attack was again deeper than in previously operated standard loops, and the data show no reproducibility. It should also be noted that in each ]oop some cold-leg a’rtack PERIOD ENDING JUNE 10, 1955 was found, whereas previously the cold legs appeared to be attack-free. To - determine whether the increase in attack could be caused by a leak during operation or by air blown into the fluoride mixture during filling, an additional series of three loops was operated. In one experiment, 1 liter of air was slowly bubbled through the fluoride mixture in the fill pot before transfer to the loop. In the second experiment, the loop was airtight, but it was operated with the helium atmosphere at atmospheric pressure. In the third, the helium atmosphere was again at atmos- pheric pressure, and, in addition, a Swagelok connection above the loop was loosened slightly so that there was a slight air leak. The data from these loops are presented in Table 5.9. A fourth, standard loop was operated as a control with the normal helium atmosphere at a pressure of 7 psi. The data indicate that to get increased attack from contamination during operation would require _ TABLE_S.’_B. RESULTS OF OPERATION OF IDENTICAL INCONEL CONTROL LOOPS . Fi!lifig Maximum Metallographic Notes Average Final Loop No. Attack Chromium Content ‘ Order (mils) Hot-Leg Attack Cold-Leg Attack (opm) 684 | 3 ' N Moderate to heavy intergranular Light, general, to a 850 ' voids depth of 1 mil 685 2 14 Moderate to heavy intergranular Very light, general, to - 750 ’ - voids ‘ a depth of 1 mil ' ] 686 o - 9 Moderate to heavy intergranular Very light, general, to 770 voids a 'depfh of 1 mil Operotlng t:me L Hot Ieg femperofure 1500 F B - 500” hr SREERL TR " Maximum Attack Loop No " Variable Chromium Content s s T - {mils) ) - ~ SR (pom) 687 10 o 800 - 689 ) 4 t_,ro;Jg oring'mal fluortde mlx’rure ?flfi: A "9- o 950 T i beFore tronsfer fo loop ‘ ' o ST e 698 : Control {7 psi heliuvm pressure) 10 775 93 Average Final k ¢ ! ¥ T I T T T ANP PROJECT PROGRESS REPORT both an air leak and a loss of helium pressure, ‘a combination which does not seem likely to occur. It also appears that the small amount of air that could be trapped in a transfer line would be enough to cause difficulty., One other possible source of the increased attack may be the cleaning cycle. Therefore, a group of loops cleaned by various procedures have been operated, but the results ‘are not yet available, . | 'G'ENERIAL CORROSION STUDIES _ E. E. Hoffman 7 W H. Cook _ C. F. Leitten, Jr. L Meta”urgy DlVlSlon Bruzmg Alloys on Type 310 Stainless Steel and "A" Nickel in Sodium and in Fuel Mixtures quzmg alloys ‘'submitted by the Wall Colmonoy Corporation have been tested in both static sodium and in static NaF-ZrF ,-UF, (53.5-40-6.5 mole %). " These tests were conducted in an effort to find a brazing alloy that has good corrosion resistance to both mediums. The tests were conducted on type 310 stainless steel T-joints brazed with the brazing alloys listed in Table 5.10, which also presents the resulits of exposure to static sodium at 1500°F for 100 hr. The brazing alloys are listed in order of decreasing NICKEL PLATE corrosion resistance. The results for tests of these alloys in static NaF-ZrF ,-UF, (53.5-40-6.5 mole %) are given in Table 5.11, where, as in Table 5.10, the brazing alloys are listed in order of decreasing corrosion resistance. A comparison of the results in Tables 5.10 and 5.11 indicates that brazing alloy B-13 (9% Si- 2.5% P-88.5% Ni) has the best corrosion re- sistance to both test mediums. Metallographic examination showed no attack along the surface of the braze fillet when tested in sodium (Fig. 5.9a). The cracks shown in Fig. 5.9z are not the results of corrosion, but rather were caused by the brit- tleness of the alloy. Similar cracking was" ob- served in many of the brazed T-joints listed Tables 5.10 and 5.11. Brazing alloy B-13 is shown in Fig. 5.95 after being tested in NaF-ZrF ,-UF, (53.5-40-6.5 mole %) for 100 hr at 1500°F. Surface attack to a depth of 1 mil can be seen along the entire brazed fillet. No cracks caon be observed in this joint. Several conclusions can be drawn from a study of the metallographic notes in Tables 5.10 and 5.11. Brazing alloys containing relatively high per- centages of phosphorus appear to be inferior in sodium. However, additions of silicon tend to improve the corrosion resistance. On the other hand, brazing alloys having high percentages of NICKEL PLATE Fig. 5.9. Type 310 Stainless Steel T-Joints Brazed with Brazing Alloy B-13 (9 wt % Si, 2.5 wt % P, 88.5 wt % Ni) After Exposure for 100 hr ot 1500°F to () Static Sodium and () Static NuF-ZrF -UF (53 5-40-6 5 mole %). Etched with aqua regia. 200X. Reduced 26%. s ar PERIOD ENDING JUNE 10, 1955 TABLE 5.10. BRAZING ALLOYS ON TYPE 310 STAINLESS STEEL ' TESTED IN STATIC SODIUM AT 1500°F FOR 100 hr Alloy Alloy Composition Weight Change . . : —_— tall hi t Designation* (wt %) o) ) Metallographic Notes B-13 9 Si—-2.5 P—-88.5 Ni 0 0 ' No attack along surface of fillet; . several cracks in fillet P-11 10.5 Si~7.5 Mn—82 Ni ~ —0.,0002 -0.018 No attack along surface of fillet; _ : : several cracks in fillet P-12 9 Si-15 Mn=76 Ni -0.0003 ~0,033 No attack along surface of fillet; - : several cracks in fillet P-10 _ . 16.27 Si~5.9 Mn-77.83 Ni +0.0004 +0.043 No evidence of attack; several ' ‘ large cracks throughout fillet P-13 7.5 $i-22.5 Mn-70 Ni -0.0008 -0.085 Erratic surface attack to a depth of . o 0.5 mil; several cracks in fillet B-12 © 10.5 Si—1.25 P—88.25 Ni 0 0 Erratic surface attack along fillet S i ; . o to a depth of 1 mil B-14 7.5 Si—-3.75 P—88.75 Ni -0.0001 ~0.01 Uniform surface attack along fillet . ‘ _ ' S to a depth of 1 mil P-14 6 Si—30 Mn~64 Ni —0.0002 -0.022 Erratic surface attack to a depth of ' 1 mil; large cracks throughout o S : R Lo fillet ' G-20 9 P-1 ].49 W-.79.51 Ni -0.0012 -0.14 Maximum attack of 4 mils along - - : ' ' surface of fillet - $-10 14, 31 Cr—9.34 $i-2.56 0 0 Subsurface voids to a depth of 3 T Mo=19.32 Fe-54.37 N: mils along fillet surface B-15 6 Sl—5 P 89 Ni ' 0 0 Maximum attack of 4 mils along L N S : surface of fillet B-16 4.5 51—6.25 P—~89.25 Ni ~0.0001 ~0.01 Small subsurface voids to a depth o o L foT . of 4 mils G-21 ' 10.98 P—-6.16 W—82.86 Ni ~0.0001 -0.0M Uniform surface attack ciong entire o o o : fillet to a depth of 4 mils B-17 38i-7.5 P-89.5 Ni -0.0012 -0.14 Maximum ettack of 4 mils along CL L o : oo surface of fitlet L-20 38 Ni-5 Cr-57 Mn —-0.0004 —-0.043 Attack in the form of stringers to a v . maximum depth of 5 mils; not uniform *quzing alloys listed in order of decreasing corrosion resistance to sodium. CEpT fncke| _’_suhcon in the absence of phosphorus tend to ‘be mah‘acked in the fluoride mixture, ' the same bruzmg ollcy after bemg ‘tested - ';Flg 5,108 No qflfcck can be seen’ along the e conducted on S V'hove' fair corrosion resnstance fo bofh medla are '..:a hot-zone femperafure&of 1500°F A femperature " the 90 wt % Ni<10 wt % P alloy, the 8O wt % Ni— dlfferenhal ‘of ‘about 400°F was maintained in all 10 wt % Cr~10 wt % P alloy, and Nicrobraz. A 1.5- the seesaw tests, mil layer of small subsurface voids can be seen in All the brazing alloys listed in Table 5.14 had 95 _-|0|m‘ brazed with Nlcrobraz and tested in sodlum, NaF;ZrF'--UF4 (53.5-40-6.5 mole %) is shown in and the reéults are glven - pbged ifor 900 hr ot i o i ANP PROJECT PROGRESS REPORT TABLE 5.11. BRAZING ALLOYS ON TYPE 310 STAINLESS STEEL TESTED IN STATIC NaF-ZrF ~-UF , (53.5-40-6.5 mole %) AT 1500°F FOR 100 hr Alloy Alloy Composition Weight Change . : ~ Metallographic Notes Designation* (wt %) (g) (%) G-20 9 P-11.49 W~79.51 Ni -0.0017 -0.152 No attack on surface of fillet;- - ' : several cracks in fillet B-15 6 Si—-5 P—-89 Ni —~0.0046 -0.510 Uniform surface attack along fillet o _ _ ’ to a depth of 0.5 mil B-16 4.5 Si—6.25 P-89.25 Ni -0.0057 ~0.562 Surface attack along fillet to a s A . depth of 0.5 mil G-21 10.98 P-6.16 W—-82.86 Ni -0.0061 ~0.674 Erratic surface attack to a depth of e 0.5 mil along fillet B-17 3 8i-7.5 P-89.5 Ni —~0.0053 —-0.516 Surface attack along fillet to a ' Cee ; depth of 0.5 mil; several cracks L e e _ in fillet . _ ‘ B-13 ' 9Si-2.5 P—88.5 Ni —0.0027 —0.279 Surface attack along entire fillet to R o a depth of 1 mil P-12 95Si-15 Mn—76 Ni ~0.0034 ~0.331 Surface attack in form of small . , ‘ _ ' voids to a depth of 1 mil; several e - - cracks in fillet ‘ 7 B-14 7.5 Si-3.75 P~88.75 Ni —~0.0052 ~0.479 Surface attack along entire fillet to C : : ~a depth of 1 mil P-11 10.5 Si=7.5 Mn~-82 Ni -0.0042 ~0.401 Surface attack to a depth of 1.5 , mils; several cracks in fillet B-12 10.5 5i-1,25 P-88.25 Ni -0.0018 -0.171 Surface attack to a depth of 2 mils; large cracks throughout fillet P-10 16.27 Si~3.9 Mn—-77.83 Ni —~0.0026 -0, 290 Uniform surface attack along fillet to a depth of 3 mils S-10 14.31 Cr—9.34 Si-2.66 —0.0030 -0.344 Surface attack to @ maximum depth Mo—19.32 Fe-54.37 Ni of 7.5 mils along entire fillet P-13 7.5 5i~22.5 Mn=70 Ni —0.6043 -0.455 Complete attack of entire fillet L-20 38 Ni—5 Cr—=57 Mn -0.0078 -0.74 Complete attack of entire fillet P-14 6 S$i-30 Mn—64 Ni -0.0099 -1.06 Complete attack of entire fillet *Brazing alloys listed in order of decreasing corrosion resistance to the fluoride mixture. g Y g fair corrosion resistance to both sodium and the fluoride mixture, with the possible exception of the 65 wt % Ni~25 wt % Ge~10 wi % Cr alloy. The Coast Metals No. 52 alloy had good corrosion resistance to the fluoride mixture in the seesaw test in contrast to its poor resistance in the static test, A large degree of porosity was present in Coast Metals No. 50 alloy that hampered the evaluation of the corrosion data. These two brazing alloys will be retested in order to check the results of these seesaw tests, | Scféénfng Tests of Solid-Phase Bonding . The a”bys which have been proposed for use in the fabrication of the ANP reactor fuel and coolant lines have a tendency to form a solid-phase bond at elevated temperatures in liquid metals and in 96 fused fluoride salts. Bonding of this type makes them unsuitable for use in valves, bearings, and seals; thus cermets and ceramics that do not form a solid-phase bond are being investigated. The cermets (metal-bonded ceramics) appear to be the more promising because of their high corrosion resistance and other chemical and physical characteristics. The test apparatus being used is shown in Fig. 5.11. This apparatus was adapted from that designed for lever-arm stress-rupture tests.® The container for the central components, for the test specimens, and for the liquid metal or fused fluo- ride salt serves as a housing for the compression tube and upper platen. By using this arrangement 4R. B. Oliver et al., ANP Quar. Prog. Rep.. Mar. 10, 1955, ORNL"‘864, Fig- 702, P 105, el iac i °% O PERIOD ENDING JUNE 10, 1955 TABLE 5.12, BRAZING ALLOYS ON *‘A" NICKEL T-JOINTS TESTED IN STATIC NaF-ZrF ~UF (53,5-40-6.5 mole %) AT 1500°F FOR 100 hr Alloy* Composition Weight Change (g) (%) Metallographic Notes (wt %) 80 Ni-10 Cr—-10 P 0 0 No attack on braze fillet 0 0 No attack on braze fillet 50 Ni—-25 Mo—-25 Ge Nicrobraz 69 Ni-5 B~15 Cr~5 Si~5 Fe-1C Electroless nickel 90 Ni—10 P 65 Ni—25 Ge—10 Cr Coast Metals No. 52 90 Ni—-4 B—4 Si-2 Fe General El..ectric No. 81 66 Ni—10 $i—19 Cr—4 Fe—1 Mn 35 Ni-=55 Mn=10 Cr 60 Mn—40 Ni 68 Ni—32 Sn -0.0004 -0.016 -0.0004 -0.013 -0.0014 ~0.05 -~0.0003 —0.012 —-0.0111 ~0.48 —-0.0159 -0.59 —0.0998 ~3.49 Mo attack on braze fillet No attack on braze fillet Small subsurface voids to a depth of 0.5 mil along braze fillet Nonuniform attack to a depth of 6 mils along braze fillet Nonuniform attack to a depth of 12 mils along fillet Complete attack of braze fillet Complete attack of braze fillet Joint partially dissolved ot fillet surface *Brazing alloys listed in order of decreasing corrosion resistance to the fluoride mixture, TABLE 5,13, BRAZING ALLOYS ON *‘A” NICKEL T-JOINTS TESTED IN STATIC SODIUM AT 1500°F FOR 100 hr Brazing Alloy* Composition (wt %) Weight Change Metallographic Notes Electroless nickel 90 Ni-10 P Coast Metals No. 52 90 Ni-—4 8—4 Sl 2 Fe . 80 Nl-]O Cr—-'IO P . Gen‘erdl E'[ecfr'lc No 8] 66 Ni-'lO Sl-'|9 Cr--4 Fe—l Mn "N:crobraz & ~© 69 Ni=5 B-15 Cr-5 5i-5 Fe-1C [ '50 Nl-25 Mo-25 Ge' - o 65 Nl_zs c;e..lo Cr e Mn—40 N. - s unelo 68 Ni-32 Sn -0.0171 (g) (%) ~0.0004 —0.018 ~0.0019 —0.068 ’"°-°.°.°'.‘,_7 ~0.08 S ~0.0022 -0.082 _0.0009 —0.036 \_ ‘"Q‘-__OZ_O_'#-' | ':—o 085 *00005 | _’%0;620 | ~0.540 No attack along fillet surface Surf;:!ce attack along fillet to a depth of 0.5 mi Nonuniform attack along fillet to a depth of 1 mll Attack on surface of hllef to a depth of 1 mil 1.5 mil layer of small subsurface voids along Tillet edge Surf?ce aftock along fallet to a depth of 2.5 mils Unlform surface attack along flllef to a depth of 3 mlls Unlform qtrqck ulong entire fliiet to g depth of 9 mlls Small voids in from surfuce' of fillet to a depth of 13 mils Complete attack of whole fillet *Brazing alloys listed in order of decreasing corrosion resistance to sodium. 97 ANP PROJECT PROGRESS REPORT TABLE 5. 14, BRAZING ALLOYS ON INCONEL T~JOINTS EXPOSED IN SEESAW APPARATUS TO SODIUM AND TO NaF-ZrF UF4 (53,5-406.5 mole %) FOR 100 hr AT 1500° F ' Bruzing Alloy* Composition _ Bath Materiol vt %) _ Weight Chonge Metallographic Notes (9) (%) " Coast Méfals No.-:'52‘ _ Fluoride mixture 89 Ni-5 Si~4 B-2 Fe 7 Sodium Low-melhng 'I\:Ii.crol':raz V Fluoride ;.nixture - 80 Ni-5 Cr-'-_6 Fe—3 B-5Si-1C Tl Sodium .Céésf"Méfa'].s l\:.lo.'SO ' Fluoride mixture ] __93 Ni-3.5 $i-2.5 B-1 Fe Coadln » Sodium 70 Ni--r-}lr;l_ Cr—6 Si—13 Ge .Fluoride mixture Sodium Nicrobraz Fluoride mixture 70 Ni—14 Cr—6 Fe~-5 B-4 Si--1 C Sodium 65 Ni—-25 Ge—10 Cr Fluoride mixture Sodium ~0.0008 -0.052 N.Vonu‘n‘ifo.rm‘ .s'firfcce'fif.foé-k ' along fillet to a depth of 0.5 mil ~0.0011 ~0.073 No attack along surface of fillet ~0.0008 ~0.063 Nonuniform surface attack to a depth of 0.5 mil along fillet Subsurface voids to @ maximum depth of 1.5 mils along surface : of flllef —0.0007 —0.051 ~0.0014 ~ —0.085 Uniform surfuce aficlck ulong f:!lef to a depth of 0.5 mil ~0.0012 ~0.077 'Very erratic surface attack olong ' co flllef to a depth of '| 5 mlls —0.0011 ~0.067 Nonumform attack to a dep’rh of : 1.5 mils along surface of fillet —0.0023 —0.139 Nonuniform attack along surface of fillet to a depth of 2.5 mlls ~0.0005 —0.030 Errahc surfcce attack olon? fillet to a depth of 1.5 mi Very erratic stringer attack to a maximum depth of 4 mils along surface of fillet ~0,0010 —0.056 Stringer-type attack to a maxi- mum depth of 4 mils in a few localized areas -0.0019 -~0.113 intermittent surface attack to a maximum depth of 4 mils along fillet *Brazing alloys listed in order of decreasing corrosion resistance to both test mediums. the test specimens can be brought into mutual compression between the platens, Corrosion-resistant cermets and hard-facing alloys have been tested in an apparatus fabricated from Inconel for solid-phase bonding in NaF-ZrF ,-UF, (53.5-40-6.5 mole %) for 100 hr at 1500°F. The test specimens were dimensionally the same (+0.0002 in.) and the contacting surfaces had roughnesses of less than 10 pgin. The test was . begun by holding the contacting surfaces of the - test specimens apart while the apparatus was flushed with NaF-ZrF -UF, (53.5-40-6.5 mole %) at 1500°F This was done to ensure that any surfczce films on the specimens would be removed prior to the application of the compression load. A fresh charge of NaF-Z¢F 4-UF, (53.5-40-6.5 "mole %) at 1500°F was ’rhen put into the test . 98 150,000 psi. chamber, and the test specimens were pressed together at the desired contact pressure. At the conclusion of the 100-hr test period, the NaF-ZrF - UF, was removed, and the system was cooled to room temperature and disassembled. ' The contact surfaces of the test specimens were examined with a low-power microscope for signs of solid-phase bonding. The seating was not perfectly uniform in any of the tests, and therefore the contact pressures between the test specimens were probably in excess of the calculated values in certain areas. Table 5.15 is a summary of the solid-phase-bonding screening tests made at calcu- lated contact pressures of 6,600, 10,000, and In the most severe tests, that is, those in which the calculated contact pressure was 50,000 psi, no bonding was observed, except L ! i M 5, i i i 5 g i g i for a slight amount in the tests of K150A vs K152B - and K162B vs K162B. Since uniform contact was extremely difficult to obtain, the slight bonding that occurred in these tests may have been caused DR ‘ UNCLA_ssu;;ED Y-14 0.01 PERIOD ENDING JUNE 10, 1955 by contact pressures in excess of 50,000 psi. This explanation seems even more reasonable when it is considered that the quantity and compo- sition of the binder metal of the test specimens, UNCL ASSIFIED Y.15144 BNICKEL PLATE 0.03 0.04 Fig. 5.10. *“A” Nickel T-Joints Brazed with ay () aqua regia, 100X. Reduced 24%. TABLE 5.15. %) Compositions: Nicrobraz Afief Static Sodium and (b) Static NaF-ZrF,-UF, (53.5-40-6.5 mole %). Exposure for 100 hr at 1500°F to () Etched with (a) 10% oxalic acid and RESULT'S'OF SOL'!D-PHASE'-BON.DING SCREENING TESTS OF VARIOUS CERMETS AND ALLOYS EXPOSED TO NuF-ZrF +UF, (53.5-40-6.5 mole %) AT 1500°F FOR 100 hr AT VARIOUS CONTACT PRESSURES K150A (80 wf % TiC=10 wt % NbTaTnC -10 wt % Ni) K151A (70 wt % TiC—10 wt % NbToT|C ~20 wt % Ni) _K'|52B (64 wt % T:C 6 wt % NbTuTtC -30 wt % Nl) £ 50,000 KI50A vs K150A° Contact S Contact ‘Pressure on Pressure Contacting Specimens’ Results* (psh) (psi) 50,000 K150A vs K151A No bonding K150A vs K152B Some bonding K150 A vs K162B No bonding K151A vs K151A No bonding K151A vs K152B No bonding K151A vs K162B No bonding K152B vs K152B Ne bonding K152B vs K162B Ne bonding K162B vs K1628 Some bonding *There was inconsistent bonding of test specimens to the supporting Inconel platens in all tests. 99 0oL PARTIALLY ASSEMBLED DISASSEMBLED Fig. 5.11. Apparatus for Solid-Phase-Bonding Screening Tests. COMPRESSION TUBE PULL ROD UPPER PLATEN TEST SPECIMENS LOWER PLATEN ——— 7] UNCLASSIFIED 12:{ LEVER BELLOWS HELIUM ATMOSPHERE IS MAINTAINED ABOVE LIQUID METAL. ( ATMOSPHERE ATTACHMENTS NOT SHOWN ) ~—— MUFFLE OF ELECTRIC FURNAGE ORNL-LR-DWG 7514 LEVER LOAD—/ \LIQUID METAL LEVEL 140d I SSTYD0¥Ud L1I3rodd dNY > 90 . particularly that for K150A vs K152B, show no apparent relationship to the occurrence of the bonding. On the basis of these datq, it is believed that none of these Kentaniums would form solid- phase bonds to each other in NaF-ZrF,-UF, (53.5-40-6.5 mole %) at 1500°F in 100 hr if the true contact pressure between the cermets did not exceed 50,000 psi. A recheck of the pairs that did bond is planned that will conclude the screen testing of the Kentaniums, K150A, K151A, K152B, and K162B, for solid-phase bonding in all contact combinations with each other in NaF-ZrF4-UF4 (53.5-40-6.5 mole %) at 1500°F for 100 hr, Sodium in Inconel A 1000-hr test was recently completed in which sodium was circulated in an Inconel thermai- convection loop. Samples of the sodium used in this test were analyzed and found to contain approximately 0.03% oxygen. The oxygen content was admittedly high and could have been reduced Fig. 5.12. -‘Sfie;:iméns of Héf and Cold Légs of an Inconel Thermal-Convection Loop After Ciréulating HOT LEG ( 1500° F) " PERIOD ENDING JUNE 10, 1955 by a factor of 10 by adequate cold trapping; however, this test did show how serious that mass transfer can be even in a thermal-convection loop when the oxygen content is high. The hot leg of the loop was held at a temperature of 1500°F, and the coldest section of the cold leg was held at 1200°F. The mass transfer that occurred was concentrated in an area in the cold leg where an air blast had impinged on the tube wall during the test, as shown in Fig. 5.12. Very little mass transfer was detected in other sections of the cold leg. A similar test will be conducted in which the oxygen content of the sodium will be held below 0.005% to determine the effect the oxygen content has on mass transfer in a sodium-Inconel system. Lithium in Type 347 Stainless Steel Lithium was circulated in two stainless steel thermal-convection loops for periods of 1000 and 3000 hr, respectively. The hot- and cold-leg temperatures were 1000 and 550°F, respectively. UNCLASSIFIED Y-15138 Sodium for 1000 hr. Crystal deposition resulting from mass transfer may be seen in the cold leg. 101 G ANP PROJECT PROGRESS REPORT The loops operated satisfactorily during the test periods, and macroscopic examination revealed no mass-transferred crystals in the loops or in the lithium drained from the loops. - The hot zone of the loop that operated for 1000 hr had subsurface voids and a ferritic surface layer to a depth of 0.3 to 1.0 mil. Lithium metal had penetrated to this depth. The weld zone of the pipe in this area was attacked to o depth of 3 mils. The cold leg of the loop was unattacked; however, there were a few small (0.2 mil) crystals attached to the surfdce, Similar crystals were previously identified as carbides. ‘The attack in the hot leg of the loop that oper- ated for 3000 hr was similar to that found in the 1000-hr test specimen; however, the spongy ferritic surface layer was 1.0 to 1.5 mils thick, as shown in Fig. 5.13. The attack in the weld zone wos intergranular and extended to a depth of 4 to 5 mils, ~as shown in Fig. 5.14. The preferential attack in o the weld zone is due to attack of the grain boundary carbides by the molten lithium, A deep groove in the pipe wall may be seen at the weld zone—parent metal interface. The cold-leg section (Fig. 5.13) of this loop was very similar in appearance to that of the loop oper- ated for 1000 hr. A few small (0.2 mil) carbide particles were attached to the wall of the tube. It is believed that the corrosion resistance of type 347 stainless steel to lithium in this tempera- ture range would be improved by lowering the carbon content of the steel and by using seamless instead of welded pipe. The austenite-to-ferrite transformation detected in the hot legs of these loops is attributed to leaching of nickel from the type 347 stainless steel by the lithium, Versene Cleaning of Beryllium-Inconel Systems Versene has been proposed as the cleaning agent for the ART, and therefore the corrosion resistance " - . .‘ ?fi&??\g: ;.J.‘ .. ‘ o ) E o UNCLASSIFIED ! NICKEL PLA‘TV , 001 . NIKEL AE Y-14937 T Q = 0.02 0.03 0.04 - ';Fig.; 513. "Hot- and Cold-Leg Surfaces of Type 347 Stainless Steel Thermal-Convecfib.fi..Loéb ‘After Circulating Lithium for 3000 hr at a Hot-Leg Temperature of 1000°F and a Cold-Leg Temperature of 550°F. Note austenite-to-ferrite phase transformations which occurred on hot-leg surface (a) and small o " crystals deposited on cold-leg surface (b). Specimens nickel plated after testing. Etched with a_qpc .- regia. 1000X. PERIOD ENDING JUNE 10, 1955 S UNCLASSIFIED Y-14933 S0°0 00 0 =z O €I 20’0 YO'O Fig. 5. 4. lhei;ie wellfief Hef-Leé Wel“l'.l’ from Type 347 S-t(.:.i-niess Steel Therlr-;fi‘a]‘ (::oev'ecfio'n Loop. After Clrculutlng Lithium for 3000 hr ot a Hot-Leg Temperature of 1000°F, Note deep (4 mils) attack in weld zone as compcred wfih 1- m|| of’rock on parem‘ mefol of berylhum 1rodync:mlc Versene was stuched Thefi"s testing procedure was designed to simulate the ~cleaning procedure used in the ARE, whlch was clecned wn'rh a Versene solution. ple was placed within the insert was 3/ in. OD, / in. 1D, and 3 in. in length. A ]% Versene‘ soluhon was pumped fhrough each'_ pump. " Disodium’ versoncn‘e was used, and the _concentrafion’ was determined on a welghf-vo[ume basis. The fesfmg temperature for both loops was 180°F, and the Versene was circulated in each for a period of 24 hr. comp!efed with’ fwo loops ‘con- 2| ‘with a berylhum msen‘. “The _ I : mloop‘-'fl between ~ two 5|ml|ar lnconel inserts whlch were__jf:_" held in place by crimpmg’ The tubular berylhum):"' Efched with qqua regla. TOOOX Upon complehon of the flrs‘r ‘test fhe Versene was drained while the loop was still at test temperature. When the loop was sectioned, a small amount of Versene was found in a static region of " the loop located in the annular gap between the _beryll_:um_ insert and the Inconel sleeve. No trace of retained Versene was found in this section of the second |oop, ‘which’ wcs flushed ‘with dls’nlled ““water after the Versene’ was drained. Macroscoplc examination showed the berylhum inserts in both loops to be quite similar. Each insert’ had retained its orlgmal polished ap- pearance. No effect of the Versene solution was "found by macroscopic or metaliographic exami- nation of the inner surface of the beryllium insert or the Inconel in either test. The outer surfaces of the beryllium inserts, however, were attacked 103 ~ANP PROJECT PROGRESS REPORT by the Versene solution. The attack was in the form of erratic pits that varied in depth from 0.5 to 2.5 mils, UNCLASSIFIED Y-15010 INCH 1 [ ] | J Fig. 5.15. As-received {(a¢) and As-tested (%) Beryllium 1Insert from Second Inconel Loop in Which Versene Cleaning Solution Was Circulated at 180°F at a Rate of 2 gpm. I INCH 0.02 003 0.04 The erratic attack of the Versene on the outer surface of the beryllium insert used in the second test is shown macroscopically in Fig. 5.15. The unattacked inner surface and the erratically at- tacked outer surface of the beryllium insert used in the first test are shown in Fig. 5.16. The 2.5-mil outer surface attack shown in Fig. 5.16 is representative of that found by metallographic examination of the beryllium inserts used in both tests. In each test the beryllium insert lost weight, as shown in the following tabulation. Test No. 1 . Test No. 2 Original weight, g 16.8715 16.7970 Final weight, g 16,8680 16.7941 Weight loss 9 0.0035 0.0029 % 0.021 0.017 A chemical analysis of a portion of the Versene solution used in the second test revealed a beryl- fium concentration of 0.0024 mg/ml. Since 1.5 liters of 1% Versene solution was used in this test, the total amount of beryllium in the solution was 3.8 mg, which agrees fairly well with the 2.9-mg weight loss of the beryllium insert used in this test. UNCLASSIFIED Y-14755 .V'Fig. 516 " Inner (@) and Outer (b) Surfaces of Beryllium Insert From First Inconel Loop in WHich Versene Cleaning Solution Was Circulated at 180°F at o Rate of 2 gpm. Etched with oxalic acid. 100X. R‘edu ced _24%.‘ 104 a) n " FUNDAMENTAL CORROSION RESEARCH G. P. Smith Metallurgy Division Film Formation on Metals Jo Vo CthCGrf Metallurgy Division Most of the work done in the past on the oxida- tion of metals has been devoted to a study of the oxidation characteristics of the ‘‘heavy’’ metals and alloys such as copper, nickel, iron, aluminum, the stainless steels, etc, These materials either were structurally important or their physical proper- ‘ties were such as to make them particularly amen- able to oxidation studies. Comparable investigations of the alkali and alkaline-earth metals are almost entirely lacking. Rather elaborate oxidation theories have been devised for the heavier metals, but it has been assumed, in general, that the alkali and alkaline- earth metals exhibit a linear oxidation rate, in accordance with the old crack theory of Pilling and Bedworth.® Their theory was that a metal, for which the ratio of the density of the oxide to that of the parent metal is greater than unity, should obey a linear oxidation law. It was reasoned that the oxide film formed would be highly subject to cracking and that, consequently, there would be a constant re-exposure of fresh metal surface to oxygen. Thus the rate of oxidation should be independent of the thickness of the oxide film, the quantity of oxide formed being directly proportional to the time of exposure to oxygen, In order to test the crack theory of oxidation directly, as well as to obtain experimental data on the oxidation of the alkali metals, an mveshgoflon: . of the omdcfl;on charccterlstlcs of sodium was '_‘;undertaken. Sodiu h'near oxmfcmon_ _rcfe. " e tho 'oughi i SN. B. Pilling and R. E. Bedworth, J. Inst. Metals 29, - 529 (1923); see also, U, R. Evans, Metallic Corrosion, Passivity, and Protection, Longmans, Green and Co., 1948, New York, p 102, ! mary experlmenfs mdicateJ fhc’r? estlgo’red ‘ PERIOD ENDING JUNE 10, 1955 oxidation mechanisms of copper, aluminum, and other structural metals. On the basis of the results obtained with sodium, it was concluded that a fundamental error existed in the currently accepted oxidation concepts for metals that obey a linear oxidation rate law and that it was therefore desirable to investigate care- fully the oxidation of some metal that does exhibit a linear oxidation rate, Columbium was especially suitable for this study, Below approximately 400°C, it obeys a parabolic oxidation rate law, while above this temperature its oxidation rate is linear. At 400°C the initial stages of oxidation also appear to follow a parabolic rate, but after several hours " of oxidation the rate increases and becomes almost linear.® Thus a careful investigation of the structure and composition of the oxide films formed at or near 400°C should provide valuable information as to the conditions which lead to a linear oxidation rate. An added incentive for the study of the oxidation properties of columbium was that this metal pos- sesses very desirable high-temperature structural properties, Its use has been limited largely by its excessive oxidation at high temperatures, It was believed that further information concerning its oxidation mechanism would be helpful in overcoming this defect. The experimental procedures used and the results of the studies of sodium and columblum oxidation are presented below, 7 Sodium Oxidation. The oxidation rate of sodium was followed by measuring the change in pressure in a closed reaction chamber as the reaction pro- ceeded. A sensitive, differential manometer in which Octoil-S diffusion pump oil was used as the manometric fluid served as the pressure-sensing iy 'I“f.llls complefely the PlH,ng-*-f-:,,_q.dewce. “and Bedworth crl'rerlon'ifbr a’ metal‘thcf exhlbl'rs a\ Thls stpdy was initiated The apporcx‘rus used is shown in Fig. 5.17. The j"‘lsodlum reservoir consisted of a glass bulb closed ~at one end with a thln-walled break-off tip. The " ‘teéservoir was filled, under vacuum, through a side “arm, with sodium that had been purified by repeated “vacuum dnsflllahons at a pressure of 1078 to 1077 “mm Hg. Affer the Slde ‘arm was removed, the ‘reservoir was atfcched to a tube below fhe oxndc- tion bulb as shown in Fig. 5.17. A small furnace /a _hlfch__‘could be plcced around the entire ) the sodium reservoir and the 6H. Inouye, Scaling of Columbium in Air, ORNL-1565 (Sept. 1, 1953). : 105 ANP PROJECT PROGRESS REPORT BALL BEARING —_{ | BREAK-OFF TIP —{- 2-mm CAPILLARY -] » "TO VACUUM PUMP OXIDATION BULB " UNCLASSIFIED ORNL-LR-DWG 7416 ocToIL "s" DIFFERENTIAL MANOMETER L BALL BEARING 2T /BREAK~OFF TIP PURIFIED SODIUM | Fig. 5.17. Sodium Oxidation Testing Apparatus. . manometer protruding from suitable holes. The " entire apparatus could thus be baked-out under vacuum before any sodium was admitted to the ~ system. Care was taken to load the manometer by the vacuum distillation of QOctoil-S from a separate bulb into the manometer arms. This procedure 106 assured that the manometric fluid would not act as a source of appreciable gas, ' After a bake-out period of 16 to 20 hr the system was allowed to cool to room temperature, and the break-off tip above the sodium reservoir was crushed. The sodium was then distilled into the H " tube below the oxidation bulb, and the sodium reservoir was removed, Finally, the sodium was distilled into the oxidation bulb and allowed to condense on the walls, Both these final distil- lations were made under a pressure of approxi- mately 2 x 1077 mm Hg. As the last step before the admission of oxygen, the tube below the oxida- tion bulb was sealed off, as were the vacuum leads to the two reference bulbs, A 50-cc flask served as an oxygen reservoir. Prior to being attached to the oxidation apparatus, it was filled with carefully purified oxygen to a pressure such that when the gos was allowed to expand into the reference and oxidation bulbs the resultant pressure was 200 mm Hg. The purification of the oxygen was accomplished by passing it over hot copper oxide (to remove hydrogen) and Ascarite (to remove carbon dioxide) and finally drying it over magnesium perchlorate and in a liquid-nitrogen trap. - ' ' The tube leading from the oxygen reservoir to the break-off tip was made of 2-mm capillary tubing. By thus minimizing the volume above the reservoir, it was possible in the low-temperature runs to im- merse the reservoir in a cold bath to precool the oxygen to the desired reaction temperature. When the break-off tip above the oxygen reservoir was smashed, the oxygen passed through the T connection above the manometer and simultaneously filled the reference and oxidation bulbs to identical pressures. The infersection of the tubes of the T was then collapsed as quickly as possible with a hand torch, The three arms of the manometer were thus separated and any change in pressure in the oxndo’r!on bulb resultlng from the reccflon of sodlum" 25, 35, c:_nd 48°C, an apparatus identical to that _ neds "remenfs have been'made at five temperatures: —79, —20, 25, 35, and 48°C, At PERIOD ENDING JUNE 10, 1955 shown in Fig. 5.17 was used, except that only one reference bulb was found to be necessary. For reasons described below, the second reference bulb proved very helpful, however, for experiments at the two lower temperatures. For all experiments the oxidation and reference bulbs were immersed in a constant-temperature bath contained in a 4-liter Dewar flask. Mineral oil served as the bath liquid for the three higher temperatures. At -20°C a saturated sodium chloride—ice bath proved to be satisfactory, and at —=79°C a slurry consisting of powdered dry ice and a 50-50 vol % solution of carbon tetrachloride and chloroform was used to attain the desired temperature. The densities of the liquid phase and the dry ice in the latter bath were approximately equal, and thus a slurry was produced in which there was little tendency for the segregation of the dry-ice particles, At —79°C the sodium oxidized at a very slow rate. The maximum pressure change which oc- curred in the system as a result of the oxidation was 0.2 to 0.3 mm Hg. Therefore it was necessary to pay particularly careful attention to factors which could cause spurious pressure readings. One obvious source of error was a temperature dif- ference between the reference and oxidation bulbs. This difficulty was overcome by utilizing a relo- tively thin slurry in the cold bath and stitring it very vigorously, A less easily corrected source of trouble was that, of necessity, the reference and oxidation bulbs were maintained at dry-ice temper- ature, whereas the manometer and connecting tubes were at room temperature, A simple gas law calcu- lohonxshows thof, in such a system, any change in perature of elther the cold bath or the Ster will produce, in generol a difference in presswo “between the oxndohon bu[b cnd the refer- ence bulb A specnol cose ‘that is an excep’r:on to th:s rule occurs when the volumes of gas at the fwo tempero’rures ‘are |denhco| on both s:des of the o he mcgmtude of the pressure dif- ference is ‘dependenf both on fhe ratio of the volume Tof ¢ gas “held af 'rhe two temperature extremes “and on thé dlfference in femperoture. Thus, this effect of no grecn‘ |mportance for experlmenfs ‘close fo room temperafure, bu'r I'|' beccme sngmflcon'r at ' "t “was “impossible to maintain exactly equal ‘volumes on both sides of the manometer, and the attainment of precise temperature control in the 107 i ANP PROJECT PROGRESS REPORT cold bath at ~79°C was almost as difficult, The equilibrium temperature of any dry-ice bath is - determined by the sublimation point of carbon : 7_-di:o$ Na,0-Ni0 + H, the hydrogen pressure or the nickel concentration was determined as a function of time and tempera- ture. The hydrogen pressure was measured with 115 ANP PROJECT PROGRESS REPORT OUTSIDE 80'0 L0°0 900 0 0 z O X £0°0 00 1070 Fig. 5.24. Inconel Bucket Exposed to Sodium Hydroxide for 100 hr. Bucket temperature, 650°C; cold- finger temperature, 550°C. 75X. the reaction chamber connected to a manometer, and the nickel concentration was determined by analyzing the contents of a quartz-jacketed metal capsule after the desired exposure time. Since the two sets of data were necessarily determined on separate systems, there was the possibility of unknown variables being present, even though every effort was made to keep the variables identical. ‘ "_This.experiménml difficulty has recently been “¢ounferacted by constructing an apparatus that makes possible simultaneous determinations of - the hydrogen equilibrium pressures and the equi- librium nickel solubility. A charge of purified “sodium hydroxide is loaded into a hydrogen-fired mckel ‘capsule which is sealed under helium and then placed in a quartz tube that is evacuated and sealed. The jacketed capsule is heated to the test " temperature for the desired time period. If the exposure is sufficiently long, the hydrogen pres- sure developed should be in equilibrium with the melt, since it should not diffuse through the quartz. The pressure should be nearly equal inside and outside the nickel capsule, because hydrogen diffuses easily through this metal at the tempera- tures used. Upon completion of the high-tempera- ture exposure, the capsule is placed in a flanged metallic cylinder which, after evacuation, is con- nected to a mercury manometer. A small metal pin held near the quartz capsule by a metallic bellows extending through a lateral hole in the cylinder can then be tapped to break the quartz jacket. The pressure established in the system can be read on the manometer, and a simple calculation will give the pressure inside the quartz capsule before it was broken. ' Calibration tests with capsules containing hydrogen at a known pressure showed that the PERIOD ENDING JUNE 10, 1955 i 4 UNCLASSIFIED 1 Y-15171 OUTSIDE o o o o o o o O O O o O INCH o o) O w ~ o H w n - Fig. 5.25. Inconel Bucket Exposed to Sodium Hydroxide for 100 hr. Buckef temperature, 700°C; cold- finger temperature, 600°C. 75X. method allowed pressures to be determined to of failure of the nickel capsules. It was found within 1%. After the pressure measurement the that the nickel content of the hydroxide remained nickel capsule was opened and the hyd"rc’kid'e was nearly constant when the time of exposure was leached out. with the total alkalinity and the dis- varied considerably, which would indicate that solved nlckel bemg “determined on the same 'sample. equilibrium was reached after a short exposure Some of ‘the quartz- |ocketed nickel capsdles “time.” The hydrogen pressures determined by this Acom‘alnlng sodium hydromde were kep'r at 800°C ~method did not level off as expected, and the for various ‘lengfhs of time in order to ascertqm\ resultmg pressure was found to be greatly in the time necessary to reach equ:llbrlum. Runs for excess of that to be expected from consideration long perlods of time were unsuccessful because of the rec:c'rlon postulated. TR 0 " 117 ANP PROJECT PROGRESS REPORT UNCLASSIFIED Y-15168 ¢ . o Jo o Je Jo o [ Jo Je Jo o Jo o o o O o QINCH|©@ o o o o o ‘ : ; B ~ o o N » ™ o o P B N Fig. 5.26. Inconel Bucket Exposed to Sodium Hydroxide for 100 hr. Bucket temperature, 800°C; cold- finger temperature, 700°C. 250X. &Y s Y % ) Coo Dec 10 1954 ORNL 1816, o 100 PERIOD ENDING JUNE 10, 1955 6. METALLURGY AND CERAMICS W, D. Manly J. M. Warde Mefcllurgy D:vus:on Addmoncl fcbr:caflon studles of Hasfelioy B have mcreased the evidence that the poor hlgh- 'tempercture fcbrlcabtllty is related to the impurity content from which the strength may be derived. Several nickel-molybdenum binary and ternary al- loys were studied in oxidation tests and in room- and elevated-temperature tensile tests. The ter- ~nary alloys all included 20% molybdenum plus nickel and a third heavy element. The results of -additional stress-rupture and tensile property studies of the nickel-molybdenum alloys are presented. For comparison with previously obtained data on the static oxidation of several brazing alloys, cyclic tests were run; the static and cyclic data were similar for tests at 1500°F. Cyclic tests at ]700°F are under way. ' Fabrlcqhon was comple'red of a 20-tube Inconel fuei-fo-NaK heat _exchanger, the fuel-to-NaK inter- mediate heat exchanger No. 2, two 500-kw NaK-tfo- air radiators, and a liquid mefala'ro-olr radiator de- signed by the Cornell Aeronautical Laboratory, Special extrusions were made of three billets of vanadium and four high-purity molybdenum billets containing 0.7% titanium, and studies of flow pat- terns of duplex and three-ply materials during im- pact extrusion were continved. The fabrication of clad columbium was investigated, and diffusion barrier studies were made. Additional information was obmmed on the properties of B 6C-Cu mixtures and a magnesnum -lithium alloy fhaf are belng con- 5|dered os‘ posstble shleldm' "'materlals. 24, Inouye, J. H. Coobs, and M. R. D'Amore, ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL.-1864, p 97. e DEVELOPMENT OF NICKEL-MOLY BD ENUM ALLOYS J. H. Coobs H. Inouye Metallurgy Division M. R. D’ Amore Pratt & Whitney Aircraft Fabrication Studies Hastelloy B. Two impact extrusions of Hastel- loy B rod were made at 2000°F with good recovery of sound material. The billets were prepared from a vacuumemelted ingot of commercial Hastelloy B and were extruded in the as-cast condition. During subsequent hot reduction of the extruded rod to sheet, the material cracked severely, and no usable. material was obtained. Two extruded tube blanks that had been shipped to the Superior Tube Co. for reduction to small- diameter tubing were processed; one blank was made from wrought material and the other from cast material. The tube blank made from a vacuum- melted cast billet fractured during the first reduc- tion operation, but the blank made from a wrought billet was successfully processed from 1.5-in.-OD, 0.250.in.-wall to 0.187.in,-0D, 0.017-in.-wall seam- fess tubing. It was found that severe reduction schedules were permissible and that intermediate stress-relieving heat treatments were unnecessary. Previously reported experiments! showed that higher extrusion ratios were attainable if the Hastelloy B billet was canned in Inconel. There- fore wrought Hastelloy B billets with and without Inconel cans have been prepared for furfher ’rube : "“extrusmn experiments. © Attempts to roll extruded Hastelloy B rod at --2000°F were unsuccessful; the material cracked ”‘:‘S‘évérely after several 5% reduction passes. Since 0.03% cerium additions had previously? been found to improve the high-temperature fabricability of Hastelloy B, cdditional'me!fs were prepared with ‘various quanfmes of cerlum aclded in the form of ni 1 ]% Ce—89% Al master o”oy A small vacuum melt confaining "0.3% cerium “cracked ‘during hot reduction, and in subsequent experiments it was shown that a maximum of 0.1% cerium was useful in improving the high-temperature fabricability. 119 ANP PROJECT PROGRESS REPORT An alloy wi'rh. the nominal Hastelloy B compo- sition, 4% Fe-28% Mo—-68% Ni, but without the tramp elements vanadium, silicon, manganese, co- ~ balt, chromium, tungsten, and aluminum, was pre- pared for evaluating the effects of impurities on the fabrication properties of Hastelloy B. Three- pound vacuum melted slab ingots 0.500 in. thick were prepared and rolled at 2100°F to 0.150-in.- thick sheet. This material showed no tendency to crack during hot rolling under moderate reduction ~schedules. The microstructure of the alloy in the as-cast condition showed random distribution of a " second phase, which appeared to be an oxide. The source of this phase might be traced to the starting material or the melting practice. It is becoming increasingly evident that the poor high-temperature fabricability of Hastelloy B is related to the im- purity content, However, since the superior strength of commercial Hastelloy B in comparison with that " of Inconel might be derived from these impurities, mechanical property tests of the pure 4% Fe-28% Mo—-68% Ni alloy are under way. The preliminary room-temperature tensile data, presented in Table 6.1, are encouraging in that they show the tensile strength and ductility to be comparable to those of the commercial alloy under these test conditions. Additional mechanical property tests are under way. TABLE 6.1. ROOM-TEMPERATURE TENSILE STRENGTH OF THE 4% Fe-28% Mo—68% Ni ALLOY Tensile , Elongation Condition Strength _ ; (%) - (psi) Annealed . 130,000 40 Annealed, aged 500 hr . 137,000 2.5 ~ at 1300°F Nickel-Molybdenum Binary Alloys. The previ- ously reported® work on nickel-molybdenum binary alloys was extended by a study of an 85% Ni-15% Mo alloy. The dlloys studied previously were 80% Ni-20% Mo, 76% Ni-24% Mo, and 68% Ni- 32% Mo. An oxidation test of the 85% Ni—15% Mo alloy at 1500°F in static air indicated a lower rate of oxidation than that observed for the 80% Ni~ 20% Mo alloy, but o higher rate than the rates for ’rhe 76% N|~24% Mo and 68% Ni-32% Mo alloys. ,3H. Inouye and J. H. Coobs, ANP Quar. Prog. Rep. Dec IO 1954, ORNL-18164, p 103. 120 Therefore additional oxidation tests are planned to supplement the present data. The presently avail- able data appear to indicate that the rate of oxida- tion of these binary nickel-molybdenum alloys is not a function of the molybdenum content, Two thermal-convection loops for circulating fluoride mixture in the 85% Ni-15% Mo tubing are being fabricated. ' Both room- and elevated-temperature (1100 to 1650°F) tensile tests of this alloy have been com- pleted. Erratic results were obtained in the room- temperature tests of specimens aged 284 hr at ele- vated temperatures (1300 to 1650°F), The tensile strength was low at all test temperatures, and low elongations were observed at test témpero’rures above 1300°F. The results of these tests are given in Table 6.2, Nickel-Molybdenum Ternary Alloys, A screening program was initiated in orderto observe theeffects of additions of alloying elements to nickel-molybde- num alloys. The alloy compositions investigated have all included 20% molybdenum plus nickel and a third element. The ternary systems that have been evaluated are: 2 to 10% Cb-20% Mo~bal Nij 3 to 10% Cr~20% Mo-bal Ni 2% V-20% Mo-78% Ni 1% Zrw=20% Mo-~-79% Ni 2% Al-20% Mo —78% Ni 1% Ti-20% Mo—79% Ni The physical property data obtained for alloys con- taining 2 to 10% columbium were reported previ- ously.y The aluminum-molybdenum-nickel alloys were investigated because of the exceptional creep properties reported in the literature for this alloy system, Hot-forgeability studies were conducted on 100-g arc melts containing from 2 to 10% alumi- num, but only the alloys containing less than 5% ~aluminum were found to be forgeable at 2100°F. Further studies will be conducted on this alloy system. ' S Vanadium, zirconium, and titanium additions were investigated in an effort to improve the elevated- temperature ductility of the 20% Mo-~80% Ni alloy. No difficulties were experienced in fhe hot rolling of these alloys. Preliminary room-temperature tensile strength data for the alloys containing aluminum, vanadium, zirconium, and titanium in both the annealed and aged conditions are presented in Table 6.3. All specimens aged at 1650°F for long periods of time 41bid., p 100. »} © @) PERIOD ENDING JUNE 10, 1955 TABLE 6.2. RESULTS OF TENSILE TESTS OF AN 85% Ni—15% Mo ALLOY Elongation in 2-in. Yield Point, Tensile iti Test T Condlflon of | est e;nperc:ture 0.2% Offsot Strength Gage Lengths Test Specimen* ("F) ] . . (psi) {psi) (%) " Annealed 1 hr at 2100°F Room 32,100 60,400 15 Annealed, aged 284 hr ot 1300°F Room 32,700 70,200 17.5 Annealed, aged 284 hr at 1500°F Room 32,000 | 99,700 50 Annealed, aged 284 hr at 1650°F Room 31,500 68,400 21 Annealed 1 hr at 2100°F 1100 18,7007 39,900 16.5 1300 17,850 30,300 7.5 1500 17,000 . 26,000 5.0 1600 16,600 23,200 6.3 1650 16,000 20,700 9.0 *Sheet, 0.065 in. thick. TABLE 6.3, ROOM-TEMPERATURE TENSILE STRENGTHS OF SEVERAL NICKEL-MOLYBDENUM BASE TERNARY ALLOYS Tensile Strength Elongation Alloy Composition Condition of o (wt %) Test Specimen {psi) (%) 2 Al—20 Mo—78 Ni Annealed 150,500 32 Annealed, aged 500 hr at 1500°F 118,500 59 . Annealed, aged 284 hr at 1650°F 70,500 12 2 V20 Mo-78 Ni Annealed 117,000 32 : o Annealed, aged 284 hr at 1300°F 127,000 30 1 Zr—20 Mo~79 Ni Annealed | 101,000 33 - Annealed, aged 284 hr at 1300°F 115,000 58 1 Tie20 Mo..79 Nj Annealed 115,000 60 L e _Aged 284 hr at 1650° 73,700 s : show lower s'rreng’rhs cnd ducflllhes ‘than fhose _ unaged or aged at fower temperctures. ‘The speci- mens ‘aged at 1650°F were exposed to a hydrogen o '."-atmosphere, ‘while the ofher specnmens were treated " Under a hehum afmosphere or in evacuatec -~ _‘,.-'f'cqpsules. The possnb|||ty of hydrogen embriih‘le-::" . meént of 'rhe SpeCImens aged m‘ ]650°F is bemg Y nves'r:gated . o The oHoys confomlng 3% 10% chromlum have | S been evolu 3y - tensile fes'rs. “made for corrosion 'restmg ‘has been ilml'red how- “ever, because of shattering during extrusion.” Al- ed i §|dqt|on, ‘s‘tr' ss-rupture, “and The quantity “of seamless fubmg though sound rod and tube blanks have been made at an extrusion temperature of 2200°F, consistent ) femperafure range ‘was unsuc _results could nof be obfc:med A study of these olloys hos led to the supposmon thc’r the shrink cavities within the mgot are the cause of the frac- ' turlng The billets are currently being mspecfed by x-ray and gqmma-radlahon techmques, prior to exfrusmn, in an ‘effort to correlate the soundness of the starhng maferucrl with the extrusmn result. " The hot rollmg of extruded rods of chromlum-_: 'molybdenum-nlckel alloys in the ]95Q to 2200°F_ ul because of ‘severe crackmg Subsequem‘;sfud:es showed that “the hot shor‘ ess was probably caused by oxygen‘ ‘contamination of the electrolytic chromium used, It was found that additions of 0.1% cerium for each 3% chromium in the melt would render these alloys 121 i i S ANP PROJECT PROGRESS REPORT hot forgeable. Further experiments have been con- ducted with 100-g arc melts containing up to 2% titanium and 2% aluminum as deoxidants. The aluminum additions were found to be ineffective, “ but the titanium additions served to reduce the hot - shortness, It was determined that approximately " 2% titanium would render a 5% Cr—~20% Mo—-75% Ni - alloy completely hot forgeable. The extruded rods of the hot-short alloys could readily be rolled to sheet at room temperature. Work-hardening data for a 3% Cr-20% Mo-77% Ni alloy are presented in Table64 “The alloys containing between 3 and - 10% chromium work-hardened at the same rate. TABLE 6.4. WORK-HARDENING OF A © 3% Cr-20% Mo-77% Ni ALLOY " Reduction (%) ' Hardness (VPN) o 168 104 260 211 ' 312 30.9 337 45.5 409 524 441 63.1 480 Room-temperature tensile strength data for chro- miumemolybdenum-nickel specimens subjected to various heat freatments are listed in Table 6.5. The data reported are the average of the results of two to three tests of each alloy. Included in the tests were alloys that contained additions of 0.5% columbium or up to 0.25% cerium. The minor ad- ditions of columbium appeared to incfe_osé slightly the tensile strength and ductility of these alloys, and an average decrease of 5% in elongation and a slight increase in tensile strength were noted for the alloys with small amounts of cerium added. The elevated-temperature tensile strengths of chromium-molybdenum-nickel alloy specimens after various aging treatments are given in Table 6.6, All the specimens were tested at the aging temper- atures listed. The results are given for single tests, The elevated-temperature ductilities of these alloys are low compared with the ductilities at room temperature. The minor additions of cerium apparently aided in obtaining higher elevated- temperature strengths in these alloys, but more tests are needed to verify this observation, No relationship between the physical properties and the chromium content is readily apparent in either room- or elevated-temperature tests. In the temperature range above 1300°F the alloys containing up to 10% chromium appear to be single- TABLE 6.5. ROOM-TEMPERATURE TENSILE STRENGTH DATA FOR CHROMIUM-MOL YBDENUM-NICKEL ALLOY SHEET SPECIMENS Aging Time: minimum, 284 hr; maximum, 500 hr Alloy Composition Condition of Tensile Strength Elongation (wt %) Test Specimen (psi) (%) 3»Cr—20 M9—77 Ni Annealed 107,000 61 Aged at 1300°F 109,500 60 Aged at 1500°F 111, 100 64 Aged at 1650°F 67,500 19 5 Cr=20 Mo~75 Ni Annealed 120, 000 | 58 Aged at 1300°F 119,000 59 Aged at 1500°F 115,000 | 62 Aged at 1650°F 72,500 14 7 Cr-20 Mo=73 Ni Annealed 116,000 | 63 | | Aged at 1300°F 116,000 | 63 Aged at 1500°F 114,000 59 o o | ~ Aged at 1650°F 69,000 14 10 Cra20 Mo—70Ni Annealed 110,500 n : C ' Aged at 1300°F 111,600 63 122 »t N PERIOD ENDING JUNE 10, 1955 TABLE 6.6, ELEVATED-TEMPERATURE TENSILE STRENGTH DATA FOR CHROMIUM-MOL_YBDENUM-N!CKEL ALLOY SHEET SPECIMENS Tensile Strength Elongation Alloy Composition A'ging Treatment (wt %) ~(hr) CF) (psi) (%) 3Cro20 Mo-77 Ni 500 1300 33,200 6.3 o 500 1500 27,300 5.0 _ N 284 1650 26,100 19.0 3 Cr=0.5 Cb—20 Mo—76.5 Ni 362 1300 38,100 6.3 | | 362 1500 30,000 - 5.0 5 Cr-0.5 Cb~20 Mo—74.5 Ni 500 1300 39,500 7.5 500 1500 29,400 5.0 5 Cr-0.25 Ce~20 Mo—74.75 Ni 362 1300 59,700 16.5 ' _ 362 1500 43,100 11.0* 7 Cr-20 Mo~73 Ni 362 1300 40,800 9.0 ' 362 1500 29,800 3.5% 7 Cr—0.5 Cb-20 Mo~72.5 Ni 500 1300 48,800 12.5 500 1500 38,500 7.5 | - 284 - 1650 29,800 3.8 7 Cr=0.1 Ce—~20 Mo—72.9 Ni 362 1300 54,500 s , | 362 1500 40,500 8.8* 10 Cr—0.5 Cb—20 Mo—69.5 Ni 500 1300 38,300 8.8 : 500 1500 43,600 9.0 1650 29,900 6.3 284 *Specimen fractured outside 2-in. gage length. phase. Additional studies of the microstructures will be conducted on alloys made with high-purity chromium as the starting material, Further evi- dence of the equdnbnum structure is shown in aged VSpec:mens. No sngmflcont changes in tensile ' sfrength or duchhfy are obServed in specumens'“' -~ aged at 1300 and 1500°F The Iow_ sfrength and ~ ductilities of spec:mens aged at 1650° " '-been cq'used by fhe hydrogen atm _sphere used ‘ of ‘stress-ru pture tests N""creasmga ’rhe s'-*rength ‘and’ the duc’rlhty 'rhroUgh ' cerium additions appears to be promising, since the alloys with cerium additions have shown better may have_ (presenfed_ _ h n) have shown low sfrengths and‘__ physicd[ properties during sfreés#rupture testing. It has become apparent that improved deoxidation practices are necessary in melting these alloys, and efforts are being dlrected toward the production of sounder mgo’rs. 0xnduhon Studies Oxidation studies were continued on nickel- "“molybdenum alloys containing from 3 to 10% chro- w‘mium, The results reporfed prevnouslys for these ““alloys were based on oxidation tests of small arc-melted specimens that had not been chemlccily:' inalyzed, The alloys hove therefore been retested by using specxmens from chemlcally anclyzed “heats of material, qnd ‘the results are presented in fTable 6.7. In comporison “with the previous re- ~sults, the oxidation rate for the 3% chromlum ‘com- “pesition was’ slightly Iower, ‘and mcreoses ‘were “noted for the alloys containing” 5, 7, and 10% SH. Inouye and M. R. D*Amore, ANP Quar. Prog. Rep, Mar. 10, 1955, ORNL-1864, p 104, 123 L e T ANP PROJECT PROGRESS REPORT TABLE 6.7. OXIDATION OF CHROMIUM-MOLYBDENUM-NICKEL ALLOY DURING ‘ 168 he AT 1500°F IN STATIC AIR Alloy Composition Weight Gain 2 Remarks (wt %) (mg/cm*) 3 Cr=20 Mo—77 Ni 5.17 Oxide spalled completely during cooling to room temperature ' 5 Cr.20 Mo .75 Ni 5.30 Oxide spalled completely during cooling 7 _ _ to room temperature ' 7 Cr=20 Mo~73 Ni 3.14 Oxide spalled badly during cooling to s room temperature ' ’ 10 Cr=20 Mo-70 Ni 0.45 No spalling occurred chromium. It appears that about 10% chromium is necessary to form a nonspalling protective oxide ‘on the ternary alloy, However, it has been ob- served '.fh'at the oxidation rate under static condi- tions can be reduced 50% by the addition of 3% chromium. STRESS-RUPTURE STUDIES OF NICKEL- MOLYBDENUM ALLOYS D. A. Douglas J. H. DeVan J. W, Woods Metallurgy Division Hastelloy B A series of creep-rupture tests on Hastelloy B in the solution-annealed condition has been com- pleted in an argon atmosphere at 1500 and 1650°F. The results are summarized in the design curves presented in Figs., 6.1 and 6.2. A similar series of tests in fused salts is nearing completion that provides an interesting comparison with the argon results, ‘ _ } Stress-rupture plots are presented in Fig. 6.3 for Hostelloy B in the two environments at 1300, 1500, and 1650°F, and it may be noted that at each temperature the values obtained in fused salts are octUaIl_y superior to those obtained in argon. Al- though no adverse effect on physical properties was expected, in view of the absence of corrosive attack of the fused sclts on Hastelloy B under static conditions, the reason for the increase in ruptire strength is not entirely clear at present. The test chamber must be flushed with a cleaning charge before the fluoride mixture to be used for the test is admitted, and therefore the specimens tested in fused salts were at temperature approxi- 124 mately 5 hr longer before stress was applied than were the specimens tested in argon. Consequently, the slightly longer aging of the specimens tested in the fused salt may account for the improvement in strength, There are objections to this theory, however. A specimen aged 70 hr prior to stressing and testing in argon showed inferior properties relafive to those of a comparable unaged specimen tested in argon. Also, there is some doubt as to whether aging can account for the improvement at 1650°F, since the phase diagram, if minor ele- ments are neglected, indicates a one-phase region at this temperature. (Photomicrographs, however, have shown evidence that aging at this temperature under stress does occur.) Other tests are being carried out to establish the exact causes for the apparent anomaly in strengths in the two environ- ments. S Tests in hydrogen and air also have been made in order fo determine the effects of these environ- ments on the creep-rupture pr0perfie$ of Hastelloy B at 1500°F, The creep curves obtained at 12,000 psi in hydrogen, argon, air, and fused salt are shown in Fig. 6.4, In comparison with the effect of argon, the effect of hydrogen is apparently negligible, and the effect of air follows closely the pattern observed for Inconel and *‘A’* nickel in air, insofar as reduced creep rate and longer rupture life are concerned. However, the final elongations of Hastelloy B in air are equivalent to or lower than those in argon, whereas the elon- gation of other nickel-base alloys are markedly greater in air than in argon. A fest program simi- lar to that carried out on solution-annealed "Hds?erifi loy B has been initiated on solution-annealed specimens daged 100 hr at 1300°F to determine 0 e Hastelloy*B from the stondpom’r 6\"" 20,000 18,000 16,000 14,000 12,000 STRESS {psi) 10,000 8000 6000 PERIOD ENDING JUNE 10, 1955 UNCLASSIFIED ORNL-LR- DWG 7558 RUPTURE 100 1000 _ 10,000 TIME (hr) Fig. 6.1. Sfress-Rup?ure Curves for 0, 060-m.-Thzck Sheet Specnmens of Hastelloy B Solution-Annealed at 2100°F and Tested in Argon at 1500°F. whether aging treatment results in serious em- brittlement at service temperatures, Modified Nickel-Molybdenum Alloys Creep-rupture tests of modified nicke l-molybdenum alloys at 1500°F in argon are under way. The results of tests on these alloys, prepared by - vacuum meltlng, have been somewhat dlsappomhngm ‘ - with respect to duchllty attained before rupture, ; 20% Mo but, at 8000 psi, it had a final elongation of 28% compared with 16% for Hastelloy B and 7% for a similar alloy to which no deoxidant (cerium) was }{H is currenfly beheved that fhe low duchiity maymi' “be atfrtbufed to meifmg prachce ‘rather than to “the anure of thé aHoy sysfems. ‘A heat to whtch”_" a sma“ amq_un. of deoxtdanf was cdded showed a ;:-is;gnlficqn'r Improvemen‘r in_properties compared_:d _strength data for this material after heating at ~with those of other smgle-phase alloys tested to _date, The alloy had the composition 74 75%”N1-“ h Cr~0 25% Ce and was inf rior - from _these heat treatments have been studied and added, The results of recent creep-rupture tests of several modified nickel-molybdenum alloys are presented in Table 6.8, TENSILE PROPERTIES OF HASTELLOY B 'P. Patriarca R. E. Clausing Meta”urgy Division It has been shown® that the physu:dl properties of wrought Hastelloy B alloy are directly influenced by the precipitation which occurs at temperatures ..within the intended operational range of ANP reactors and heat exchangers. An extensive pro- ..gram is now under way to obtain short-time tensile _tempercfures from 1100 to 1600°F for times from 100 to 1000 hr. Typical microstructures obtained _photographed, and a correlation is being made P, Patriarca et al.,, ANP Quar, Prog Rep, Mar, 10, 1955, ORNL.-1864, p 116. 125 He work . ANP PROJECT PROGRESS REPORT UNCLASSIFIED - ~ ORNL—-LR-DWG 755¢ 14,000 ‘ 13,000 12,000 ~ 44,000 RUPTURE 40,000 < : . N R N 2% A \ 9000 - " 8600 < = 3 N 3z \ \ \\\ \\ \k\ ' 0.5% ELONGATION N W 7000 \\ o~ ™, N NN i N ™~ ™ NN & N N NG TR @ . N ™~ \\ ’ GOOO \ \\ \\ N ™ \ N o \ \\ \\\ 5000 N SN 4000 1 10 100 1000 10,000 TIME (hr) "Fig. 6.2. Stress-Rupture Curves for 0.060-in.-Thick Sheet Specimens of Hasfel.by B Solution-Annealed at 2100°F and Tested in Argon at 1650°F. TABLE 6.8. CREEP-RUPTURE TESTS OF MODIFIED NICKEL-MOLYBDENUM ALLOYS Alloy Composition Stress Time to Rupture Elongation (wt %) (psi) (hr) (%) 77 Ni=20 Mo~3 Cr 8000 90 7 5000 675 9 75 Ni_20 Mo~5 Cr _ 8000 80 7 74.75 Ni~20 Mo-5 Cr~0.25 Ce 8000 444 28 73 Ni=20 Mo~7 Cr 8000 380 | 9 befween physical properfles and the correspondmg 'mlcros’rruc’rures. As a result of this investigation, ‘more comprehenswe information on the high-tem- perature properties of this material is being ob- . tained, The relative merits of various preaging " heat trectments of Hastelloy B are also being )sfudled and it is hoped that, as a result of this a procedure can be developed which will stabilize the microstructure sufficiently to reduce the sensitivity to high-temperature aging, A summary of the short-time tensile data that have been obtained to date is presented in Flgs. 6.5, 6.6, and 6.7, and in Table 6.9. It may be noted in Fig. 6.5, which shows the results of the test in which the variable of aging time was in- vestigated, that the tensile-strength and yield-point LT . . e ol o kil L s L STRESS (psi) PERIOD ENDING JUNE 10, 1955 -“EeneT ORNL—LR—DWG 7560 60 50 40 30 n o 10 | FLUCRIDE FUEL: NaF—ZrF4—UF4 (50-46—4 mole %o 10 100 1000 10,000 TIME (hr) Fig. 6.3. Comparison of Hastelloy B Stress-Rupture Data in Argon and in the Fuel Mixture NoF.ZrF,- UF, (50»46-4 mole %) at ]300 1500, and '|650°F TABLE 6 9. ROOM-TEMPERATURE TENSILE PROPERTIES OF HASTELLOY B IN THE :SOLUTION ANNEALED AND SPHEROIDIZED CONDITIONS ‘Yi'e'rld "S_t'i'é-n'gfh ‘Elongation - curves for 500— - max1mum Tat ag Y ofSpecimen (p's'i)"" - (%) - Spheridized 175000 10 185 000 100, ooo 0115000 151020 ‘Solution-anneal 125,000 10 130,000 50, 000 to 60,000 55 to 65 - tempera?ure ’rensule properhes obtamed by a specml ~ spheroidization heat treatment® are compared in "Figs. 6.6 and 6.7 with those of solution-annealed ‘and 'IOOObeZ agmg ’rlmes. reoch a e'l)-/' 1300°F, while the -hes o _minimim at this samer_’ _j‘dUCfl[Ify Curve ré femperature. “The results of meféllographlc studles,:" ‘discussed prewously, “indicafed that this was | caused by the ‘exfensive precipifation of a second phase — beta or a combination of two phases, beta and gamma — throughout the matrix. The high- “material in the unaged and in fhe uged conditions, "'i"'Although ‘the sphermd:zahon ’rrecn‘ment was in- " tended to produce a stable microstructure between 1300 and 1600°F, it may be noted that the ductility is lowered considerably in the 1100 to 1300°F 127 ANP PROJECT PROGRESS REPORT O CRt ORNL—LR—DWG 7561 TESTED IN HYDROGE TESTED [N AIR TESTED IN ARGON ELONGATION (%) TESTED IN thF--ZrI-",_‘—UF4 {50--46--4 mole %) o 200 420 600 800 {000 1200 1400 1600 1800 2000 2200 2400 2800 TIME (hr) Fig. 6.4. Creep Curves for Hastelloy B Sheet That Was Solution Annealed ot 2100°F and Tested at 1500°F and 12,000 psi in Various Environments. UNCLASSIFIED temperature region. Even though very high tensile ORNL-LR—DWG 7562 and yield strengths were obtained with this treat- 1000 HR ment, the low high-temperature ductility would "O-0— 500 HR probably prevent useful application of this material % "o 100 MR above 1000°F. Investigations of other promising < pretreatments are now in progress, as well as investigations of the composition variables which 1o -~ 400 W0 O o] o TENSILE STRENGTH &/‘%\ G 2, M o 5 S . €70 SILE STRENGTH & have been shown to be of significance. o 60 5 — N g o P PO & DEVELOPMENT OF BRAZING ALLOYS < ‘{\E\' DO//!/ . , E 0 YIELD PO P. Patriarca . G.. M. Slaughter = *—e Metallurgy Division © 30 g & e % R. L. Heestand Y 20 * 2 Pratt & Whitney Aircraft 2 DUCTILITY | DUCTILITY S = S L DUCTILITY High-Temperature Oxidation Tests 1100 1200 1300 1400 1500 1600 Tests for evaluating the static oxidation re- TEMPERATURE (°F) sistance of several brazing alloys were conducted U _ previously at 1500 and 1700°F, and cyclic tests Fig. 6.5. High-Tempercture Tensile Properties have now been initiated, In the 500-hr cyclic of Hastelloy B That Wes Solution Annecled at tests the samples are subjected to 190 air cools 2100°F Prior to Aging at the Testing Temperature from 1500°F, The results gbtained thus far are for Various Times, compared with the results of the static tests in 128 a) »n UNCLASSIFIED ORNL-LR~DWG 7563 140 130 120 110 100 20 80 70 = SPHERQIDIZED - SOLUTION ANNEALED 2 HR AT 2100°F 60 50 40 40 30 30 TENSILE STRENGTH AND YIELD POINT ( psix{0™ 3} 20 20 ELONGATION (%) 10 DUCTILITY 10 0 O 1100 1200 1300 1400 1500 1800 ‘ TEMPERATURE (°F) Fig. 6.6. High-Tempefufdre Tensile Properties of Hastelloy B in the Solution Annealed and Spheroidized Conditions. Table 6.10, alloys tested have good resistance to oxidation under both static and cyclic conditions. Tests are now under way for evaluating the attack resulting from cycling from 1700°F. An apparatus is also being prepared for testing the oxidation resistance of brazed joints in moving It may be seen that many of the moist air. Physwal Properfy Tests _4 An mveshgcn‘lon “has been conducfed of accumu-" S "_ICmVe effecfs on the physical properhes of InconeI/ tubmg a5 a@” result of fcbrlcatlon of 'rube-to-fmj_ :-_,;':-.'IJOII‘H‘S by hlgh temperature brazmg “with Coast . Metals cxlloy No. 52 Brazed fube ‘|’O-f!n specnmens' ~with_at least 1 in, “of ‘stacked fins were machine = o ""ground To 'rhe orlgmcl fubmg dimensions. Threaded | : "henv brozed to fhese tubes' to Room- cmd hlgh ’remperofure fens;le tests were “then conducted on The as-brazed specimens and on specimens that were subsequem‘ly heated in a vacuum for periods up to 500 hr at 1500°F prior to testing. The results indicated that no signifi- - heat treatments. N g2 PERIOD ENDING JUNE 10, 1955 UNCLASSIFIED ORNL~LR—-DWG 7564 140 430 120 ¥y 140 o Py r " Q > ‘@ 2100 '_ = o 90 o, O 3 80 W > g 70 ELECTRODES . MIXING CHAMBER | | i \ I I * TEST SECTION FREEZE VALVE MIXING CHAMBER SUMP FILL TANK Fig. 7.3; ‘{.S.c'l;é'r.na-fii:'.'Iji'agrom of Small-Scale Pumpfrigl Sysfemfot Heat Transfer Studies with Fused Sst. e o ' o s E HYDR'O'DYNAMICS 0% J,Q Brgdfufe"”f“‘ " F. E.Lynch s rer curves 1c o L.D. Palmer o r-;-f_r_:exchcnger are shown -il;.l '."fFlgs 7-5 "dhd 7.6-_,_‘_,;_ N Reoctor Expenmem‘ol Englneerlng D|v15|on _/Fvlgure 75 is based on 18 data points that fall - G. L. Muller o Prafl & Whn‘ney AII'CI'G‘H' 7 Furfher flow studles ‘_were conducted w_n‘h ’rhe quar’rer scole model_;'of thé'_ ]8~|n. ART core ::rfhrough fl'me core w:fhouf ro’rqtlon have been tdken, “and " quantitafive veloc:ty “profiles have been »«‘i-’ \’,’\4-’ L “mined by u’rlhzmg the tube bundle as an electrical- resistance heater. Results of the sfudy should obtained. further substantiate the curve shown in Fig. 7.5. A vaned section was added at the inlet to the 151 TR T I I N R T T R OEST mEe Ro pot ALTE ANP PROJECT PROGRESS REPORT TS wwn»‘unnw e, = z S STorisiinal d‘tlo.fctcoooo;,:ccfsv.'fl e A iorrereYse TSt T Tas ol T et TR i ST T ot 3 : <3 ey : % b - T ) o Tae 2 % D B st 2 ) & eIy amIstTaes e tOowooHNN“oc a3 s e e T e e 5 > v e red. oz o e 2335 SITeTAEL. N sl AveELriT 152 iment. -to-NaK Heat Exchanger Exper ART Fuel 4 7 Fig. e M w i ey e R S ' | PERIOD ENDING JUNE 10, 1955 core model to give the fluid a rotational component (approximately o 45-deg helical angle ot the o ’ ORNL~LR-DWG 7322 7 40 ' o b entrance). The velocity profiles were then ob- 2 A served qualitatively with the flow-visualization " : Nu/PrO9 0 025 Reo,;_w/ Technique. Separation was observed next. to the ' L7 island wall rather than the shell wall, which was : 20 /,/ W% the case with sfrdight-throug!h flow. The axial 7 / component profiles are shown in Fig. 7.7. 3. //' In addition, a series of 16-mesh screens was ; //’ ' / added at the inlet, both with and without the 10 / ' rotation-producing vanes. Little effect attributable // to the screens was noted in the straight-through P case, except that the separation region appeared /’ to be much less turbulent; indeed, the negative /_ - profiles were easy to see. They appeared to be 5 / approximately parabolic, and thus they suggested . _ 000 2000 5000 10,000 LEUREY REYNOLDS MODULUS, A, ORNL-LR-DWG 7324 an. 7.5. Prehmmary Heat Transfer Churacier- istics of the Fuel Side of the ART Fuel-to-NaK : m\U] T Heat Exchanger as Determined by Water-to-Water ' Tests. | Rl . . : = ORNL—LR=DWG 7323 . 02 - >~ / / W ». Qi S /—/ - “ . REVERSE FLOW o N | W , & - . 2 N 2 ‘ _ g > g E - S~ 2 005 — -~ o ) : . SMOOTH PIPEL_| : U ! " Fig. 7.6. of the Fuel ‘Slde of the ART Fuel-fo-NaK Heat Exchanger. Flow “Through o Model of the ART Core af a Reynolds Number of 3000 with a Rotational Ve- locity Component at the Inlet. 153 Fig.'7/7.‘ Qualltaflve Axial Yelocity Profiles of’ - sl Qo na? AL ANP PROJECT PROGRESS REPORT a laminar type of flow, No effect of the screens could be observed with rotational flow. A two-dimensional diffuser has been designed which permits a variation of diffuser geometries and cross-sectional area ratios. The entire diffuser section is fo be made of Plexiglas so that the flow can be studied by the flow-visualization system. The assembly is now being fabricated. A preliminary feasibility study is at present being made on the control of the boundary layer in the core for the purpose of preventing separation of the boundary layer. A boundary layer suction “technique is being considered. A plastic 10/44-scale model of the 21-in. core has been designed and is now being fabricated. The odd scale was chosen to make the model fit the available testing facility. REACTOR CORE HEAT TRANSFER H. F. Poppendiek N. D. Greene L. D. Palmer Reactor Experimental Engineering Division The hydrodynamic studies being made for the ART core indicate that when separation regions exist they are often characterized by repetitive, short periods of flow stagnation. A study was made of the temperature and tensile stress fluctu- ations in the Inconel core wall that result from these periods of flow stagnation. The following idealized system depicting the ART core wall during a short period of flow stagnation was con- sidered. It was postulated that a layer of fuel contiguous to the Inconel wall suddenly stagnated for a period of 0.1 sec under the volume heat flux condition of 5 kw_/cms. This flux is representative of the conditions in the ART because of the high flux peak at the wall. After a 0.1l-sec time interval the fuel and wall temperature would rise about 260°F if no heat transfer were present; transient heat conduction into the Inconel wall, however, would reduce this temperature rise. A numerical composite-slab heat transfer analysis was made for this problem, and the results indicate that the " Inconel-fuel interface temperature would rise about 60°F in 0.1 sec. This temperature rise would, of course, be much smaller if the thermal conductivity of Inconel were not so poor. The corresponding eicshc thermal stress for this temperature fluctu- -'? ation was found to be about —15,000 psi. These results suggest that the fluid flow in the core should not be allowed to fluctuate, 154 All mechanical components, including the power supply, of the volume heat source experiment for reactor cores with nonuniform flow cross section have been constructed and installed. System leak testing and flow calibration are now being carried out. An “infinitely’’ adjustable power input (that is, 15 to 100 kw) to the test section will be pos- sible by means of several saturable reactors that have been shown to possess excellent voltage regulation characteristics under the desired load. The electrical and power instrumentation is nearing completion. The temperature structure within an insulated, divergent test channel will soon be studied for low power densities. Several heat transfer analyses that are useful in predicting the thermal structure within a circulating- fuel reactor core have been completed. A report is being prepared which tabulates some of the detailed temperature profiles derived for the forced- convection volume-heat-source systems described previously.l'g These temperature data will make it possible to determine rapidly the complicated, radial, fuel-temperature profiles in circulating-fuel reactors whose pipe or channel ducts are being cooled at the walls. Also, o transient temperature solution was derived for the case in which fuel stagnates momentarily next to the Inconel core shell. Another analysis was concerned with a boundary layer temperature solution where a volumetric heat source exists in the fluid; thermal and hydrodynamic boundary layers are not presumed to be equal. ' FREE CONVECTION IN FLUIDS HAVING A VOLUME HEAT SOURCE D. C. Hamilton F. E. Lynch Reactor Experimental Engineering Division The objectives of the free-convection reseorch and the progress were previously reported Both the theoretical and experimental analyses of the three-parallel-plates system have been completed, and a report is to be issued.? The three-parallel- ]H. F. Poppendiek and L., D. Palmer, Forced Con- vection Heat Transfer in Pipes with Volume Heat Sources Wztbm the Fluids, ORNL-1395 (Nov. 5, 1953). 4. E. Poppendiek and L. D, Palmer, Forced Con- vection Heat Transfer Between Parallel Plates and in Annuli with Volume Heat Sources Within the Fluids, ORNL. 1701 (May 11, 1954), 3D. C. Hamilton and E. E. Lynch, ANP Quar. Prog. Rep. Dec. 10, 1954, ORNL-1816, p 113, 4D, C. Hamilton and F. E, Lynch, Free Convection Theory and Experiment in Fluids Having a Volume Heat Source, ORNL-1888 (to be issued). sl Do st AL plates system is shown schematically in the upper inset of Fig. 7.8; it consists of three parallel and equally spaced vertical copper plates, 8 in. wide and 36in. high, with a channel width, x4, 0f 0.5 in. Plastic bottoms and ends make leakproof containers of the two free-convection channels, which are filled with a dilute electrolyte (HC! in HZO)' With the two outer plates grounded and the center plate maintained at a given a-c voltage, heat is generated uniformly within the electrolyte. The heat thus generated is transferred by free con- vection to the outer walls and then through the outer walls fo the coolant channels in which cooling water flows vertically upward. Thermo- couple probe wells were provided so that the temperature within each wall could be measured at various vertical levels in the system., The variables measured were coolant flow rate and temperature rise, power input to the apparatus, and the three wall temperatures at various levels. A theoretical analysis for laminar flow was made that was based on the postulates that, in the region awdy from the top or bottom, the velocity profile becomes fully established and unidimensional and that the temperature becomes linear with the vertical coordinate, z, and has the same vertical gradient, A, as the coolant mixed mean tempera- ture. Both velocity and temperature solutions were obtained. The maximum temperature at a given level occurs at the center wall. The equation for the dimensionless center wall tempera- ture, ®(0), follows, and (D‘(O), Ny, and N, are defined on Fig. 7.8: sin?2 A + sin A sinh A ~ A sin A cosh A — A cos A sinh A PERIOD ENDING JUNE 10, 1955 Semiquantitative visual observations of the paths of suspended droplets indicated that the free- convection circuit existed as one long cell and that the velocity profile became established in the middle region and was similar to that predicted by the theoretical analysis. The onset of turbulence occurred at a Grashof modulus of 5 x 107, the same as that for ordinary free convection. A curve is given in the lower inset of Fig. 7.8 that shows the typical vertical variation of the experimental ®(0). |n comparing the laminar regime minimum ®(0) data with the theory, it is seen that, for N, less than 1300, the data are about 20% lower than the theory; this is good agreement. For these data, 4 was uniform and approximately equal to A , the coolant mixed mean temperature gradient. For the data (not shown here) for which Ny was greater than 1300, A was definitely not uniform; it was as much as ten times greater than A and the ®(0) data were 50% lower than the theory. From the turbulent regime data, it is seen that free convection can reduce the maximum tempera- ture difference in such systems to at least one tenth that which would be present if the heat were transferred by conduction alone. Variables not defined on Fig. 7.8 are: %, thermal conductivity; g, acceleration due to gravity; 3, temperature - coefficient of expansion; a, thermal _ diffusivity; v, kinematic viscosity; and §, volume heat source term. o ®(0) = )\fi’ [sin A éi_nh A - c-z_(_]!-—_ cos A cosh)\) + c(cos)\ si_nh_}t ~ sin A cosh M, sinh2 A — sin? A s (cosh)\ — cos A) (smh X+ sin A) — 2X sin Asinh c = - sinh2 A — sin? A 155 b ! ANP PROJECT PROGRESS REPORT " UNCLASSIFIED . ORNL—LR-DWG 7325 T — T T T ] LAMINAR REGIME e TURBULENT REGIME CENTER WALL N = ~— T LEFT OUTER WALL Ot I — T~ 4 | s Sl AT D [ ¢ .ZS.""--..__ T T 0.5 — — EQUATION 1+ o K s LEFT COOLANT PASSAGE S et ~ 1, y 7 d ""--.\‘. A — - ~= T~ ~ -~ w5 oz . 07§ J—‘\—-.. N ~ Jol = O MINIMUM $(0) EXPERIMENTAL DATA o U‘-"“&L 5 \\"13 i ® DARKENED POINTS GO WITH THEORY T of —— T— S oy A MAXIMUM $(0) EXPERIMENTAL DATA e & —— AVERAGE LINE DRAWN THROUGH DATA —— MINIMUM ®(0) FROM LAMINAR REGIME THEORETICAL ANALYSIS (EQUATION 4) 1.0 0.05 T J ‘ I i { H l - | < MAXIMUM ${0) | MINIMUM &(0 T T I $(0) @(/7 | ——— 7, CENTER WALL TEMPERATURE 1653 Bg 4 ? #, OUTER WALL TEMPERATURE ' av 0 | : TOP BOTTOM 0.02 —— Sx2 HEIGHT ' §.(0)= —= 810 TYPICAL VERTICAL VARIATION : 24 M= 5354 OF EXPERIMENTAL &{O0) " a’c oo L L | L L || 10° 2 5 10 2 5 10° 2 5 10° 2 W I Fig. 7.8. Dimensionless Temperature Function for N\, < 1300. HEAT CAPACITY Liquid (546 to 899°C) W. D. Powers Ho —Hyso o =~9.8 + 0.3508T - (5.39 x 10~°)72 Reactor Experimental Engineering Division c =0.3508 — (10.79 x 10~5)T The enthalpies and heat capacities of four P o AH = 63 at 510°C fluoride compositions and of lithium hydride have fusion been determined by using the copper-block calori- NaF-ZrF ,-UF , (56-39-5 mole %) meter. The results are listed below: Solid (]37 to 503°C) | H ~ Hygo =—4.3 +0.1596T + (5.15 x 107 °)12 NaF-ZrF ,-UF , (50-46-4 mole %) | c, = 0.1596 + (10.29 x 107°)T Liquid (540 to 894°C) ieuid (567 10 892°C) —ye? iqui to Hp ~Hygoo ==3.3+0.3178T - (4.28 x 107°)T Hy ~ Hygo o = 0.6+ 0.3033T — (3.24 x 10-%)1? =0.3178 — (8.56 x 10~5)T -5 “p c, = 0.3033 - (6.47 x 10~>)T AH, . =57 at 530°C NaF-ZrF, (50-50 mole %) | NaF-LiF-ZeF -UF , (20-55:21-4 mole %) Solid (54 o 488°C) ~ Liquid (582 to 900°C) Hp - Hygo o = ~44+ 0.1798T + (2.69 x 10512 Hop = Hygo o ==25.9 + 0.43WT ~ (7.42 x 10-5)7?2 c,= 0.1798 + (5.38 x 10=3)T c,= 0.4314 — (14.85 x 10~3)T 156 "f‘"\/f‘scosity measurer ' fluoride mixtures, and the results are tabulated’ /QAMM Lithium Hydride Solid (100 to 490°C) Hyp ~ Hypo =~38.5+ 0.939T + (6.6 x 1074)72 c, =0.939 + (0.132 x 10T In the above expressions, H; = enthalpy in cal/g, c, = heat capacity in cal/g-°C, p T = temperature in °C, AH, . = heat of fusion in cal/g. The enthalpies and heat capacities of the first two mixtures had been previously determined by using the Bunsen ice calorimeters. In view of the present importance of the zirconium fluoride-base fuels, however, it was felt desirable to determine these properties with the more precise copper- block calorimeters. The heat capacities of liquid fluoride mixtures may be predicted within about 15% on the basis of their chemical composition. It has been found that the product of the heat capacity (cal/g:°C) and a function M/N is remarkably constant. This function is defined as: M =S M, ? N =2xN;, —— z M = average molecular weight, . = molecular weight of component, . = mole fraction of component, = average number of ions, = number of ions in componenf funchon M/N For the ]7 fluorlde mixtures being *VLS*Ud'ed at bfesent ‘the “average of the préduct of:"' “and M/N is 9.0. The corresponding average producf for the six fluoride mixtures containing 'approxlmctely “equal ‘molar amounts of NaF and o IRy and from O to 7 mole % UF was found to S _'be 8.1 ' in Table 7.1. The data are expressed in the form nts were made on seven PERIOD ENDING JUNE 10, 1955 ST ORNL-LR-DWG 7326 1.0 ° NQF-ZrF4 MIXTURES WITH O TO 7 mole % Ul"'4 ® ALL OTHER FLUORIDE MIXTURES . ® 05 N\ p N 7 2 v::;,= 8.0 Hfi_u © A 9 >— k= NN g et TN . - M % o N T N\ * u& ‘ \ 0.2 e\ o.1 10 20 30 40 50 60 MIN Fig. 7.9. c, Vs M/N for Molten Fluoride Mixtures. where T is in °K. These viscosity data have been plotted in Fig. 7.10. No equation is listed in Table 7.1 for salt a because of the slight curvature in the data, which may be noted in the figure. Salts b, ¢, and 4 are mixtures which had been studied prior to the recently completed viscometry refinement program. Salts a, e, f, and g are Flé;Jre 79 shows s p[of of heaf cupacny T theff*-"mi)émres ‘which have been formulated recently. Measurements were made on all the mixtures - (except salt g) by using both the Brookfield and capillary viscometers; the results obtained by these two completely different instruments were in satisfactory agreement (deviations from the average line through the data were within 112%), Measurements on salt g, which contained BeF_, e made in a separate beryllium facility. Data 5 taken wnh 'rwo cuplllary wscometers to “furnish a check. =:Figure 7.11 presents a plot of the viscosities of seven mixtures containing BeF2 Mixtures g, i, and & were studied at ORNL. Mound Laboratory investigated mixtures » and j, and mixtures and m were studied at KAPL. The formulas of these mixtures in mole percentages, as well as the 157 D sl AT _ANP PROJECT PROGRESS REPORT TABLE 7.1. SUMMARY OF CURRENT VISCOSITY MEASUREMENTS a Vis&osity Miqufe Coniposition (cp) B Reference a - NaF-LiF-ZeF -UF .UF, At 550°C, 13.5 (b) o (20.9-38.4-35.7-4-1 mole %) At 800°C, 4.3 | _ b NaF-ZF 2 UF, At 570°C, 11 0.1307 3730/T () . (50-46-4 mole %) At 870°C, 3.4 ¢ NaF-UF, At 700°C, 10.25 0.1715 &>784/T (d) (66.7-33.3 mole %) At 900°C, 5.1 d NaF-KF-UF, At 600°C, 17 0.0866 ¢4611/T (e) (46.5-26.0-27.5 mole %) At 900°C, 4.4 e NaF-LiF-ZrF ;-UF ,.UF, At 600°C, 12 0.061 481V/T (N (20-55-21-3.6-0.4 mole %) At 850°C, 3.7 / NaF.LiF-ZrF, At 600°C, 12 0.061 4617 () (22-55-23 mole %) At 900°C, 3.1 g NaF-LiF-BeF, A+575°C, 7 0.105 3560/T (g) (56-16-28 mole %) At 800°C, 2.9 VISCOSITY (cp) 4See Fig. 7.10. bS. I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 72, ORNL CF-.55-3-61 (Mar. 8, 1955). €5, I. Cohen and T. N, Jones, Measurement of the Viscosity of Composition 30, ORNL CF.55-3-62 (Mar. 9,,1955), dS. f. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 43, ORNL CF.55-3-137 (Mar. 16, 1955). 95, 1. Cohen and T. N, Jones, Measurement of the Viscosity of Composition 2, ORNL CF-55.4-32 (Apr. 1, 1955). /s.1. Cohen and T. N, Jones, Measurement of the Viscosities of Composition 81 and Composition 82, ORNL CF.55-5.58 (May 16, 1955). 83, I. Cohen and T. N. Jones, Measurement of the Viscosity of Composition 78, ORNL. CF-55-5-59 (May 16, 1955). UNCLASSIFIED ORNL-LR-DWG 7327 ' TEMPERATURE (°K) 600 700 800 900 1000 1200 50 20 { 300 400 500 600 'TEMPERATURE {(°C} 700 800 {000 '_ Fig. 710 .V‘is'c:osi;ies of Fluoride Mixtures Currently Being Studied. ' 158 VISCOSITY {cp) 600 300 Fig. 7.11. Viscosities of Some Fi.uc.ariae_Mik)-&br.‘esi ; UNCLASSIFIED ORNL-L.R-DWG 7328 TEMPERATURE {°K) 700 800 900 400 500 Containing BeF,. 1000 m 600 700 800 TEMPERATURE (°C) o 1200 1000 . G oieode o i g A ok e Qo pod dlts PERIOD ENDING JUNE 10, 1955 TABLE 7.2, THE VISCOSITIES AT 700°C OF SEVERAL MIXTURES CONTAINING Be'F:Z - a' . o Viscosity BeF, Content \ Mixture Composition (cp) (wt %) Reference ~ g Ber-LiF-NaF 4,1 32.2 (&) 1 . ' (28-16-56 mole %) : ' ¥ b ' BeF NaF ' ‘ 4.25 40.0 (¢) o (38361 7 mole %) ’ ' ; . BeFyLiF 5.0 44.9 (@) (31-69 mole %) | i BeF2-NqF 6.25 47.0 {c) (44.4-55.6 mole %) k ' BeF -NaF | 6.75 45.8 () o (43-57 mole %) ' 1 BeF.z—NuF-U F4 ' 9.6 53.9 {e) (51.9-47.9-0.25 mole %) m Ber—LiF-ThFd-UF4 17 58.6 () (49.43-49.43-1.03-0. 12 mole %) %See Figs. 7.11 and 7.12. S, 1. Cohen and T. N. Jones, Measurement of the Viscosity of Composztzon 78, ORNL CF.55-5-59 (May 16, 1955). “Communication from J. F. Eichelberger, Mound Laboratory 45, 1. Cohen, ANP Quar. Prog. Rep. Mar, 10, 1955, ORNL-1864, p 143. . ?J. K. Davidsen, R. J. Herbert, and B. T. Morecroft, Fused Fluoride Homogeneous Reactor System for Submarine “ . Propulsion, (SAR Phase Il Study), KAPL-992. ' Personal communication from J. K. Davidson, Knolls Atomic Power Laboratory. : UNCLASSIFIED 1 T oriCLASSIFIED BeF, contents in weight percentages and the 20 e viscosities at 700°C, are listed in Table 7.2. Figure 7.12 shows a plot of the viscosity of each of these mixtures at 700°C vs the weight percentage of BeF, in each of the mixtures; note that the viscosity decreases as the BeF2 content decreases. Mixture g was formulated because it appeared from this tfrend that the kinematic viscosity might compete with the viscosities of salts in the NaF-ZrF ,- -UF, system. The kinematic viscosity of mixture g, BeF -LiF-NaF (28-16-56 mole %), was found to be ubout 15% higher than that of NaF-ZrF, UF (50-46-4 mole %), the salt presently being consudered as 'rhe fuel for the ART. "VISCOSITY (cp) + THERMAL CONDUCTIV!TY 3 - 0 0 200 30 40 50 e0 70 o ) W. D. Powers - - U Ber, CONTENT OF MIXTURE (wi %)~ Reactor Experlmentcl Englneermg Dw:snon ) T ?:Two methods for ‘the defermmdtton of fhermal ’ -~ Fig. 7.12. Viscosities at 700°C vs B‘éF2 Content conductivities of liquid fluoride mixtures are being i of Seven Fluoride Mixtures. Compositions of the investigated. A radial thermal conductivity cell mixtures are given in Table 7.2, has been redesigned and is being fabricated. The 159 T T TErT— i i, i s QDo it Al ANP PROJECT PROGRESS REPORT major difficulty in using the apparatus as originaily designed was the lack of sufficient guard heating to assure pure radial heat flow, A flat-plate system is now being investigated. The liquid is contained in cells consisting of two parallel flat plates of metal joined at the edges with 10-mil-thick sheet metal. The conductivity of the liquid in the cell can be calculated by knowing the area of the plates and the distance between them and by measuring the amount of heat passing from one plate to the other and the tempera- ture differences between them. Some preliminary data on a ZrF ;-base fluoride mixture are currently being obtained that are in agreement with previous conductivity data obtained with the variable-gap device. 160 ELECTRICAL CONDUCTIVITY N. D. Greene Reactor Experimental Engineering Pivision Conductivity studies of several aqueous electro- lytes have indicated the presence of a considerable amount of polarization within the experimental, platinum conductivity cell. Accordingly, a second potential electrode will be installed, which, when used in conjunction with an electrometer having a 10-megohm input, should reduce the effects of polarization. Similar conductivity measurements of some nonaqueous electrolytes (molten salts) will be made to compare the differences in the degree of polarization between the two types of electro- lytes. 2 s had been irradi . :;L',‘."_-};‘ANP Quar Prog. Rep Dec 10, 1954 ORNL 1816, p 120 2Chemical ‘analyses by the General Analysis Labo- PERIOD ENDING JUNE 10, 1955 8. RADIATION DAMAGE D. S. Billington J. B. Trice Solid State Division Preliminary results from a series of MTR irradi- ations of Inconel capsules designed to compare UF;- and UF ,-base fuels show no corrosion, no significant uranium segregation, and no changes in concentrations of impurities in the fluoride mixture. The loop in which a fuel mixture was . circulated in the LITR has been sectioned, and preliminary examinations have been made. The miniature loop designed for operation in a vertical hole .in the LITRis in the final stages of assembly, and charcoal traps to delay fission gases in the event of a rupture have been tested and incorporated in the cooling-air off-gas line. A tube-burst stress-corrosion apparatus has been assembled for insertion in the LITR, and an LITR- irradiated stress-corrosion rig is ready for exami- nation. The creep apparatus inserted in the MTR has completed six weeks of irradiation and is to be returned to ORNL for measurements. A theo- retical study has been completed which indicates that xenon will be sparged from the ART, as presently designed, in sufficient quantity to alleviate the control problem, MTR STATIC CORROSION TESTS W. E. Browning G. W. Keilholtz Solid State Division H. L. Hemph.iil Analytical Chemistry Division Two more of a series of paired 'c'apsules are contained 2 mole % UF3, and in the other capsule the fluoride mixture contained 2 mole % UF,. A control capsule for each mixture was also opened and analyzed. No corrosion could be found in any of these five capsules. One irradiated capsule containing UF that was examined earlier had shown slight corrosion. The next series of capsules that will become available for analysis consists of four capsules that will have been irradiated three times as long (six weeks) as those analyzed thus far. The capsules in this series will be more likely to reveal any difference that might exist between the two fuel materials. The starting fuel mixtures were analyzed? for trivalent uranium, total uranium, iron, chromium, nickel, sodium, and zirconium. The UFB-becring batch had more than 96% of its vranium in the trivalent condition, whereas no trivalent uranium was found in the UF ,-bearing batch when cnalyzed by a method which is sensitive to 0.005% ust, The results of the chemical analyses for uranium, iron, chromium, and nickel in the starting mixture, the control mixtures, and the irradiated mixtures were averaged, and ratios were computed to deter- mine whether there were changes in distribution or composition. The results of these statistical analyses, which are presented in Table 8.1, indicate that there was no significant uranium segregc’non and that there were no chonges in concentration of urqmum, iron, chromium, and nickel. being ‘irradiated for two weeks each in the MTR o These Inconel capsules are bemg wradmted - order to compore the effects of UF3 and UF, - NaF- ZrF -base fuel mixtures, as described prevu- e ously Three addltlonal Inconel capsules that ha ted for two weeks were openedw ‘and anaiyzed chemlcafiliy and metqllogrcphlcaily if,}.;_ln two of fhese cop5u|es the fluorlde mlxture” 1W E Browmng, G. W. Kelrlhohz, und H. L. Hemphili, ratory, Analytical Chemistry Division, 3Metc|lographic analyses by the Remote Metallography Group, Solid State Division. Chemical analyses were also obtained of four additional irradiated inconel capsules in which the fluoride fuel mixture initially contained 4 mole % . UF,. _The uranium analyses of inner cores and outer cores of the samples agreed to within 6%. The iron, chromium, and nickel concentrations .. were between 0.01 cmd 0.1%. Additional capsules f this type are belng prepared for :rradmhon in .,,.__:th‘et MTR at 1500°F for control runs, and for thermal cyclmg tests. ' 4Anczlyses performed by the Y-12 Analytical Labo- ratory. 161 b L ke F ' oL TABLE 8.1, SUMMARY OF ANALYTICAL RESULTS FOR SERIES OF CAPSULLES TESTED TO COMPARE EFFECTS OF UI"'3 AND UF, IN N.uF-ZrF4-BASE FUEL MIXTURES - LH0dIY SSIYO0Yd LDIF0dd INV Number of Uraniom Distribution {a) Ratio of Final Ratio of Final to Initial Capsule Category Capsules U (av)/U, (av) to lnitial Impurity Concentrufion(c) a b Uranium Content( ) Fe Cr Ni All capsules in series 7 1.003 *+ 0.028 1.021 % 0.022 1.47 +>1.47 1.08 +>1.08 (d) All irradiated capsules 5 1.021 t 0.033 1.038 £ 0.024 1.45 £ >1.45 0.48 + 1.5 1.48 + 0.56 All control capsules 2 0.96 = 0.0240 0.989 * 0.02(e) 1.53 +>1.53 3.0 + 1.2 (d) Irradiated capsules containing 2 1.026 £ 0.044 1.031 = 0.04(e) .67 + 1.47 0.0 + 1(2) 3.9 + 1.1 2 mole % UF4 Control capsule containing 1 0.947 + 0.028 0.975 + 0.024 (D {d) (d) 2 mole % UF'4 lrradiated capsules containing 3 1.001 £ 0.048 L051 £ 0..05(6) 1.31 +>1.31 0.54 + 0.6 0.95 £ 0.05 2 mole % UF3 Control capsule containing | 0.972 1 0.020 0.994 + 0.02(e) 1.53 £>1.53 3.0 + L2 {(d) 2 mole % UF3 ' (a)Ua is the uranium content of the inner core of a sample, and Ub is the uranium content of the outer core of a sample. (b)Correcfed for burnup. (C)Confidence limits are ot 95% level. (d)Vcflues obtained were influenced by a sample which was probably contaminated with Inconel and therefore are not significant. (e)Esfimufed. .y - Lo sl LITR HORIZONTAL-BEAM-HOLE FLUORIDE-FUEL LOOP 0. Sisman J. G. Morgcn C. D. Baumann M. T. Morgan W. E. Brundage A. S. Olson R. M. Carroll W. W. Parkinson Solid State Division The operation of the loop for circulating fluoride- base fuel in the LITR horizontal beam hole to study the effect of reactor irradiation on corrosion and fuel stability was described in the previous report.> The fuel mixture was composed of 62.5 mole % NaF, 12,5 mole % "ZrFA, and 25 mole % UF,. The loop was cleaned by running it at 1500°F for 16 hr with a non-uranium-bearing salt. The actual fuel mixture was circulated in the loop for 645 hr, including 475 hr during which the LITR was at full power (3000 kw). The power generated in the loop -was about 2.8 kw, and the linear flow rate of the fuel mixture was 8 to 10 fps, which corresponds to a Reynolds number of 5000 to 6500. The dis- assembly of the loop to provide specimens for metallurgical examination and chemical analysis and to reduce the loop to pieces of convenient size for fuel recovery has been completed. After perotion of the loop in the LITR, the section between the linear seal flange and the pump was sheared with remotely controlled equip- ment to permit removal of the pump in its shield. The loop in its water jacket was then fransporfed fo the hot celis of the Sol:d qu’re Dwuszon for the remote dlsassemb!/ operation. The out-of—plle end of the jacket was drawn from the withdrawal shield with the spec:al dlsassembiy appcsrc:fus show 30. Sisman et al., ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 150, - system with nonradloactlve material. -~ pump enciosure was removed, the insulation and - the various wires and heoters were taken off to z= provide access fo ‘the pump bowl and the fuel tubes. After fhe_ fuel tubes were severed and the nds were clamped fhe pump “bowl was cut ‘usf’ above fhe fuel leve] by usmg the bcnd saw (Fi ig. . 8.5). ‘There : all” 51 through a hole in fhe rear wall of a hot cell equ;pped “-':ffi'ws;ble on the upp _ _ = the cct:vnty ‘was high enough fo |nd|cate fhe_‘ presence of a cons:derable quam‘l'ry of fnssmni ' . prov:ded access to the loop for str:pplng of thermal ™~ the upper porhon of the A v e T ’ o S * PERIOD ENDING JUNE 10, 1955 insulation, heaters, wires, and graphite shield blocks. The loop is shown partially stripped in Fig. 8.3, as photographed through a window of the hot cell. ' Tube sections about 2 in. long were cut from the fuel tubes for metallographic specimens. These specimens were taken at about 6-in. intervals from the part outside the jacket seal flange and at 2-ft intervals inside the flange, between the heat ex- changer and the irradiated section. Additional specimens were taken from the heat exchanger and the venturi for examination of a longitudinal section. The U bend of the irradiated section was cut intfo five specimens from ]/2 to 1 in. long, and additional specimens were cut from the remaining 18 in. of the irradiated section at about 6-in. intervals, The fuel was removed from these specimens by melting in an argon atmosphere. Four 2-in. sections of tube were also taken from the portion of the loop between the jacket flange and the pump, and two more were taken 6 in. back - from the ‘irradiated section to furnish samples for chemical analysis of the fuel. Disassembly of the pump was complicated by its size and by the large stainless steel enclosure around the bowl (Fig. 8.4). A 300-amp, d-¢ welder equipped with a high-frequency-arc starter was used to cut the enclosure successfully. It was -possab[e to guide the cutting electrode with a - manipulator while cutting the sheet metal in a manner that prevented contamination of the fuel After the o’f fuel (~1 g) ofrthe pump “baffles, buf too high to permit unshlelded dlsmanflmg and examination. !t was assumed that fission products which had vaporized from the fuel were the sources 163 S " ANP PROJECT PROGRESS REPORT UNCLASSIF [ED PHOTO 13766 R RENCH FOR MANIPULATOR s YDRAULIC MANIPULATOR CANS FOR SHIE B L | L VACUUM CLEANERS FOR SAW “SHIEL T R FRTPITR S LN | | k4 D W% ) FURNACE POT WASTE CAN [ T = ' il i N Fig.» 8.1. Top View of Hot Cell and Equipment for Dfsassémbly of LITR Fluor.i'de-.FuAei. Loop. o AR AR B, i MG R SR s T o el . A sl s Sl o FUEL TUBES AND HEATERS= o 165 il E e i e Do so? LM ANP PROJECT PROGRESS REPORT UNCLASSIFIED PHOTQ 13889 FUEL TUBE gl ’ ~ Fig. 8.3. Partly Disassembled LITR Fluoride-Fuel Loop. of the activity, and, since an analysis of the condition of bearing and seals was precluded, this portion of the pump was discarded. The fuel- containing parts were stored for uranium recovery. The metallurgical specimens have been prepared and photomlcrogrcphed A preliminary analysis indicates that subsurface void formation was generally uniform throughout the loop to a depth of 0.8 mil. The uniformity of attack possibly resulted from ihe Icck of a femperature differential in the loop. The temperqture record shows that the fuel temperature could not have been more than 20°F lower in the out-of-pile parts of the loop than in the in- p||e parts. Some sections showed grain growth, which is an indication that some tempera- tures were h\gher than those recorded, but even the areas of grain growth showed no significant __chq\nge in depfh of corrosion. , " The fuel in the nose of the loop was in a maxi- 166 mum perturbed flux of approximately 0.8 x 10" neutrons/cm?+sec, which corresponds to a fission power density of 0.5 to 0.6 kw/cm3. The power density gives a measure of the severity of con- ditions imposed upon at least one section of the loop tubing by radiation effects. No difference in the corrosion could be observed between this region and other, less severely irradiated, regions. Also, about the same depth of attack was found in the venturi section of the loop as was present in the rest of the loop, even though there was a large velocity gradient present. No intergranular attack was noted. N _ The six sections of tubing tcken for chemical analysis were sampled by a drilling method de- scribed previously,® After the end of a tube section 8¢, C. Webster and J. G. Morgan, Solid State Semiann. Prog. Rep. Feb. 28, 1954, ORNL-1677, p 27. A RS . e o o i A G e, SR il L o o “w . .“.‘. - wrrh a Flg 8.4. LITR Fluoride-Fuel Loop Pump Priot ‘to Removal of Enclosure for Providing @ Helium Atmosphere. “had been cleaned by drlllmg a /-|n. hole abouf l/ in. deep in the fuel, each sample was bored out ./16"”‘ bit. Tungsten carbide—~tipped bits used to avoid conrammahng the fuel with ‘rhe place where ‘this llne passed through the pump “shield. The chemical analyses of the samples arewr “almost complete, “and radiochemical analyses for - Cs 71_37__, and Zr 95 have been “made__”_of two_ of the _. amples.” elecfrlcal measuremem‘smade durmg 'operaflon of the loop The resul'rs fromfl‘fhe Cs _ 7_analy5|s indicate escape of the Xe 137 oarent from the fuel. In orderfo measure the fission power distribution along the irradiated section of the loop, an experi- nonrad:o‘achve _iron, chromium, or nickel, = Un- ' lrradlared control(samples were obtalned from the flll lme between the charge ‘tank and the loop at“ ' : tamed markedly more Ry taken upstream. This difference is exactly that “The frssron power of 2.7 ancl 55 kw R - \m'dtcared for the two samples by ‘the Zr®3 aont_entm co pareslfqvombly with fl'ie ? 8 kw_‘vesfjimaiecl from, “ 2.1 and 1.5 kw, “are cons;derably lower,fl and they removal rate of Ru PERIOD ENDING JUNE 10, 1955 ment was conducted in the LITR with a mockup of the loop. The mockup was designed to duplicate both the- arrangement of structural materials ‘and the macroscopic cross section of the fuel so that effective thermal-neutron flux values could be ‘measured inside the fuel and then could be used to calculate the power generated by fissioning. ‘The results are being compared with the activation of cobalt monitor foils attached to the loop and with additional data from the activation of borings taken ~from the metallurgical specimens. The results will be reported when all the data have been assembled and analyzed. DEPOSITION OF Ru'%3 IN LITR FLUORIDE-FUEL LOOP M. T. Robinson Solid State Division, T. H. Handley Analy’rlcal Chemistry Division The deposn‘:on of the fission- producr 40 day Rulq3 on the surfaces of the fuel circuit of the " ARE was reported prevrously Supporhng evidence for this phenomenon has now been obtained from a siudy of samples taken from the loop described above. Two samples of tubing from the loop were examined by gamma-ray scintillation spectrometry. One sample was taken upstream from the high-flux region of the loop, and the other was taken at an equal distance downstream. The ratio of Ru'®3 to Zr?3-Nb%5 should have been 0.6 if the Ru'®® had remained uniformly distributed throughout the loop. However, the ratio for the sample taken upstream was 5.33, and that for the sample taken downstream was 8.07. Within - the experimental error, the Ze?5-Nb93 activity analyzed the same for both samples. The' sample """"" 103 fhan did the sample to be expected from rapid reaction of the ruthenium in the fuel with 1he container metals. These ‘numbers, comblned wn'rh operahonal and dimensional data, yielded a hczlrc llfe of 0.5 min for the rate “constant for removal of l?uw3 from the fuel in the loop. It was assumed for “the. calculatlon that the 103 was a first-order function of the concentration. M. T Robinson, S, A, Reynolds, and H. W, Wright, ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 13. 167 e No st delle ANP PROJECT PROGRESS REPORT UNCLASSIFIED PEOTO 14134 %( 23 Fig. 8.5. Bottom View of Pump Showing Accumulation of Fuel on Upper Parts of the Baffles. After a delay of 53 days, the two pipe samples were re-examined to determine the apparent half lives of the two components of the gamma-ray spectrum. For the Ru'%3 peak (0.50 Mev), the half lives of two samples were 42 and 43 days, respectively, and amply confirmed the assignment of that activity. For the Zr?5.Nb?5 pedk, however, the results were 40 and 43 days, respectively, instead of the nearly 65 days expected. The contribution to this gamma-ray energy (0.78 Mev) must be largely from 37-day Nb%3, the 65-day Zr’3 being very low in relative amount (a few per cent). It appears very likely that niobium deposits on the Inconel pipes along with ruthenium. The deposition rate for Nb?3 cannot be estimated, at present, since it is produced very slowly from its Zr?5 parent, rather than rapidly from fission as is the case with Ru 193, Radiochemical analysis of a fuel sample for Ru'®® indicated that the amount was below the 168 limit of detection by that technique.® This demon- strates that the efficiency of ruthenium removal was comparable to that attained in the ARE.” No direct neutron activation of the elements of Inconel was observed. MINIATURE IN-PILE LOOP W. R. Willis M. F. Osborne H. E. Robertson G. W. Keilholtz Solid State Division The miniature fluoride-fuel loop for insertion in a vertical hole in the LITR (described previously®) was charged with fuel and operated on the bench. It is now in the final stages of assembly for in-pile testing. An exact mockup of the experiment has been insertedin the reactor and has been withdrawn 8 . . . W. W. Parkinson, personal communication, 9W. R. Willis et al., ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 147. aliainie o i ol W by using the special withdrawal cask. The operation proceeded smoothly DELAY OF FISSION GASES BY CHARCOAL TRAPS D. E. Guss, United States Air Force W, R. Willis, Solid State Division In the event of a leak in the miniature in-pile loop mentioned above, the fission gases released into the cooling-air stream must be contained, and at the same time, the cooling air must continue to flow until the temperature of the system has fallen below the control set point. To accomplish this, two traps, 24 in. long and 10 in. in diameter, filled with 6-14 mesh activated cocoanut charcoal, have been mcorpora’red :ri the cooling-air off-gas line. These traps, whlch are connected in series and operate at room temperature can be expected to hold krypton activity in the system at the design air flow rate of 20 cfm for 30 sec before the release of appreciable activity. If the cooling air then continued to flow at the design rate, the krypton " activity, which would be released over a period of about 5 min, would reach a peak in about 2 min. However, since the air flow would stop in less than 1 min after a leak occurred, virtually all the krypton activity would be contained in the traps and would trickle out over a Iong period of time. Xenon, the other active gas present in quantity, will be held back for a much longer period of time; hence, the trap deS|gn was based on the krypton holdup time. Since the action of the charcoal is to increase the effective volume of the trap space by a factor VWthh depends dlrecfly upon the amount of chcrcoal_ e used, it was possnble to eshmafe ‘the frap size S neededwforrfhe_m p||e experlment sumply by scahng N expenmental_' Japparatus mcheolly in Fig. 8.6. Kryp'ron was |n'rroduced |n'ro"xi_(:_'” a‘reservoir beneath the frap ond c'r time zero, was drwen ‘through the” Yrap with air. "It then passed through a flowmeter and over a counter, where the activity as a function of time was observed. All the runs were made at 23°C. N PERIOD ENDING JUNE 10, 1955 The two small traps were tested at flow rates in the region of interest, and all the elution curves obtained were similar, the activity rising quite rapidly and almost linearly to a peak and falling off more gradually. A typical elution curve is shown in Fig. 8.7, in which the time coordinate is UNCLASSIFIED ORNL-LR-DWG 7330 RADIATION DETECTOR CHARCOAL TRAP Fig. 8.6. Schemahc Diagram of Charcoal-Trap Experiment. i ORNL-LR-DWG 7334 10 * o 0.9 / \\ / ® AIRFLOW RATE: 6600 cm>/min 08 ] \ > L = | \ »n 0.7 = . \ = o/ . z 06 \ o - \ o / ® < 05 \ o ) | \ = < 0.4 4 ® w \ / e : //0 o - T e e \.. | J , | \. 5 | 4 0 20 40 60 ‘TRAP VOLUMES OF GAS ELUTED (cm?) Fig. 8.7. Elution Curve for 334.cm® Charcoal Trap Containing Krypton. 169 " &OMW ANP PROJECT PROGRESS REPORT v;ekpres.sed' in trcp vol'umes of gas eluted. A plof of - the inverse of the flow rate vs time to peak activity for the 23.5-in.- -long, 1.05-in.-dia trap (shown in “,_',’VFlg 8. 8) indicates a linear relationship. The point . on ‘|’|‘|IS curve corresponding to 'rhe design flow rate ' -*57:;1'15 shown. _ The e[u‘rlon curve obtained at the in- -pile design . flow rate is shown in Fig. 8.9. Since the effect ~ of two traps in series is to double the time to peck “activity and to broaden the elution curve, it is planned to rely upon two charcoal traps 24 in. long and 10 in. in diameter to delay any burst of fission gases from the in-pile loop for 30 sec before an appreciable amount of activity is released to the sfcck. CREEP AND STRESS CORROSION TESTS W. E. Dav:s J. C. Wilson N. E. Hinkle J. C. Zukas Solid State Division Heat transfer experiments have shown that heat dissipation in the longitudinally finned, vertical cylinders in the water-jacketed, helium-filled enclosure of the tube-burst, stress-corrosion appa- ratus, as presently designed, % will limit fhe fluoride salt power densities to about 500 w/cm°. Therefore fins have. been added between the salt container and a surroundmg water |acket and the heat transfer has been increased sufficiently to permit opercmon at power densities greater than 1000 w/cm®. The first complete apparatus has been assembled and awaits f||||ng before irradi- ation in the LITR. Five more rlgs are bemg fabricated. | - Tests are under way to determine the suitability of pressure-regulating valves for supplying gas to stress the tube-burst specimen. Two pressu're'- volume fransformers with 1:1 and é: I ratios were designed and are being fabricated. These units are intended to isolate from the remainder of the system the gas used to stress the specimen so that rupture of a specimen will result in a known pres- 195, C. Wilson et al., Solid State Semiann. Prog. Rep. Feb, 28, 1955, ORNL- ]851 p 3. PeRET . ORNL—LR—-DWG 7332 20 . / 2 15 »—0—7'/ > o 1 £ ' / < < E ¢ / ot u )P £ 10 - 1 =z / o . 3 re-1FEB 1,1955 - ' A MAR 91955 u DESIGN POINT a FOR LITR TRAP s 5 | g = /—0—4/ ’I-.-l b;é-f_'_' .fi 0 0 100 200 300 400 50C 600 700 DELAY TIME TO PEAK OF ELUTION CURVE (sec) ) FlgSS Inverse of the Air Flow Rate vs Time to Peak Activity for o 23.5-ifln:-'l._.<;‘n§", 1.05-in.-dia ~ Charcoal Trap Containing Krypton. s R D iR S o &MW - SECRTT ORNL-LR-DWG 7333 C 850 40 30 | | | | | | | l | \ AR FLOW RATE: 20 cfm | | | | RELATIVE RADIATION INTENSITY 20 ][ \- 10 } % 0 100 200 300 TIME (sec) " Fig. 8.9. Eluhon CU;Ve for 30:,900-cm3.Char<.:ou| Trap Containing Krypton. : _to 0.0001 in. or better, so that the stress may be accurate to 1%, several methods of checking wall ~ thicknesses are being studied. PERIOD ENDING JUNE 10, 1955 The stress-corrosion rig irradiated in the LITR previously ! will be opened when hot-cell arrange- ments have been completed. A transfer carrier for bringing subsequent rigs from the LITR to the hot cells in Building 3025 was designed and is being fabricated. The MTR creep apparatus has just completed six weeks of irradiation in the MTR and will be returned to ORNL for measurements. A THEORETICAL TREATMENT OF Xe!33 POISONING IN THE ARE AND THE ART M. T. Robinson Solid State Division A theoretical study of xenon poisoning in a circulating-fuel reflector-moderated reactor was made in an attempt to understand the behavior of the ARE and to exirapolate this experience real- istically to the ART. The system was assumed to consist of two phases: the liquid fuel and the sparging gas (helium), The theory deals only with volume-averaged concentrations and neutron fluxes. Turbulent motion of the two fluids was assumed to assure thorough mixing within each phase. With these assumptions, the differential equations describing the behavior of the poisoning were derived and solved. The various processes are iliustrated schematically in Fig. 8.10. The rates of all processes, except the two Xe'S3 production rates, were assumed to be first order; that is, the rates are proportional to the Xe 135 concentration in the starting phase. The rate constants for the two phase-transfer operations are related by application of the law of mass action, cm:di thus the eqwhbrlum solubilify of xenon in the .. fuel is mfroduced The Xe 135 pouséhmg of a fluid- fueled reactor '\__under steady-state condmons is gwen by (a r a,)k °°,.-~ 2 -mf&?ifffif_ Xw= sfeady state Xe]35 -poi'sbni.ng (%), Ty = 'IOnyeaf T, . ay = 00y /o,, ”W. W. Davis et al., ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 155, 171 Do et AL ANP PROJECT PROGRESS REPORT SECREY SSD-A-1167 ORNL—LR—DWG 6430A PRODUCTION FROM DECAY OF 1'°° LOSS BY FLOW OF TRANSFER OF xe'3° ‘ LOSS BY N o GAS FROM SYSTEM Xe' 29N LIQUID TO GAS xe'3d N THERMAL-NEUTRON ABSORPTION o ' SPARGING -GAS 135 | LIQUID-FUEL — . lLOsSBY PHASE TRANSFER OF Xe™> PHASE _ RADIOACTIVE DECAY GAS TO LIQUID * LOSS BY RADIOACTIVE DECAY DIRECT PRODUCTION IN FISSION Fig. 8.10. Processes Governing Xel35 Poisoning in Liquid-Fueled Reactors. k]=a4+)\f+)\L, ky=a, + azfi)\f+ Ag s o, = RTS, B = VL/VG ! A'I_. = O-Xe(’l‘S /\g = vG/VG , )‘f = rate constant for transfer of Xe 135 from liquid to gas, )/xe, y, = fission yields of Xel35 1135 f' Xe = thermal-neutron cross sections for Y235 tission, u23s absorption, S “and Xe 135 absorption, e d4' e radioactive decay constant of Xe " ¢ = volume-averaged thermal-neutron 0 flux, 135 VL’VG = volumes of liquid, gas phases, ve = sparging gas volumetric flow rate, S = séiub”ity of xenon in fuel at l-atm ~ Xe pressure, . T = absolute temperature, 'R = universal gas constant. - The results of the calculations made by the equation and the ART data given in Table 8.2 are presented in Fig. 8.11, which is o design chart for the estimation of the sparging-gas flow rate for conceivable levels of Xe'3® poisoning. The - parameter 7, (Fig.8.11) is the phase-transfer mean life from the liquid to the gas: 1 / R A Thé \-r;:iue of ;rf 1"ol Ee chosen is unknéwn, but ARE 72 experience indicates a value of about 20 min for that reactor. |t is to be expected that the value for the ART will be considerably smalier. A study of the poisoning behavior of the ART was made for two different types of shutdown. |n each case, it was assumed that 7, = 5 min and that = 1000 STP liters/day. At stead¥ state during full nuclear power operation, the Xe'>> poisoning would be about 0.40%. If the nuclear power was reduced to zero but sparging was continued, the poisoning would not rise by more than 1 or 2% of the steady-state value before starting to decrease. If sparging was discontinued, the poisoning would rise in about 11 hr to a maximum value of about 12%. If at this point sparging was resumed, about 36 min would be required to reduce the poisoning to below the steady-state value. Intermediate values of 10, 5, 2, and 1% poisoning would be reached in times of about 1, 7, 17, and 25 min, respectively. It appears highly probable that no difficulties will be encountered from Xe'3® during short shutdowns, since removal of the xenon to acceptable levels by sparging can be made within the time necessary to force the fuel from the dump tanks into the reactor. A series of calculations has been performed on the Oracle to study the approach of the Xe 135 poisoning to its steady-state value. In all cases of practical interest, this approach is controlied by the rate of 1'35 production. In fact, except for a change in ordinate, the poisoning is found to follow the iodine growth curve. It is concluded that the removal of Xel3% by sparging with helium (or other inert gas) appears to be a satisfactory means of controlling poisoning in the ART. No difficulties seem to exist in connection with shutdown behavior of the poisoning at the sparging rates selected for ART operation. £ ‘s PERIOD ENDING JUNE 10, 1955 TABLE 8.2. DATA USED TO CALCULATE POISONING IN ART ; o 7 - ' : A " Numerical Data * - ag = 0.254% R = 82,0567 cm3-01'm/moie-°K : ay = 4.74% | T = 1033°K (1400°F) ; 0 a, = 0.0509 S = 6 x 1077 moles/cm>+atm Q, = 2.09 x 10 s sec-] Oyo = 1.7 % 106 barns* ) e _ . Reactor Data™* - ARE ' ART v, 535 5.64 f1° ' .3 | 3 VG 1 .fl' | 0.31 ft v 0.25 em>/sec 1000 STP liters/day ¢. 8 x 10! neutrons/cm? sec | 1 x 10" neutrons/cm?-sec B 5.35 | 18.2 A’L 1.4 x 10;6 sec“] 1.7 x 10_4 sec—] - A K_. Ergén:-c;na{ HW Bé}fi‘ni',‘A'NP indr.:t;P'r?o"g. Rep. Ma;. IO, 1955, ORNL- 1864, p 16. "'-,‘_;' **ARE datae from ARE Nuclear Log Book and from J. L. Meem; ART data from J. L.. Meem and W. T. Furgerson. ECRE™ ~ SSD-B-1168 ¢ 5 ORNL—LR—DWG 64 31A A e T 1 Ty = 20 min “xe'3% POISONING (%) ;= 10 min =5 mlin Tf=2m‘in L [ “Tf=Om‘in" P . T 3000 0 40000 5000 " SPARGING-GAS FLOW RATE (STP liters/doy) Fig. 8.11. Stéu'dy-Sfafe Xe'®3 Poisoning in_ART as a Function of Sparging-Gas Flow Rate for Various Assumed Values of the Phase-Transfer Mean Life, Ty 173 W ' Do ot bl - ANP PROJECT PROGRESS REPORT 9. ANALYTICAL CHEMISTRY OF REACTOR MATERIALS C. D. Susano J. C. White Analytical Chemistry Division Modifications of the method for the determination of uranium metal in fluoride-base reactor fuels were completed. Further studies were made on the evaluation of the methylene-blue method for the determination of trivalent uranium in fluoride-base reactor fuels. Investigations were continued on methods for the determination of oxygen as metallic oxides in fluoride salts. An improved separation method invelving anion-exchange resins was de- veloped for the determination of alkali metals in fluoride salts. A volumetric method for the deter- mination of zirconium in fluoride-base fuels was proposed. o DETERMINATION OF URANIUM METAL IN FLUORIDE SALT MIXTURES A. S. Meyer, Jr. B. L. McDowell Analytical Chemistry Division The apparatus for the determination of uranium metal in fluoride-base fuels according to a method based on the decomposition of the hydride in an CLOSED-END MERCURY-SEALED I \ OIL MANOMETER TO VACUUM Cu0 TUBE PUMP " LIQUID NITROGEN COLD TRAP MERCURY PRESSURE INDICATOR COMBUSTICON TUBE FURNACE atmosphere of oxygen at reduced pressure! is shown in Fig. 9.1. This apparatus has been cali- brated and has been applied to the determination of uranium metal in UF; and KF-UF . Samples of the uranium metal are placed in a platinum boat in the combustion tube and heated to 250°C for 1 hr in an atmosphere of hydrogen in order to convert any uranium metal present in the sample to the hydride. The excess hydrogen is evacuated from the system after the sample has been cooled to room temperature in an atmosphere of hydrogen. The UH, produced from the uranium metal is oxidized by heating at 400°C for 20 min in an atmosphere of oxygen: ' AUH, + 70,—> 4U0, + 6H,0 The combustion gases are passed through a copper oxide tube at 500°C to ensure complete conversion of the hydrogento water. The water vapor produced IA. S. Meyer, Jr., and B. L. McDowell, ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 158. UNCLASSIFIED ORNL-LR-DWG 7334 TO KIPP GENERATOR ASCARITE SAMPLE y 7 LIQUID NITROGEN ’\\ COLD TRAP N FURNACE 0z ALCOHOL AND MgC102 DRY ICE " Fig. 9.1. Diagram of Apparatus for the Determination of Uranium Metal. 174 B el oo ik L " Do et by the oxidation of the combined hydrogen is collected in the cold trap, and, after the oxygen has been removed by evacuation, the water is allowed to expand into a system of known volume. The pressure of the water vapor is measured on a closed-end mercury-sealed oil manometer similar to that described by Naughton and Frodyma.? The pressure registered on the manometer is related to the weight of the uranium present in the sample by empirical calibration of the apparatus. A preliminary calibration of the apparatys with BaCl,-2H,0 as.a standard for hydrogen in the form of water indicated a linear relationship between scale reading and weight of hydrogen for the range 0.18 to 0.8 mg of hydrogen. This linear relation- ship does not apply to larger amounts of hydrogen because the pressure in the system is approxi- mately equal to the vapor pressure of water. Results which have been obtained for the cali- bration of the apparatus with samples of pure uranium metal are in agreement with the BaCl,-2H,0 " calibration, and they indicate that a scale reading of 1 mm represents 0.36mg of uranium (Fig. 9.2). The results of the calibration with uranium metal have a coefficient of variation of 7% for the range 6 to 60 mg of uranium. ' ' UNCLASSIFIED ORNL—LR-DWG 7335 25 T T T T e OBTAINED FROM CALIBRATION WITH BaCly- 2 H,0 AS A STANDARD = 20 —— FOR HYDROGEN (N THE FORM OF WATER -——8 = O OBTAINED FROM CALIBRATION € 45 L. WITH URANIUM METAL | a i ‘ = o i : w 10 /‘/ - : < - . 2 & . . o/d : Flg. 9.2. Calibration of Apparotus For Determl- ation of Uranium Metal T e R samples represented portions of the material J. J. Naughton and M. M. Frodyma, Anal. Chem. 22, 711 (1950). 2 i . L- 9. r'hICh were onrolyzed for uranium metal were fcken VPrOg Rep. Dec: 10, 1954, ORNL-1816, p 12 by drillmg with” |ncreasmg|y |crger bl’rs so that the e PERIOD ENDING JUNE 10, 1955 successively nearer the wall of the reaction tube. Samples of UF; and fluoride salt mixtures which were to be analyzed for uranium metal were ground in an atmosphere of hydrogen to prevent oxidation of the uranium metal. The samples, along with the necessary equipment for grinding, were placed in a plastic bag, which could be alternately evacu- ated and filled with hydrogen. The samples were ground to a particle size that would pass a standard 325-mesh sieve. Several samples of UF; which had been produced by the reduction of UF, with uranium metal have been analyzed for uranium metal in the new appa- ratus. The results for the uranium metal content paralleled those obtained by petrographic analysis of the material. The standard deviation for dupli- cate determinations on the samples of UF, was approximately the same as that obtained for the calibration of the apparatus with uranium metal. The difficulties encountered earlier in the decom- position of the UH, in certain gaseous atmos- pheres" have been completely eliminated in the present method. The period of several hours which was required for the decomposition of the hydride in an atmosphere of carbon dioxide was reduced to approximately 20 min when the ignition was carried out in an atmosphere of oxygen. Tests indicate that no interference is introduced by the presence of fluoride salts or trivalent uranium in the samples, and therefore the procedure should be applicable to analyses of all types of fluoride fuels. A topical report on this investigation is being written. DETERMINATION OF TRIVALENT URANIUM INFLUORIDE FUELHS A. S. Meyer, Jr. W. J. Ross Analytical Chemistry Division Oxidation of Trivalent Uranium by Mefhy|ene Blue ' jtohiparison ‘of the methyle'\e-blue “and the hy'drogen evolution® methods for the determination A, s Meyer, Jr., and B, L. McDowell, A S. Meyer, Jr., D. L Mannmg, and W J. Ross {151\}1’ Quar Prog. Rep. Mar. 10, 1955, ORNI -1864, p D L. Meanning, W. K. Miller, and R. Rowan, Jr., Methods of Determination of Uranium Trifluoride, ORNL.- 1279 (April 25, 1952). 175 of trivalent uranium in fluoride-base fueljs has been ANP”Quar | T T o R, LD»MM ANP PROJECT PROGRESS REPORT completed. Portions of eutectic mixtures, which had been pulverizedand sampled under inert atmos- vheres, were analyzed simultaneously by both methods in an effort to minimize the effects of the heterogeneity and the instability of such samples. An evaluation of the results showed that satis- factory agreement, comparable to that reported previously? for other eutectic mixtures, was achieved in the analytical results for samples of LiF-BeF,, NaF-LiF, and NaF-BeF,. The results obtained by the methylene-blue method in the determination of trivalent vranium in KF-UF ,-UF, eutectics continue to exhibit negative bias when compared with the results from the hydrogen- evolution method. No further studies were made of NaF-ZrF -UF ; eutectics. All the mixtures studied could be dissolved in methylene-blue solutions that were 1.5 to 6 M in HCl and saturated with AlCIS. No hydrogen is evolved during the dis- solution of the most reactive trivalent wuranium ‘compounds, such as U'Ci-s, in 1.5 to 3 M HCI solutions of methylene blue if efficient agitation is maintained while the solvent is being added to the flask. Methylene-blue sclutions that are 1.5 to 6 M in HCI are reduced, during dissolution at room temperature, to methylene white by finely divided metallic chromium, nickel, iron, and uranium-nickel alloy. Zirconium is only very slightly soluble in such solutions over a 2-hr period. The presence of metallic impurities in fluoride-base fuels causes highly erratic results when ftrivalent uranium is determined by the methylene-blue method, but such impurities have an even greater adverse effect on the hydrogen-evolution method. Dissolution of the metallic impurities in the hot acid solvent results in quantitative evolution of hydrogen. A more comprehensive description of the theoretical and practical aspects of the methylene-blue method for determining trivalent uranium will be issued in the form of a topical report. Simultaneous Determination of Trivalent Uranium and Total Uranium The inherent advantage of the methylene-blue method over the hydrogen-evolution method, that is, the possibility of determining trivalent uranium and total uranium in the same sample, is still being investigated. The studies have led to the postu- lation of an interaction species of pentavalent ‘uranium and methylene white.* A detailed study 176 of this postulation has been initiated. The method of Vosburgh and Cooper® will be applied to es- tablish the stoichiometry of the complex. Attempts to prepare standard solutions of methylene white through the reduction of methylene biue by zinc amalgam have been unsuccessful because of the ease with which zinc ions react with both the oxidized and reduced forms of methylene blue. An attempt is being made to prepare pure methylene white through electrolytic reduction of methylene blue. DETERMINATION OF LITHIUM IN LiF-BeF, AND LiF-.ZrF4--UF4 A. S. Meyer, Jr. D. L. Manning Analytical Chemistry Division A method for the determination of lithium in LiF-BeF, and LiF-ZrF ,-UF, has been developed that is based on a separation of lithium by an anion-exchange resin, Dowex-1. In this procedure a sulfate solution of the fluorides that is approxi- mately 1 N with respect to H,50, is passed through a column of Dowex-1 in the hydroxide form. In order to avoid the depletion of the resin in the column by the H,50,, the solution is equilibrated with about 20 ml of a water slurry of the resin in the hydroxide form in a beaker before being placed onto the column. Beryilium, uranium, and zirconium are retained on the column as the beryllate, zirconate, and diuranate anion, respectively, while the lithium passes into the effluent as LiOH, The following equations represent the probable reactions, where ROH is the hydroxide form of the anion-exchange resin: H,50, + 2ROH —> R,50, + 2H,0 Zr(SO,), + 8ROH — R,Zr0, + 2R,SO, + 4H,0 BeSO, + 4ROH —> R2Be02v+ R,50, + 2H,0 2U0,80, + 6ROH— R,U,0, + 2R,30, + 3H,0 27277 Li,50, + 2ROH —> R,SO, + 2LiOH It may be seen from the equations that LiOH and water are the only species which are found in the effluent. The lithium is determined by titrating’ the hydroxide with a solution of standard HCI. If the fluoride mixtures contain sodium fluoride in addition to lithium fluoride, the sodium will react in the same manner as the lithium and will 6W. C. Vosburgh and G. R. Cooper, . Am Chem. Soc. 63, 437 (1941). i i B L Do pa? Ao be found in the effluent as NaOH. A subsequent titration of the effluent with standard HCI will give the sum of sodium and lithium as the hydroxides. Lithium is then determined by extracting the LiCl, after titration with HCI, with 2-ethyl-1-hexanol and titrating the chloride ion in the nonaqueous medium according to the procedure developed by White and Goldberg.” If a single alkali metal is present in the fluoride mixture, it can be determined directly by titration of the effluent with standard HCl. When more than one alkali metal is present, the sum of the alkali- metal hydroxide concentration in the effluent is obtained by the method just described. Lithium is then determined by the 2-ethyl- 1-hexanol procedure,7 po'rastm, by the tetraphenyl boron gravimetric method;® and sodium, by the difference. W||||ams‘c:nd Vaughan (cf. *“ANP Service Labo- ratory”’) have utilized a two-stage anion-exchange resin column for the determination of the alkali metal fluorides in the presence of beryllium, zirconium, ‘and uranium fluorides. The first column ¢ontains the anion-exchange resin, Dowex-1, in the citrate form, whereas the second column contains the resin in the hydroxide form. Beryl- lium, zirconium, and uranium remain in the citrate column as the anionic citrate complexes, and the alkali metals pass into the effluent as the alkali metal hydroxides. The alkali metal concentration is then determined by titrating the effluent with standard HCi In the method described above, a single column of anion-exchange resin in the hydroxide form is used, and thus the need for the citrate form of the anlon-exchange ‘resin s’ completely eliminated. © The method has been tested in the ioborafory ‘and .‘found' to be schsfactory in every ‘respect. The ™ T , no of the mefhod is of ’rhe" " order of 1% in the rangémlo to 30 mg “of ]n‘h:um A & "','}"(:'toplcul reporf on the method |s belng wrlt’ren ' ‘preserice of sulfate has been repor’red by Fritz S Johnson.? Excess disodium dihydrogen ethylene- diaminetetraacetate (EDTA) forms o very stable PERIOD ENDING JUNE 10, 1955 complex with zirconium in acidic solutions. The excess EDTA is then back-titrated with bismuth nitrate; thiourea is used as the indicator. The end point is noted by the formation of the yellow bismuth-thiourea complex. "An advantage of this method in comparison with previous methods for using EDTA in the determination of zirconium is that anions which form complexes with zirconium, such as fluoride, sulfate, phosphate, thiocyanate, and tartrate, do not interfere, This volumetric method for the zirconium determi- nation appeared to have several advantages over the gravimetric mandelic acid method now being used, but it was found that the yellow thiourea end point was obscure, especially in the presence of hexavalent uranium, which is also yellow. The precision was consequently poor. The possibility of modifying the method by back- titrating the excess EDTA with a solution of iron{[1l), with disodium- 1, 2-dihydroxybenzene-3, 5-di- sulfonate (Tiron) as the indicator, is now being investigated. This reagent forms an intensely purple complex with iron(ltl) in an acetate-buffered solution having a pH of about 5.0. Two moles of Tiron combine with 1 mole of iron under these conditions. This compiex, which has a pK of about 10, is relatively stable. ! The pK of the iron-EDTA complex’ is about 25, which is approximately the same as that of the zirconium-EDTA complex. It is possible, therefore, to add a solution of iron{lll) to a solution of EDTA and Tiron at a pH of 5,0 so that when the EDTA is completely complexed the color of the solution will change from yellow to purple. Qualitative, pre- Ilmmory tests revealed this change in color at the “end pomf ‘to be well defined. The end point was Cvery shorp when a solution of 0.05 M EDTA " “containing about 50 mg of Tiron was titrated with a “solution of 0.05 M iron(lll}. The end point corre- """ sponded to the correct stoichiometry of the iron- EDTA reachon in which 1 mole of iron cemplexes »_,wflh '| mole of“Er_D‘TA Future 'wcrk w:” lnclude The volume‘rrlc determmahon of zirconium in the - "fi}'{—'i‘.’J. C. White und G. Goldberg, Applz'cation of the Volbard Titration to the 2-Ethyl-1-Hexanol Separation Method for the Determination of Lithium, ORNL-1806 (Nov 4, 1954). sy 8¢. R. Williams, ANP Quar Prog Rep. Mar. 10, 1955, . ORNL 1864, p 162, J S. Fritz and M, Johnson, Volumeinc Determmatwn of Zirconium and EDTA Method Involving Back-Titration with Bismuth, 15C-571 (Feb. 1, 1955). 1OA E. Harvey, Jr., and D. L.. Manning, J. Am. Chem. Soc. 72, 4488 (1950). 177 - ik bopoi e ANP PROJECT PROGRESS REPORT the determination of zirconium by the modified method of back-titrating the excess EDTA and a study of the possible interferences, particularly fluoride and tetravalent uranium. It is expected that tetravalent uranium would interfere because it would complex with the EDTA in the same manner as the zirconium would. DETERMINATION OF OXYGEN IN FLUORIDE FUELS “A. S. Meyer, Jr. J. M. Peele Analytical Chemistry Division Further tests were carried out on the determi- nation of oxygen as oxide in fluoride fuels'! by electrolysis of solutions of the samples in fused KHF,. Quantities of oxygen in excess of 90% of the theoretical value were recovered when known amounts of oxygen were introduced as water by the addition of samples of Na 2CO,4, which react with the fused blfluorlde in accordance with the following equcmon Na,CO, + 2KHF ,— 2NaF + 2KF + H,0 + CO, The yields were not quantitative, because the rate of oxygen evolution decreased as the concentration of water in the electrolyte was reduced. Since traces of hydrogen were deiected in the insoluble gases, even though a high concentration of AgF was added to the electrolyte, methods were adapted for the direct determination of oxygen in the effluent gases. For samples containing large quantities of oxygen the Orsat'? method for the determination of oxygen in gases was applied by passing the insoluble gas, which was collected over KOH, into a solution of alkaline pyrogallate and meas- uring the decrease in volume of the gas. A modification of the Winkler method '3 for the determination of oxygen in water was adapted for samples which contained smaller concentrations of oxygen. The apparatus was modified by con- verting the sweep gas to purified helium and bubbling the effluent cell gases directly into an alkaline solution which contained Kl and a sus- pension of Mn(OH), in which the oxygen was A s, Meyer Jr., and J. M. Peele, ANP Quar. Prog. Rep Mar. 10, 1955, ORNL-1864, p 159. 2w, w, Scott, Standard Methods of Chemical Analysis, 51‘1’1 ed., ll, 2349, Van Nostrand, New York, 1939 13| . W. Winkler, Ber. 21, 2843 (1888). 178 absorbed according to the reaction 4Mn(OH), + O, + 2H,0—> 4Mn(OH), The absorbed oxygen was determined by acidi- fying the absorber solution and thus liberating an equivalent quantity of iodine, which was then titrated with standard Na,5,0, solution. If only microgram quantities of oxygen are present, the iodine is measured by extracting it info orthoxylene and determining the concentration of iodine in the organic phase spectrophotometrically according to the procedure of Silverman, Bradshaw, and Taylor. 14 Experiments are now being carried out in an attempt to reduce the time required for the quantitative evolution of oxygen. No significant increase in the rate of oxygen generation was obtained by carrying out the electrolysis at a temperature of 250°C rather than at 100°C. The effect of the design and current density of the anode on the efficiency of the generation of oxygen is now being studied. It has been found that the use of o mercury cathode, which is introduced by placing a Teflon cup in the electrolytic cell, simplifies the electrolysis, Fluctuations of the current are decreased, and the evolution of fluorine is greatly reduced. By introducing the sweep gas below the surface of the mercury, the problem of slugging of the entrance line is eliminated. Since AgF is incompatible with mercury under these conditions, it cannot be used to reduce the rate of generation of hydrogen. DETERMINATION OF OXYGEN IN 7 METALLIC OXIDES BY BROMINATION J. P. Young G. Goldberg Analytical Chemistry Division Extension of the method of Codell and Norwitz '3 for the determination of oxygen in titanium to the determination in fluoride-base fuels was continued during this period. Several modifications of the apparatus described previously '® were made. New traps of smaller dimensions were designed, and, in addition, an ice-salt trap was placed between the ignition tube and the first dry-ice~alcohol trap to 4[... Silverman ond W. Bradshaw, Determination of Oxygen in Certain Gases, NAA-SR-892 (April 15, 1954). M. Codell and G. Norwitz, Chem. Eng. News 32, 4565 (1954). 6;. c. White, G. Goidberg, and J. P. Young, ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 161. R’ i G, A i DY - - oxygen con’rammchon were found. =+ In ‘several attempt Bo not Alel ‘remove more completely the excess bromine vapor. [+ was found that if the flow rate of the mixture of helium and bromine vapor were too rapid, the second dry-ice—alcohol trap became plugged with solid bromine during the course of a determination. In the previous application of the bromination method to the de’rermlnohon of oxygen in BeO, considerable oxide confcmlncmon was found in the NaF FeF flux that was used. A small quantity of the lmpure flux was therefore treated with bromine vapor at 750°C, and then the prefreated flux was used in the determination. Essentially complete recovery of the oxygen in the BeO sample was obtained =when this mixture was treated with bromine vapor at 950°C. Several samples of CrF, were analyzed for oxide contommahon, and reasoncble values were found when the samples were treated with bromine vapor at 950°C The accuracy and optimum conditions for this determ!noflon will be investigated with known mixtures of CrF3 and Cr,0O Also, several samples of Nazsz were analyzed for oxide contamingtion. The bromination was performed at 950°C, _and reasonable values of During the analysis of two scmp]es of Li,ZrF,, a precipitate formed on the surface of the Bc:(OH)2 bubbler and caused inefficient flow of gas through the appa- ratus. It is believed that the occurrence of this precipitate was due to a reaction mvoivmg the hydrogen fluoride or fluorine which was present or was formed in the samples being analyzed. A trap of NoF and KBr is belng pre pared for removal of these gases. The accuracy ‘and op'rlmum condl’rlons for the defermmaflon of oxygen contamma'rlon in - ZrF W||| fhen be mveshgated ;wcs found that the occurrence of various side reactions made the method impractical. s aoply-" fhe flrst reaction as a method for determmlng omde present as BeO, it PERIOD ENDING JUNE 10, 1955 It was impossible to secure reasonable blanks in the absence of BeO, and furthermore, when BeQO was present, no evidence of stoichiometry wds found. In these attempts, samples of BeO and BeF,-BeO were placed in contact with solutions of KF, as concentrated as 20% (W/V), at tempera- tures of 80°C. The concentration of hydroxyl ions presumed to have been formed during the dissolution of the sample was then determined by titration with standard acid. Values from 7.5 to 9.0 pH units were- chosen as arbitrary end points for these titrations. Dissolution of either BeO or BeF, (with oxide contamination) was more rapid in higher concentrations of KF and at pH values of 7.5, A fine crystalline precipitate was formed during the dissolution of all samples containing BeF,. From the results of x-ray diffraction analysis, it was found that this precipitate contained K,BeF,. DETERMINATION OF TRACE AMOUNTS OF NICKEL IN FLUORIDE FUELS WITH SODIUM DIETHYLDITHIOCARBAMATE J. P. Young M. A. Marler Analytical Chemistry Division The use of sodium diethyldithiocarbamate in the colorimetric determination of nickel in trace amounts was investigated. The work of Chilton'’ served as a basis for these studies. Sodium diethyldithiocarbamate is a very sensitive colorimetric reagent for nickel; however, this re- agent forms colored complexes with a wide variety of cations. Many of the metal complexes of sodium diethyldithiocarbamate are extractible into organic solvents. The molar absorbancy index for nickel-diethyl- dithiocarbamate extracted into carbon tetrachloride or" 1,2-dichlorobenzene from an aqueous solution “~ whose pH was 9 was found to be about 34,000 at ‘7 a wavelength of 328 mu. The molar absorbancy -. index of nickel dimethylglyox'ime is about 11,000 .. at the wavelength of its maximum absorbancy. '.'/_L(:hllton'|8 subsequently reported that a more ef- _:W_flment exfrcctlon of the mckel cpmplex occurred | i the pH of the ‘aqueous phase were first basic .+: to bromeresol green (pH ~5). Without the addition _of the complexing agents described by Chilton, it “Was found that uranium interfered in fhis method, e STl 7). M. Chilton, Anal. Chem. 25, 1274 (1953), 18 M. Chilton, Anal. Chem. 26, 940 (1954), 179 ANP PROJECT PROGRESS REPORT as did iron and molybdenum to a very slight degree. -With the addition of the complexing agents de- scribed by Chilton, it was found that the inter- “ference of uranium was still present, although “somewhat decreased. Of the other ions present in fluoride-base fuels, chromium and the alkali ions presented no interference, and evidence indicates that zirconium does not interfere. Absorbancy “spectra of carbon tetrachloride extracts were —~determined for aqueous solutions containing sodium ~ diethyldithiocarbamate and all the cations mentioned -in this discussion. ANP SERVICE LABORATORY. W. F. Vaughan C. R. Williams Analytical Chemistry Division The number of determinations of the oxygen content of metallic sodium increased sharply during ‘the quarter. Efforts are being made to increase the accuracy of this determination, and two steps have been taken to achieve this goal. More ef- ficient removal of the thin film of adsorbed moisture on the outer surface of the glass bulbs used for the determination is being achieved by immersing the bulbs in acetone and then quickly rinsing them in dry ether before placing them in the hexane-butyl bromide mixture. Also, the reagents, both hexane and butyl bromide, are being desiccated over phosphorus pentoxide so that the water content is less than 5 ppm. The major portion of the work in the service laboratory continued to be the analysis of fluoride salts, with the emphasis being on the determination 180 of the following components: Na, Zr, Li, K, F, U3, total U, Ni, Cr, Fe, and Mo. A new procedure for the determination of lithium was proposed and tested in which lithium and other alkali metals are first separated from zirconium, beryllium, uranivm, and sulfate ions by means of two anion- exchange resin columns in series. The first column is prepared in the citrate form and the second in the hydroxide form. The effluent from the second column is a solution of lithium or alkali metal hydroxides, the concentration of which is deter- mined by titration with standard acid. Work by Manning (cf. ‘“'Determination of Lithium in LiF- BeF, and LiF-ZrF4-UF4”) has subsequently shown that only the hydroxide form of the anion resin is required for the separation. ' A total of 1673 samples was analyzed, on which 8678 determinations were made. The backlog consists of 151 samples. A breakdown of the work is given in Table 9.1, TABLE 9.1. SUMMARY OF SERVICE ANALYSES REPORTED ' Number of Number of Samples Determinations Reactor Chemistry 1139 6143 Experimental Engineering 525 2515 Miscellaneous 9 20 Total 1673 8678 i ¥ ! Y ‘which — PERIOD ENDING JUNE 10, 1955 10. RECOVERY AND REPROCESSING OF REACTOR FUEL F R Bruce D. E. Ferguson W. K. Eister H. E. Goeller M. R. Bennett J. T. Long G. I, Cathers R. P. Milford S. H. Stainker Chemical Technology Division E. E. Hoffman C. F. Leitten, Jr. Metallurgy Division PILOT PLANT DESIGN The design of the pilot plant for recovering ANP fuel by a fused salt—fluoride volatility process is expected to be completed by August 15. The engineering flowsheet is 85% complete. Of the 29 pieces of process equipment now contemplated, 10 are completely designed or specified and on order; 2 of the 10 have been received. The scheduled construction completion date is still December 31, 1955, The plant will be located in cells 1 and 2 of Building 3019. Equipment for highly radioactive materials — the fluorinator, the ARE fuel dump tank, and the vessel for melting other salts to be processed (for example, salts from the in-pile loops) — will be in cell 1. Equipment for less radicactive materials will be in cell 2. The remainder of the equipment will be in the operating gallery and on the roof above cells 1 and 2, The present plan of operation is to lower the ARE dump tank into the cell, melt the contents, and pass the molten salt info the fluorinator by means of'm"f'rogen pressure. |he process differs from thc’r prevnously descnbed] in the following ~ details: “only two cold traps are provided, one at -40°C qnc! one at ~62°C; fhe UF will be w"‘recycled “together “with fluorine gas, through the - absorber ‘and cold traps to effect additional decon- -~ tamination nc necessary to meef product spec:fl-' | "'.':":“"ca'r:ons. o s . The bcrren molfen sqlf will be removed from the i fluorlnm‘or with pressure and pu'r mfo metal cans _ezzhburled The sodium fluoride ab- " sorber 'con’ral'nhx'ng the volqtlle Ffission- product floe- rides will be transported to the burial ground, - where it will be dumped and the contents buried. Excess fluorine in the off-gas will be reacted with 'p, E. Ferguson et al., ANP Quar. Prog. Rep. Mar. 5 to 10% aqueous potassium hydroxide in a spray tower, ' Two refrigeration units, each consisting of a« Freon F-22 and F-13 cascade system, will be used to chill Freon F-11 to —40 and —-62°C for recircu- lation through the cold traps. PROCESS DEVELOPMENT In further studies on an absorption bed for removing volatile radicactive material from the Ui:é-l:2 gas stream of the fused salt—volatility process, sodium fluoride was found to be much more effective than calcium fluoride, Decon- tamination was poor in second runs with the same absorbent bed. |t is believed that the femperature of the bed is important, but control of the bed temperature was difficult, The absorbent beds were 1-in.-dia tubes that contained 90 g of either calcium fluoride or sodium fluoride in a 9-in. length or 180 g in an 18-in. length, The CaF_ used was made by fluorinating CaSO,. Both the CaF, and the NaF were 12 to 40 mesh. The gas from a fluorinator was passed directly into the absorbent bed. The fluorination ‘reaction was carried out with 365-g charges of the ARE-type fuel NaF-ZrF, -UF, mole %) containing 30 g of uranlum. (53.5-43.0-3.5 Previous “work? had been done with similar material on a sccle of a 67-g charge of fuel "The NaF was ten times as effective as reported ‘previously., The amount of fluorine used for the fluorination ‘was much less than the ninefold "_:excess used prewously. In some of the new runs “the flucrine was passed info the molten fuel very rapidly during the induction’period and then slowly during the volatilization of the UF to make the residence time of the off-gas in the a%sorbe-r long. 2D, E, Ferguson et al., ANP Quar. Prog. Kep. Dec. 181 ANP PROJECT PROGRESS REPORT Calcium fluoride at either 200 or 650°C was about one-third as effective as NaF in removing volatile ruthenium and niobium fluoride from the UF(S-F2 gas stream. Tests were made with both 9- and 18-in.-long beds (Tables 10.1, 10.2, and 10.3). ' Good decontamination was obtained with NaF in either a 9- or an 18-in.-long bed at 650°C when the NaF /U weight ratio was 3/1 (Table 10.1, runs 1, 2, 3) or 6/1 (Table 10.2, runs 5 and 7). When the 9-in.-long bed was re-used (run 4, Table 10.1), so that the over-all NaF /U weight ratio was 1.5/1 for the two runs, decontamination factors for gross beta, gross gamma, and ruthenium beta decreased sharply. With the 18-in.-long bed the same effect was observed, although here the over-all NaF/U weight ratio for the two runs was 3/1 (Table 10.2, runs 6 and 8). The temperature of the NaF absorbent bed was difficult to maintain at 650°C. The temperature profile over the 18-in.-long bed varied 90°C. The data in Table 10.2 were obtained with the hottest point at 670°C; this temperature may be a little high for efficient operation, but operation at a [ower temperature would probably result in too much uranium retention in the cooler section, CORROSION STUDIES In 20 laboratory-scale fluorination runs at 650°C, corrosion of nickel test coupons and of the nickel reaction vessel was fairly low. Since conditions changed continually during the runs and since the various components of the vessel were attacked to different degrees, a calculated over-all cor- rosion rate would have no significance, However, it appears that a large number of ARE fuel fluori- nation runs can be made in one reaction vessel before corrosion interferes with the process. TABLE 10.1. DECONTAMINATION IN A 9-in.-LONG NaF ABSORBENT BED UF6-F2 gas stream from fluorination of ARE-type fuel at 600 to 650°C passed through 1-in.-dia bed with temperature of 650°C in hoflesf_portion; same bed used in runs 3 and 4; F"2 flow rate about 300 ml/min initially, then aobout 150 ml/min during remainder of run F2/U mole ratio: run 1 3.7 run 3 8.2 run 2 3.6 run 4 4.8 NaF/U weight ratio in absorber: 3/1 for runs 1, 2, and 3; over-all ratio for runs 3 and 4 = 1.5/1 Decontamination Factors Activity Run 1 Run 2 Run 3 Run 4 Over-all Gross 3 1.3 x 104 5800 1.0 x 10* 2900 Gross y 3.2 x 104 2.1 x 104 2.4 x 104 4500 Ry y 1700 1600 1.0 x 104 200 Zr-Nb y 3.2 x 105 7.4 x 104 7.0 x 10% L0 x 10° TRE* 5.2 x 105 5.0 x 104 4.9 x 104 1.1 x 107 Across Absorbent** Gross 3 340 200 ' 35 Gross y 930 4100 220 Ruy 940 1400 62 ZeNb y 900 1.2 x 104 2000 TRE B 8 17 8 *Total rare earths. **Caleulated on basis of activity found in absorbent and final product. 182 »” ) w PERIOD ENDING JUNE 10, 19535 TABLE 10.2. DECONTAMINATION IN AN 18-in.-LONG NaF ABSORBENT BED UF F gas stream from fluorination of ARE-type fuel at 600°C passed through 1-in.-dia bed with hoflest point at 670°C; same bed used in runs 5 and 6 and in runs 7 cnd 8; F flow rate cbou'r 200 mi/min 2/U mole ratio: run 5 8.1 run 7 8.4 run 6 9.9 run 8 10.2 NaF /U weight ratio in absorber: 6/1 for runs 5 and 7; over-all ratio for runs 5 and 6 and 6 and 7 = 3N Decontamination Factors Activity : Run § ’ Run 6 Run 7 Run 8 Over-all Gross 3 3900 1600 4300 2400 ; Gross y 9700 2700 2.0 x 104 4000 5 Ru y 1500 140 6700 250 ZeNby 3.5 x 104 2.9 x 104 9.8 x 10% 5.2 x 10% TRE 3 3.7 x 104 1.0 x 10° Across Absorbent* Gross 3 2 14 Gross y 40 | 82 Ru y 5 31 Zr-Nb y 250 700 *Calculated on basis of activity found in absorbent and final product, - _’;‘Q",'Nl, | 0 01% 7C"),‘ was also cut |ong:fudma|ly and The A’ nickel reaction vessel was 2 in. in diameter, The three test coupons were mounted in an upright position at the bottom of the reaction vessel, as shown in Fig. 10.1, in such a way that one-third the surface area of ‘each coupon extended from the quUICI mfo fhe gas phase. The coupons' were 3 in, long, % in. wide, and ¥ in, thick. Two of ’rhe cdupons were”A"n:ckel?nommal compo- | 'smon 99.4% Ni, 0, 05% C), and one of them wcs' ‘ “cut longn‘udlnqily and welded The thlrd coupon, WhIC]’I wa _ mckel (nomlncl composmon 99.4% to pro‘é'évss condlhons for a 'rotal of 30 hr. The:w ' fluorme flow rcte varled from 50 to 300 ml/mm“ “and was regulated so that 9.4 moles of fluorine was used per mole of uranium in each run, 7.5 mils in depth. Corrosion of the welded coupons {both ‘'A’" and “I."" nickel) was greater than that of unwelded ones, but in both cases the corrosion was of the solution type (Fig. 10.2), and there was fairly uniform surface removal, Dimensional and weight-change analyses also showed that corrosion was greater in welded than in unwelded coupons (Table 10.4). Thé most severe attack was on the outer surface of the fluorine gas inlet tube in the vapor zone (Fig. ]0 3). The attack on this tube in the liquid zone was more uniform and varied from 4.0 to region in contact with molten salt showed that the attack there was of a solution nature (Fig. 10.4). Recovery of uranium was high in all runs (Table 10.5). The uranium loss in the waste salt was consistently lowest in the 50-min runs at the 183 Dimensional analysis of the __‘;_,E,;;:',reachqn vessel indicated nonuniform attack of "5 to 9 mils in both the liquid and gas zones. A NaF - 'C'metallogrcph:c examination of the vessel in the" .r.,_foi' an md:vrdual run varled from 4, 58 to 0.83 hr, - the reaction’ vessel cmd fhe coupons bemg exposed f n L T A A ek sk . st i o i o o e - ~ANP PROJECT PROGRESS REPORT TABLE 10.3. DECONTAMINATION IN A CaF , ABSORBENT BED U\F6:-F-2 gas stream from fluorination of ARE-type fuel at 600°C passed Through. Ir-rir‘l‘.-diu bed; __F2 flow rate about 200 ml/min Run 9: 9_iin7-|ong bed at 650°C, F2/U mele ratio 6.5 ‘Run 10: ]8-in'.-long bed at 2000(:, F2/U mole ratio 7.5 Run 11: 18-in.-long bed at 200°C, F2/U mole ratio 8.7 Decontamination Factors Activity o Run 9 Run 10 Run 11 Over-u” | Gross 3 1900 960 1700 Gross y 1800 1200 3600 Ru y 130 | 80 200 CZrNby 5300 6400 | 47 x 104 " TREB | 2.3 x 10° Across Absorbent* Gross B ‘ 4 3 5 Gr'oss y 18 20 65 Ru y 5 3 8 ZrNby 40 110 1000 TRE 3 13 *Calculated on basis of activity found in absorbent and final product. TABLE 10.4. WEIGHT LOSS OF NICKEL CORROSION COUPONS TESTED IN LABORATORY-SCALE FLUORINATION RUNS . Original Weight Final Weight Weight Change Type of Coupon T (9) (g) (g) (%) Welded *“L** nickel 83.9878 ~ 80.3760 3.6118 4.3 o Welded "*A’* nickel 86.3445 82.7515 3.5930 42 Unwelded **A’* nickel 82.6071 80.2160 2.3911 2.9 o ¥ Do not lelels highest fluorine flow rate. This result was pos- sibly due to a smaller loss of fluorine in corrosion in the short runs than in the long runs. Out of 3935 g of salt, 341 g of uranium was actually recovered as UFé' which corresponds to an initial PERIOD ENDING JUNE 10, 1955 uranium content of 8.66%. Analyses of this par- ticular batch of fuel indicated a urenium concen- tration ranging from 8.30 to 8.76%. Even if the higher value is assumed to be correct, the total recovery was 99.0%. TABLE 10.5. URANIUM LOSSES IN LABORATORY-SCALE FLUORINATION RUNS Uranium Loss Duration Fluotine Flow Rate Number of Runs (hr) (ml/min) in Waste " (% of total) 1 4.58 55 0.11 4 2.50 100 0.02 to 0.16 5 1.35 200 ' 0.06 to 0.23 9 0.83 - 300 ' 0.01 to 0.04 e " S 7 et Fig. 10.1. Cross Section of the fluorine gas inlet tube may be seen at point A. Nickel Reaction Vessel, The pitting type of aitack on UNCLASSIFIED Y- 14847 185 Sl B b, v Sl ¢ Lo o Aelele ANP PROJECT PROGRESS REPORT UNCLASSIFIED ¥-15140 Fig. 10.2. Cross Section of Welded “‘L’* Nickel Test Coupon Exposed to Molten Sult |h a Nickel Reaction Vessel. Note uniformity of attack. Etched with KCN plus (NH )25208' 12X. - AS RECEIVED Fig. 10.3. Outer Surface Attack of Fluorine Gas Inlet Tube in Vapor Zone of Reaction Vessel. S'eéfioh taken at point A of Fig. 10.1. Etched with KCN plus (NH ) 186 ,Og 20X. » e 5ot ¥ n PERIOD ENDING JUNE 10, 1955 0.004 0.002 0.003 0.004 0.C05 0.006 0.007 0.C08 0.009 0.040 0.044 INCH Fig. 16940 Inner Surface of Specimen of ‘‘A’" Nickel Reaction Vessel Taken from Region Exposed to ARE-.Type Fuel. Note nonuniform surface attack. Etched with KCN plus (NH4)25208' 250X. 187 T T ek -rern - s n v T T R N T A i G et o i LA *® Part i SHIELDING RESEARCH Atk K sk o RS -t " g 11. SHIELDING ANALYSIS E. P, Blizard F. H. Murray C. D. Zerby ' ~ R. B, Murray Applied Nuclear Physics Division S. Auslender J. B. Dee C. A, Goetz J. Smolen Pratt & Whitney Aircraft A calculational method has been developed for tracing all the gamma radiation born within a circulating-fuel reactor; for this calculation it is assumed that the reactor is spherically symmetric and that all regions are homogeneous. in addition, the codings of two Monte Carlo problems - the calculation of the heat generation resulting from the absorption of gamma radiation in laminated shields and the calculation of the energy and angular distribution of air-scattered neutrons from a monoenergetic source — are nearing completion, The analysis of the constant-velocity transport equation has also been extended. GAMMA-RAY DISTRIBUTION IN A __ CIRCULATING-FUEL REACTOR AND SHIELD C. D. Zerby M. D. Pearson! A semianalytical Monte Carlo calculation has been initiated to determine the history of all gamma radiation born within a circulating-fuel reactor of the ART type. The calculation is a joint effort of the Boeing Airplane Company and ORNL and will be coded for the IBM 701 automatic digital cdmpm‘gr. The problems of interest will be compufed at Boeihg. The calculation w1|l be ' smphhed by ’rhe assumphons thc’r the recctor has spherlcolly symmetrlc geomefry and " that the = regions w:fhout homogene:fy (for example, the - ,heat exchanger) ure homogeneous. However, these ~ will be the only snmphflcahons, and all shells of lr'TheAcore, refiecfor, and “shield, |nc|ud|ng the _ Inconel rlconfamer she_lls, w:” be ’raken ' m’ro' _account,” T T e : The problem' is deS|gned to de'rermme the hlsfory of “all gamma radiation born within cmy one ‘spherncol shell of the confugurc:tlon and ’ro account_ ' . for the rodlal disfrlbu’non of the source., The‘:_ - results w:” mc]ude u "determination of the energy @ ————— ]Boeing Airplane Company. . experimental data, absorption resulting from the transport of the gamma radiation. These data will be presented as a function of radius. The results will also include determinations of the energy spectrum and angular distribution of the gamma radiation penetrating the shield., The data obtained will be normalized per watt of power generated in the core. The sources of gamma radiation within the core will be determined from neutron flux distribution data obtained from one of the existing multigroup neutron diffusion calculations. The power of the reactor will be determined from these same data. The energy and density normalization of the total prompt-fission .and fission-product decay gamma rays born in the active fuel region of the core will be obtained from the relationship (1) N(E) = 13,7 e~1:08E | where N(E) = total photons/fission/Mev, E = energy (Mev). The components of this total are given by the following equations: (2) Ny(E) = 7.7 e=1-02F | @) Ny = 6en12E where - N,(E) = prompt photons/fission/Mev, ~ N,(E) = decay photons/fission/Mev. Equation 2 is an empirical fit to published datq, while Eq. 3 is a best guess inferred from some 3 Equation 1 is the sum of Eqs. 2 and 3. Better data will be used as they become available. In the heat exchanger region, only the fission- i 2 .. product decay gamma rays will be considered, 2, L. Gamble, Pbys. Semiann. Prog. Rep. Sept. 10, 1953, ORNL-1620, p 15, 3R, W. Peelle, private communication. 191 ik opes S A - data on heating are obtained in a similar manner. ANP PROJECT PROGRESS REPORT since the thermal-neufron flux, and thus the - fissioning, is depressed by the boron curtains. It will be assumed that these decay gamma rays - will be uniformly distributed in the heat exchanger. The Vener'gy and density normalization of this ~radiation will be obtained from Eq. 3. In all other regions of the core the energy and density ~of the capture gamma rays will be determined from published data.4 The capture gamma-ray sources in the shield “will be mcluded when data on the neutron flux in this region become available. It is hoped that the numerical integration of the transport equation for spherical shields will be completed so that the results will be available to complete this problem. A neglected source of radiation is that resulting from inelastic neutron scattering. However, this radiation can be included without modification of the code when the information becomes available, It is intended that each region of the assembly be treated separately, and the results will be added to complete the treatment of the reactor as a whole, By treating each region separately, it wullrbe ‘possible to make parameter studies to determine optimum configurations to shield against radiation from any patticular region. For this calculation it was necessary to deviate from the straightforward Monte Carlo methods because of the many mean free paths of material to be penetrated by the core-region gamma rays; for example, there are approximately 16 mean free paths from the center of the core to the outside of the shield for 3-Mev gamma rays in a typical 300-Mw design. Many mean free paths of at- tenuation for radiation usually result in poor statistics in the solution when the problem is ~ treated by Monte Carlo methods. It was therefore necessary to use a semianalytical method, which is a considerable improvement over the straight- forward procedure. The procedure is to generate the spatial energy and angular collision density of gamma rays born in any one region by standard ~ Monte Carlo techniques and from this to calculate analytically the energy spectrum and angular distribution® of the penetrating radiation. The ' 6 In practice the analytical and Monte Carlo parts of the problem are carried on simultaneously, each modifying the other. 4P. Mittelman, Gamma Rays Resulting from Thermal .‘ Neutron Capture, NDA 10-99 (Oct. 6, 1953). 192 ENERGY ABSORPTION RESULTING FROM GAMMA RADIATION INCIDENT ON A MULTIREGION SHIELD WITH SLAB GEOMETRY C. D. Zerby S. Auslender The coding of a Monte Carlo calculation of heat generation resulting from the fransport gamma radiation through shields with ' laminated-slab geometry is nearing completion,” The data re- sulting from the calculation will include the energy absorption as a function of depth, as well as the energy reflected and penetrating the slab, A revised code makes it possible to consider the gamma-ray heating in a circulating-fuel reactor when it is idealized to slab geometry (see pre- ceding paper). As coded, 1000 histories can be calculated by the Oracle with standard Monte Carlo methods in approximately 5 min, This number of histories should be sufficient to provide good statistics for gamma rays incident on a shield with a thickness of approximately seven mean free paths, The code should be applicable to other programs as cases of interest arise, ENERGY AND ANGULAR DISTRIBUTION OF AIR-SCATTERED NEUTRONS FROM A MONOENERGETIC SOURCE C. D. Zerby A general description of a Monte Carlo calcu- lation of the energy and angular distribution of air-scattered neutrons from a monoenergetic source was given in a previous report.®. Coding of the problem for the Oracle is continuing, and slight modifications in the original problem have been made for greater utility, The program now makes possible a determination of the energy and angular distribution of air-scattered neutrons from a unit surface source on the surface of a sphere, SFor an example of the type of data to be obtained see C. D, Zerby, ANP Quar. Prog. Rep. Dec. 10, 1954, ORNL-1816, p 144 C. D. Zerby, *‘Energy Absorption Resulting from Incident Gamma Radiation as a Funchon of Thickness of Materials with Slab Geometry,"’ Reactor Shielding Information Meeting, Nov. 15-16, 1954, General Electric Co., Aircraft Nuclear Propulsion Dept., Cincin- natz, Ohio, WASH-185 (Part I), p 26 (Feb, 1955). ®Eor an exomple of the type of data to be obtained see C. D. Zerby, ‘*The Penetration of Composite Slabs by Slanf Incident Gamma Radiation (Monte Carlo So- luhon), WASH-185, op. cit., p 15. 7c.p. Zerby and S, Auslender, ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 173, 8¢. p. Zerby, ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL-1864, p 173, £l i, Rk, i idiai A i o ey e e MR i R e P The surface source can have an angular distri- bution about the normal given by P(f) = Acos™ 8, where P(6) probability of emission per unit solid angle in a direction inclined at an angle 6 with the normal, f = angle with respect to the normal, A = normalizing constant, n o= 0,1,2,0. . The sphere on which the source is located will be treated as a black body to radiation returning to the sphere. With these modifications it is possible to calcu- late the answers for several sources having different powers of the cosine distribution and then to combine them to duplicate the distribution from the surface of a circulating-fuel reactor, Detailed information about the neutron radiation reaching the crew compartment shield can thus be obtained. A study of the effect of neutrons originally emitted in the half space to the rear of the aircraft on the total flux at the crew compartment shield can also be made with these modifications, Economical shaping of the reactor shield is dependent on a thorough knowledge of behavior of this radiation. 9 Murray, ANP Quar. Prog. Rep. Mar. 10, 1955, ORNL 1864, p 173. PERIOD ENDING JUNE 10, 1955 ANALYSIS OF THE CONSTANT-VELOCITY TRANSPORT EQUATION F. H. Murray The analysis of the constant-velocity transport equation previously developed9'm has been extended.!! The inversion of the Fourier trans- forms computed in the first paper may be avoided by the use of eigenfunctions for all the separate strata or shells of a stratified medium. The course of the calculations is the same os for a system of nonhomogeneous differential equations. To any solution of the nonhomogeneous equation is added the sum of all eigenfunctions, each multiplied by a constant factor, and these factors are determined by the conditions at the boundaries between different media. In this case the con- ditions are that the product (n,v)f(S,) be con- tinvous at each surface and that the flux vanish at infinity for finite sources. The eigenfunctions for cylindrical and spherical source distributions may be calculated from those for plane source distributions with the aid of matrix representations of rotations in three-dimensional space. The solution of the transport equation with finite sources but for unbounded space represents the solution of the nonhomogeneous equation, mF. H. Murray, Anisotropic Scattering of Neutrons in a Uniform Medium with Beam Sources, ORNL CF- 54-11-83 (Dec. 3, 1954). Ne w Murray, Analysis of the Constant Velocity Transport Equation with the Aid of Eigenfunctions of the Various Medza:, ORNL CF-55-5-2 (to be published). 193 ANP PROJECT PROGRESS REPORT 12, LID TANK SHIELDING FACILITY G. T. Chapman J. M. Miller Applied Nuclear Physics Division W. J. McCool H. C. Woodsum Pratt & Whitney Aircraft ~ The second series of experiments with mockups - of 'fl-jé”<:_"'_i_fculbti'r'1§->fu‘e|' reflector-moderated reactor ' ___(RMR‘)‘ @:nd shield have been continued at the Lid Tank Shielding Facility (LTSF). This series con- ‘sists of two sets of fests: the static source tests “and the dynamic source tests. The measurements fcken to date have been concerned w:Th the static source fests REFLECTOR MODERATED REACTOR AND ' ' SHIELD MOCKUP TESTS G T Chcpman J. B. Dee : H. C Woodsum Mcmy of ‘rhe measurements in the static source ‘tests of the reflector-moderated reactor and shield mockup experiment at the LTSF have been com- pleted, As reported previous!y,] these tests were designed to determine the primary and secondary sources of radiation that reach the outside of the shield. The tests completed to date have included (1) gamma-ray dose rate measurements behind various configurations of the mockup (the standard configuration is shown in Fig. 12.1), {2) thermal- neutron flux distribution measurements within the beryllium reflector, (3) neutron measurements be- hind the mockups as a function of the thickness of the lead gamma-ray shield region, and (4) sodium activation measurements within the heat exchanger. Before any mockup tests were performed, a series of gamma-ray dose rate, fast-neutron dose rate, and thermal-neutron flux measurements were made both in plain water and in borated water contained in the iron tank that holds all the components of the mockup in the experiment. These measurements were made in order to determine the effect of the borated water in which all the mockup measure- mén"fé' 6re made. The mockup fcnk which is placed agclnst the source nlate, has a / in.-thick alumi- num window onthe source side. Rodm’rlon measure- ‘ménts were also made in the plain water of the LTSF (that is, no mockup tank used). 1J .B. Dee, ANP Quar. Prog. Rep. Mar. 10, 1955, : ORNL 1864, p 176 194 Gamma-Ray Dose Rate Meos-urem’e‘nfs The gamma-ray dose rate measurements in plain water and in borated water (Fig. 12.2) were used to determine the source strength and approximate energy of the gamma rays from the LTSF source plate. In order to separate the sources, three simultaneous equations with three unknowns were solved: - +T. . & D')’(HZO,LT) H Capture D'y(Hzo,RMRT) =1 + FH Capture + FAI Capture ! D')’(B-H20,RMRT) =T, r H Capture Al Capture + + , 66 10 Sremr 2-0$-057-66-117 3pin, Al N e R \\\\\\\ ~25em AR Ygin. INCONEL DRy e 4-in. DRY —- . o el TANK WALL 1 (Fe) Ya-in Al \f— - rel TANK WALLS <] fl\\ / {/ ALl - T TR D _ ,,‘ N WOODY 5 i1 BoRaL T i \ T (4 SHEETS) 2in BORAL | WOOD|i720% L HEAT EXCHANGER (2 TANKS] - - (4 SHEETS) ~—{ | || L | - 1lin. NICKEL 4% in. LEAD (3 SLABS) e — = BORATED WATER (195wt % BY _ Fig. 12.1. Sfandor'd‘ Confijfiraiion- (Top View) for LTSF RMR.Shield Tests. e e " the mockus of the RMR and shisld was in the LTSF, and the effects of varying dlfferenf“' ' reglons were studied. Measurements behind various T weEenE T 2-01—057~-66-~11 2 —Fa PLAIN WATER, RMR TANK o PLAIN WATER, LID TANK 5 BORATED WATER, RMR TANK GAMMA-RAY DOSE RATE {mr/hr) 10 50 60 7O 80 920 {00 {10 120 {30 140 {50 {60 z, DISTANGE FROM SOURGE PLATE {cm) ' Fig. 12.2. Gdfim&fi&ybofié Rate Measurements in Plain and Borated Water at the LTSF. where 'D,), = total gamma-ray dose rate meas- ured, " = gamma-ray dose rate from LTSF - source plate, H Capture = gamma-ray dose rate from hydrogen . capture of thermal neutrons, I, Capture = 9amma-ray dose rate from alumi- num capture of thermal neutrons, ratio of the integral of the thermal- neutron flux in pure ond boro’red woter, 66 il “and bérated woter :ro’no of the fhermol neufron fluxes'__'_‘ . the alumlnum wuncfow In pure lun_form Ieadr thlck” t F “the curves is due to”rhe 0 5 Mev‘gammo ray re- PERIOD ENDING JUNE 10, 1955 PAECHET. 2-01-057-66-116 WATER CONTAINED IN TANK WITH M WINDOW TOTAL GAMMA PLAIN WATER TOTAL GAMMA DOSE, BORATED WATER DOSE FROM SOURCE PLATE GAMMA DOSE FROM HYDROGEN CAPTURE, PLAIN WATER GAMMA-RAY DOSE RATE {mr/hr) 10 50 60 70 80 90 100 HO 120 130 140 150 160 {v0 180 z, DISTANCE FROM SOURGE PLATE (cm) Fig. 12.3. Contribution of Various Sources to Gamma-Ray Dose Rate in Plain or Borated Water Against LTSF Source Plate. beryllium and the heat exchanger showed that, for thicknesses of boral greater than / in., there was no change in the resulting gamma-ray dose rate (Fig. 12.4). This simply means that gamma rays resulting from epithermal neutron capture in the heat exchanger are not an important contribution to the total gamma-ray dose rate (see further discussion on the sodium activation below). Figure 12.5 shows a decrease in total gamma-ray dose rate with an increase in heat exchanger thickness. This decrease was investigated and was found to be just an attenuation of the primary gamma-ray dose. A lead thickness variation study was carried out “ not only to determine the resulting effect on the - gamma-ray dose rate but also to learn something econdcry gamma-ray produchon in and Results of ’rhls sfudy (Fag 12.6) sulhng “from fhermal-nelfron capture by the boron estobhshgd\ fhlcknesses of the boron curtain between the 'qf the borofed wa’rer |mmedLate|y behmd ‘rhe iecd |nsfa|1édfw\ gamma wrcy n r fl'le back of ‘the tonk Ana!yses of these curves md:cote ‘to dc're, “that most of this secondary dose is caused by boron capture gamma rays, hydrogen capture gamma rays, and thermal 195 T+ R Py ANP PROJECT PROGRESS REPORT SEL R -66-112 2™ e CONFIGURATION 2A, 13 in. BORAL 4 CONFIGURATION 2C, Y% in. BORAL - 107 "0 GONFIGURATION 2B, NO BORAL GAMMA-RAY DOSE RATE {mr/hr) 50 60 70 80 90 {00 #HO {20 {30 440 {15C 180 z, DISTANGE FROM SOURGE PLATE {cm) " Fig. 12.4. Gammo-Ray Dose Rate Beyond RMR- * Shield Mockups: Effect of Boral Thickness. PECRET 2-01-057-66-110 STANDARD CONFIGURATION (4-in. HEAT EXCHANGER) v n STANDARD CONFIGURATION PLUS 2-in, HEAT EXCHANGER GAMMA-RAY DOSE RATE (mr/hr) 3 o 10—2 o 50 60 7O 80 90 00 O {20 130 40 50 160 : z, DISTANCE FROM SOURCE PLATE (em) Fig. 12.5, 'lGhr-fimui-'R;:y Dose Rate Beyond RMR- Shield Mockups: Effect of Heat Exchanger Thick- ness. and epithermal neutron captures in the lead. For large lfead and water thicknesses (for example, .9 in. of lead and 80 cm of water), the dose rate _ from epithermal neutron capture in lead seems to "f:‘[S:"_%:"dominate'.' More experiments are planned for ",inyes'f'igating these results. The placement of the "l_éd'd will also be investigated in a later test to 196 o [ o N N N . OO, GAMMA-RAY DOSE RATE {mr/hr) 10 5 2 7Y, in. ~2 0 50 60 7C 80 90 100 #O {20 430 140 150 160 180 2, DISTANGE FROM SOURGE PLATE {cm} Fig. 12.6. Gamma-Ray Dose Rate BeyondRMR- | Shield Mockups: Effect of Lead Thickness. | SEORr™ 2-01-057-66-113 STANDARD CONFIGURATION {tin. Ni). STANDARD CONFIGURATION PLUS 1 in. Ni GAMMA-RAY DOSE RATE {mr/hr) 102 50 60 7O 80 90 100 110 120 {30 {40 150 160 z, DISTANGE FROM SOURGE PLATE (cm) Fig. 12.7. Gamma-Ray Dose Rate Beyond RMR- Shield Mockups: Effect of Pressure Shell Thick- ness. enable shield optimization and to further study the secondary gamma-ray dose in and beyond the fead. A study of the effect on the total gamma-ray dose rate of adding 1 in. of nickel to the pressure shell of the standard configuration indicated that the extra thickness of nickel merely attenuated the primary gamma-ray dose (Fig. 12.7). A study - e Motk s e ik ."f_,é:‘ . ‘"'d'-were"maéi‘e be;fh in bora’recf wqfer and in plcun waer of the effect of adding an Inconel cladding on the boron curtain at the rear of the beryllium (Fig. 12.8) showed a negligible increase in the gamma-ray dose rate with 0.020 in. of Inconel and an increase of approximately 10% with 0.125 in. of Inconel. SEOMERee 2-01-057-66-45 e CONFIGURATION 2A, STANDARD GONFIGURATION (NG INGONEL ) o GONFIGURATION 1G, STANDARD CONFIGURATION + 0.020in. INGONEL 2 4 GONFIGURATION 1H, STANDARD CONFIGURATION +0.125 in. INCONEL GAMMA-RAY DOSE RATE {mir/hr) ~2 10 : 50 60 70 8> SO {00 {0 120 {30 {40 {50 ({60 2, DISTANGE FROM SOURGCE PLATE (cm) Fig. 12.8. Gomma-Ray Dose Rate Beyond RMR- Shield Mockups: Effect of Inconel Cladding on Boral. Additional gamma-ray dose rate studies in this static source series will include variations of the thickness of the beryllium reflector region and the addition of heavy metals in the reflector region. An investigation of the gamma-ray captures in the beryllium reflector and Inconel core shell is reported below. Neutron Measuremenfs “As was mentioned V_qbove,r neu’rron e UFSIaATS ockup tank A comparason was nof cffecfed‘;fo nof|'cecxb|e degree. The dnsplacemem‘ of ’rhe L-zdvTank plcnn water CUrVBs . _cnd the mockup tunk” In order to determme the distribution of the thermal-neutron flux in the beryllium of the mock- PERIOD ENDING JUNE 10, 1955 SEemET 2-01-057-66-18 BORATED WATER IN RMR TANK | | ——PLAIN WATER IN LID TANK | /PLAIN WATER IN RMR TANK N\ 40 N % 10 PLAIN OR BORATED . \ WATER IN RMR TANK— \" \:\ 1 i N A 1 PLAIN WATER IN LID TANK . 10 < ot 5. 5, ./. N //f L~ /j 2 4 THERMAL-NEUTRON FLUX {nv) o W % g Tt /\. / " 3 o FAST-NEUTRON DOSE RATE (mrep/hr) 1G°2 12 0 20 40 60 80 100 120 140 z, DISTANGE FROM SOURGE (cm) Fig. 12.9. Neutron Measurements in Plain and Borated Water (1.95 wt % B) ot LTSF. ups, gold foils were exposed throughout the re- flector region. The foils were placed on the faces of each beryllium slab from 5 in. above the source center line in intervals of 4 in. to the top of the beryllium. Plots of the results as a function of the beryllium thickness (Fig. 12.10) are uniform and peak just after the first face of the second 4 __Lberylhum slab (that is, at about 4 / in. of beryl- “#4% [jum), These data have not yet been corrected ;- for self-absorption and flux depression by the foils. Plots of the gold foil measurements as a function ...of the vertical distance above the source center ... line (Fig. 12.11) again are uniform, with a relatively .. flat region across the source center line. As would . be expected, the flux falls off rapidly as the upper - edge of the beryllium is approached. Neutron measurements (Fig. 12.12) were also 'spoce between the source <+ made behind the mockups as the lead region of .~ the shield was increased from 0 to 9 ]/2 in. to study «the contribution to the gamma-ray dose rate by secondary gamma rays from the lead. There were a great many low-energy neutrons in the region 197 ANP PROJECT PROGRESS REPORT DEOREF ) 2-01-057-66-119 108 THERMAL-NEUTRON FLUX {nv) <% STANDARD GONFIGURATION 5 x = DISTANCE ABOVE SOURCE CENTERLINE 0P 0 2 4 6 8 0 12 14 #, BERYLLIUM THICKNESS (in} Fig. 12.10. Thermal-Neutron Flux Distribution in the Beryllium of the RMR-Shield Mockups: Horizontal Traverses. Seenss 8 2-01-057-66-120 3 2, o, THERMAL-NEUTRON FLUX (nv) N OO) CONFIGURATION #=BERYLLIUM THICKNESS 10° : 5 9 13 17 24 25 29 33 #, DISTANCE ABQVE SOURCE CENTERLINE ({in.} F.‘ig‘./"'IZU.‘“. The;mal-Nefitron Flux Distribution in ihe Bery|li--um of the RMR-Shield Mockups: " Vertical Traverses. - T 4 2-01-057-66-122 4lin. Po &in. Pb 1Y in Pb T¥2in Pb 3in. Pb THERMAL-NEUTRON FLUX {nv) in.Pb 4Y%in. Pb 3in. Pb FAST-NEUTRON DOSE RATE (mrep/hr} e0 80 10C 120 140 DISTANCE FROM SOURCE {cm) Fig. 12.12. Neutron Measurements Beyond RMR- Shield Mockups: Effect of Lead Thickness. immediately behind the lead, and the increase in fead thickness, in effect, only moved the thermal and intermediate flux out. The slopes of the fast- neutron dose rate curves (A = 4.7 cm) indicate that these neutrons were predominantly of intermediate energies. Sodium Activation in Heat Exchanger Region A study of the activation by core neutrons of the sodium in the heat exchanger region was made as a function of the boral curtain thickness and the heat exchanger thickness. In each of the six con- figurations tested (Fig. 12.13), 12 in. of beryllium preceded the first boral slab. The last boral slab behind the heat exchanger region was followed by 1 in. of nickel, 4 l/2 in. of lead, and borated water, The measurements were made by placing small pellets of sodium flyoride (p = 2.1 g/cm®) in the center of each heat exchanger tank at the source center line. The plots in Fig. 12.13 show the measured specific activity of the pellets corrected to saturation for each case. For the first three cases the effect of decreasing the thickness of "w P i i - o oL F|g. ]2 14. RMR-Shield Mockup Experiment: PERIOD ENDING JUNE 10, 1955 SEGRET- 2-01-057-66-121 R INRUANNRNRLRN NN N R~ N DEENNRNN A D T B N A RN N 10° - — 7 = - ; . — 5 - ’ | D N - - 5-in-THICK %-in-THICK {%-in-THICK - ~ 2 el = I~ BORAL SLAB Ni WALL NaF SLAB ] E % (F=1.6g/cmd) s 2 N B o @ 4 \ - = 10 [~ \ e, .. _ % — ’ N ™~ - E | \ : \ = — \ _ % 5 . \ \ ] o — Y Y \ . | 9 \\ \\ \ A N - [t N, \ o ® \ e N\ = 2 \ <[ N\ x .\ o \ g 10° v9%% 552% P ) 7 %4 $7%% 2972 7% %25 L 7 o 7h%7 %7 2 gies Vo v = 7% Vo a7 74 o 7 E 5 2277 i 747 ) 0 a v w7z 7 b 477 77 = v 9957 7 74%% 7257 V = 7277 9247 7% B4 9% 7% = v 9972 27 7254 P 2% o ] = . 0 v 4 5 =4 7254 %77 %% 5227 o Z v gl , 7z %4%% 7 2979 %7 o 1 e 7 L % 2742 7% 9772 4 w42 7! [ ] Lo % [ 4 Loy 7 o Z 7742 4 ) A 4% e A NN, B R AN NN AR AN NN R N &\\\\\\\.\\\\\\\\\\\\\\\\\\\\\\\\\\\\ ‘\\\\\\\\\\\k\%k\ ANNTINNNR N S BN “ N STANDARD CONFIGURATION Flg. 12.13. Effect of Heat Exchanger and Boron Curfam Thicknesses on Sochum Activation in RMR- Shield Mockups. the firs*_r boral curtain is apparent. The results of BORAL THICKNESS (in) the last three cases show the effect of a 6-in.-thick © \ 2 3 4 5 heat exchanger with various thicknesses of the Sowrr 2-01-057-66-1234 4-end boral curtai in. A sample calculahon of 'rhe sodlum activation in the NaK in the heat exchanger of a 300-Mw alrplane reactor gave 2000 curies. |f the NaK-to-alr radiator were unshielded, this would give a dose rate of 16 rem/hr at a distance of 60 ft, 10% i | | [ | ,,iaooo | k - first and second boral curtain. g 5 1-2-4’,/1— i . . . p. = i-3-1q B £ oo ST e For each configuration an average specific £ ] 23 | N . , . . < TR activation was determined by integrating numeri- £ i—2-2 e . : £ N85 | cally; the results are shown in Fig. 12.14. Each 2 \ ©g . . . . 3 \\ & & 2000 soea) | point on the curves is labeled with a code in 5 o 7 . . . 3 \\ "% which the first and last numbers designate the Q A ] g 3 N % 2 v e s number of boral sheets preceding and following 8 N, HEAT EXCHANGER THICKNESS (in.) N 1 the heat exchanger, respectively, and the center I s 2-2-4% N_ 9 Y % N number des:gnates the number of heat exchanger « ~, & Y\ tanks in the conf:guraflon._ The mserf of the figure = . ~ o ' o _ is a cross plot showing an mcrease in the average o £ RGN specific activation with increasing heat exchanger | e p o 2 . thickness for two dlfferent ’rhlcknesses of the first & & v = " the best availabl Slot of thaf ‘f Tom Lo 2-01-058-38—1 Rt x=20"gin. 5 x =15 '/8 in. x=10 Y% in. x=2% in. 02 RELATIVE FAST -NEUTRON DOSE RATE (cpm) HORNYAK BUTTON COVERED WITH Y in. LEAD. 5 L4 TRAVERSE ¢ (PARALLEL TO AX!S OF PLANE). ¥ =DISTANGE BETWEEN INLET AND QUTLET AIR DUCTS REACTOR AND DUCTS 2, 3 AND 4 IN FIXED > POSITION REACTOR POWER 4 kw (NOMINAL) 100 50 0 50 00 150 " " DISTANCE FROM REACTOR CENTERLINE (cm) F:g 135 GE-ANPA"-Duct Mockup E xperlmenf Fast-Neutron Dose Rate Along Traverse 4. and its position has no great effect on that portion of the radiation which penetrates the shield. In order to determine some of the characteristics of _i{he angular distribution of radiation escaping | of ‘a”5-ft-long aluminum fube U ast- e “on_the intersection of the reactor center line and ‘the ‘second leg of the large inlet duct. The curve shows that the escaping radiation peaks quite heavily in the forward direction. There is no evi- "PERIOD ENDING JUNE 10, 1955 Wttt 2-01-058-38-3R1 o & 2 & w0 2 Q o 0 2 b 3 (0925 0 15 20 25 (o) /sec) Fig. 13.6. GE-ANP Air-Duct Mockup Experiment: Angular Distribution of Fast Neutrons. dence of fast neutrons escaping through the large duct, since the curve is symmetrical in the region around 50 deg. As soon as time is available, fur- ther investigations will be made of the efficiency of this type of fast-neutron collimation and of angu- lar distributions for other shields., The results of this experiment seem to indicate that the problem of shielding the air ducts of an air-cooled aircraft engine may not be so serious as was once thought. This conclusion applies spe- cifically to the annular air duct. THE SPECTRUM OF FISSION-PRODUCT GAMMA RAYS _ R. W. Peelle T. A. Love F. C. Maienschein One of ‘ 'fh em osfdlfftculf ‘ éi;o blems in the des ign | of a o-ft-long aluminum tube “of q divided aircraft shield arises from the lack of which acted as a collimator when | ' informati gr&\:_gg_wthg__epergy ‘spectrum and angular dis- .\'rr_iblqt:i‘c}_nrbf'gamxficjlr;jys at the outside of the primary reactor shield, These data are needed for an air- ‘2), this is equivalent to integrating over angle ¢, for a fixed angle . Substitution for UNCLASSIFIED 2-01-056-7-D133 - 8,y - PS(0,y - )= A _ S(MEASURED)( ¥4, e [ bta,4)a0 [ Deta,g) a0 ‘BEAM BEAM . / \\"&u BEAM DS(MEASURED)(Q’q’ 9_62) 2 O, u;"w a0 Fig. 14.18. Relaiionr;h‘ips -fof Dért'er.miri'ing Dose Scattering Probability from TSF Experiments. 213 TR T s G Dt © - ANP PROJECT PROGRESS REPORT Ps(6,4 Z 952) of the equation at the top of Fig. - 14.18 results in the final relation for obtaining, from the experiments, the average dose scattering probobllny PS(@) for a conical shell beam. The numerator in this relation is the integral of the S scch‘ered dose, around the side of the cylinder, “with respect to angle ¢,. The denominator is the integral of the direct-beam dose rate emitted from the equivalent point source. This relationship for conversion of the experi- ~ mental dose rate measurements to a fundamentally ‘basic probability function is general. |t applies “not only to the scattered dose rate at the outside surface of the cylinder but also to the dose rate at various depths into the cylinder. [n this case the probability function is also a function of the shield- ing thickness. This relationship is also applicable for the scattered dose rate at the front or rear face . of the cylmder or for the scattered dose rate within . ;:fhe crew compartment cavity. The method of - evaluation of the scattered dose integral [numerator ‘in the PS(6) relationship] from the detector tank measurements is presented in the following paper. The method of evaluation from the experimental results of the direct-beam dose rate integral [denominator in the PS(4) relationship] is given below, Evaluation of Direct-Beam Integral A plot of the relaxation length of the direct-beam fast-neutron dose rate vs total water thickness be- tween reactor face and detector is given in Fig. 14.19. Because of the large separation distance between the dosimeter and the reactor face in the TSF experiment, the relaxation lengths measured correspond to the point attenuation kernel for the fast-neutron dose rate. For comparison, relaxation lengths obtained from BSF data and corrected for geometry to correspond to a point attenuation kernel are also plotted in Fig., 14.19. The TSF and cor- rected BSF data agree well in the overlapping region, both as to magnitude and to trend. The variation of the dose rate measured at the rear of the detector tank as a function of beam . angle @ is shown in Fig. 14.20. The ordinate is ~ 'the ratio of the dose rate for beam angle 6 to the . .dose rate for § = 0 deg, that is, when the reactor " is pointed directly at the detector tank. For a given “““dngle @, the dosimeter in the detector tank reads " the direct-beam dose rate coming off at angle 6 “with respect to the line of symmetry of the beam 214 2—-01-056-T-Ai28 o—eo—5 ** o O TSF DATA ® BSF DATA CORRECTED FOR GEOMETRY rlem) 0 0 50 ' 100 150 TOTAL WATER THICKNESS BETWEEN REACTOR FACE AND DETECTOR (cm) Fig. 14.19. Relaxation Length (\) for Fast- Neutron Dose. serRet 2-04-056-7- 129 5 DIRECT BEAM INTEGRAL* J' RECTED (ppan) 2 (9)-!2- 2 sin 8 00 =167 x10"2x ¢ 0.5 CALCULATED DIRECT on 3 02 PLUS ESTIMATED 9 SCATTERED Bl od ‘—- g MEASURED DIRECT wlul 0,05 PLUS SCATTERED Qo oo 0.02 CALCULATED DIRECT 0.04 0.005 ME ASURED CONTRIBUTION = 12.5% SCATTERED DOSE 0.002 AT SIDE 0.004 : 0 20 40 60 80 100 420 440 8, BEAM ANGLE {deg) Fig. 14.20. Angurlur Distribution of Dose at Rear of Detector Tank (p = 45 cm). e e o ki v bt " “verse R? relationship, D_ o ‘p‘|u‘s an air-scattered dose rate from the other por- tions of the beam. At § = 0 deg the direct-beam dose rate is large, and therefore the scattered dose rate is negligible. As @ increases, the direct-beam dose rate decreases rapidly, owing to the beam - collimation, and the scattered dose rate becomes proportionately larger. In order to evaluate the direct-beam dose rate integral, it is necessary to subtract out the scattered dose rate. The method used was, briefly, the following: The neutron ~angular emission from the faces of the reactor was assumed to have a cosine distribution. The direct beam from the reactor faces was attenuated along a straight-line path by using the relaxation lengths shown in Fig. 14.19. The results of this procedure for the various angles @ are shown in Fig. 14.20. The calculated direct-beam curve and the measured direct- plus scattered-beam curve agree well for ¢ below about 45 deg, where it is to be expected that the scattering contribution will be relatively small. “For @ greater than about 45 deg the calculated direct-beam dose rate drops rapldly, and, at 90 deg, the direct-beam dose rate is down by a factor of 10%. As expected from the geometry, the dose rate measured at § = 90 deg is an almost totally scattered dose rate. The scattered dose rates measured at the side of the detector tank for @ = 90 and 135 deg are also given in Fig. 14.20. The ratio of rear- to side-scattered dose rates for these two angles is about 1.6, If this ratio is assumed to be valid to 6 = 60 deg, an estimate of the rear-scattered dose rate based on the measured sco'rtered dose rate at the side can be made for 6 = 60 deg. If this is done and fhe esflmated rear-scofiered dose rate at @ = 60 deg is added fo the calculcted direct- ‘beam dose ra're, the result IS fhat gwen by fhe”, ’ o corrected the dose rate which would have been obtained on PERIOD ENDING JUNE 10, 1955 the surface of a unit sphere drawn about the equivalent point source. The element of solid angle dQ is 2% sin 6 d6. The direct-beam dose rate integral for the TSF experimental beam (at =45 ¢m) is equal to 1,67 x 1072 x {2, Calculation of Scattered Dose Figure 14.21 presents a typical plot of £2p5(6) vs the angle 0 of the conical shell beam. Note that the ordinate is the product of the square of the separation distance by the dose scattering probability for a conical shell beam for this sepa- ration distance. The dose scattering probability presented in Fig., 14.21 is for a water thickness p = 45 cm at the reactor, a water thickness ¢ into the side of the cylinder of 5.2 cm, and a separation PG 2-01-056-7-A136 O T T I T I | LET D; (8,0) = DIRECT-BEAM DOSE OBTAINED AT ) DISTANCE ¢ FOR A GIVEN REACTOR SHIELD DESIGN | THEN: 0= [ 0 (8,0} a2 P(8) dp ,u 0.05 - o4, e | s e ‘ pszffi_zué__ {12/:3(9)] I @ £ 0.02 \ & w N Qoo N = N\ \\ 0.005 \\ 30 0 780 9ol U427 T 450 180 8, CONICAL SHELL BEAM ANGLE (deg) Fig. 14.21. Dose Scattering Probability for Coni- cal Shell Beam as Obtained from TSF Experiments (o = 45¢cm, t = 5.2 cm, { = 64 #). 215 e - ANP PROJECT PROGRESS REPORT diétqni_:e ffiof 64 ft. The functional dependence of P*(6) on p, t, and 1 should be accounted for in o sccfl‘ered dose rate caleulations, In the present expenmenfs a range “of values of t is covered for a " single sepcrahon distance £ and for a single value - of p; however, some data were obtained with p - changed from 45 to 15 cm, and the effect of this - change on the dose scattering probability is dis- . cussed later. To illustrate the use of the dose " scattering probability for a conical shell beam in the prediction of scattered dose rate, assume that ~the angular distribution of the dose rate emission from a prlmary reactor shield is obtained by dose rate measurements af a distance d from the reactor. Thls dose rate d:strlbuhon obtained at distance 4 is denoted as Dd(é? d). Then D6, d)d2 is the dose ‘rate which would be obtained on the surface of a “ “unit sphere drawn about the equivalent point source. " The scattered dose rate is obtained by weighting "~ Dj(6, ,4)d? by the scattering probability PS(6) de- fermmed from the differential experiments and integrating with respect to y = cos @ {first integral in Fig. 14.21). The second integral in Fig. 14,21, which is equivalent to the first, is in terms of the £2p<(6) used in the plots. Effects of Direct-Beam Collimation Figure 14.22 shows the effect of beam collimation on the dose scattering probability averaged over the beam. These results were obtained from single air-scattering calculations. The beam in the TSF experiments had approximately a cos? distribution. The cos® and cos'® distributions represent an increasing degree of collimation of the beam. |t must be remembered that in the interpretation of the TSF experiments the assumption was made that the dose scattering probability averaged over the TSF experimental beam was equal to the dose scattering probability for a differential beam having - an angle @ equal to the angle of the axis of symme- try of the experimental beam. These results indi- -cate that this assumption is valid within 10 to 15% - for beam angles greater than 30 deg. For beam " angles between 0 to 30 deg the averaging of the dose scattering probability is quite sensitive to beam collimation, and therefore the experimental results obtained are not valid if any sharp well- jdefmed beams are emitted from the reactor shield - in this range of beam angles. Consideration of multiply scattered neutrons in the calculations "~ would have tended to reduce the effect of beam UNCLASSIFIED 2-04-056-7-A430 0.3 0.2 04 0.05 2 LB 18 0.02 0014 0.005 0.002 0.001 ' 0 30 60 90 120 150 i80 8, BEAM ANGLE (deg) Fig. 14.22, Effect of Beam Collimation on Scattering Function Based on Single-Scatter Calcy- lation. collimation shown on this figure because the sources of multiply scattered neutrons are of necessity more diffuse than the sources of singly scattered neutrons, Effect of Neutron Energy Spectrum A small amount of data was also obtained for a water thickness of 15 em at the reactor to obtain an indication of the effect of change in neutron spectrum on the dose scattering probability. Figure 14.23 presents the angular distribution of the dose rate at the rear of the detector tank for p = 15 cm of water. Again, to obtain the direct- beam dose rate, it is necessary to subtract out the scattered dose rate. The procedure descnbed for the p = 45 cm case was used, but the results are not nearly so satisfying. The calculated direct- beam dose rate has a flatter distribution than the measured direct plus scattered dose rate in the region of O =0 to 45 deg, where the scattered dose rate should be a relatively small contribution. For all the data obtained at p = 45 cm, both for the direct-beam and scattered dose rates, a plot of the dose rate vs the beam angle always showed a slight dip in the curve in the region of 6 = 30 deg. ‘This dip has been attributed to an out-of-roundness i o i deaails Gm i ol i, i ca Fig. ST 2-01-056-7~A137 ~1 2 RECT DOSE INTEGRAL= 9.6x10 'x.J 0.5 CALCULATED DIRECT PLUS ESTIMATED SCATTERED 0.2 04 MEASURED DIRECT 0.05 PLUS SCATTERED O deg 0.02 DOSE AT & DOSE AT 8 CALCULATED DIRECT .01 0.005 CONTRIBUTION= {8 ©.002 0.004 0 20 40 60 80 100 120 140 8, BEAM ANGLE {deg) Fig. 14.23. 'Afigijlur 'Distribufion of Dose ‘u;"Rear of Detector Tank (p = 15 ¢m). of the reactor tank. For the 15-cm case, the ef- fect of an out-of-roundness of the reactor tank ~ would be much more pronounced owing to much shorter relaoxation |engfhs. The discrepancy ob- tained here is believed to be due, in part, to this effect, The direct-beam dose rate integral was obtained from the curve drawn through the data points from 0 to 45 deg and faired into the calcu- lated curve from 60 to 90 deg. The contribution to the direct-beam dose lntegral of the dose rafe from _ - scaflered neutrons into the side of the detector “tank is qu:te chfferent for the p = 15 cm case com- ‘pared with the p = 45 cm case. Examination of the “"relaxation lengths for the two cases shows that they ’are approxlmotely proporhonal to the relaxchonr Ieng'rhs for the dlrect-beom dose. More dafa are | _50 to 90 deg is only about 18%. o N b Frgure 14, 2% compores\’rthe Ime-beam dose scatter- mg probobsl n‘y obfcmed fl"om the TSF me =z S XA l5|de_6f the detectoi' ank. : 14, 24 e that the effect of this dif- “ference in energy’ Spectfbm’on the air-scattering process is not large, the difference in the relative air scattering between the two cases being de- ‘PERIOD ENDING JUNE 10, 1955 SUERES 2-01-056-7-A135 0.05 0.02 12 % DIFFERENCE 0.01 0.005 0.002 Dy, £)d0 0.004 BEAM 7 0.0005 0.0002 0.0001 o) 4 8 12 16 20 249 /, WATER THICKNESS AT SIDE OF DETECTOR TANK (cm) Fig. 14.24. Effect of p (Water Thickness at Re- actor) on D s 2P0 =0, f ~ p, =0) = | L edam D d(Q,{) dQ ' termined as only 12%. However, a rough estimate of the uncertainties in arriving at this percentage difference adds up to the order of 20%. Hence, at least a £20% figure should be placed alongside the 12% difference indicated. More data for larger ‘water thicknesses at the reactor are needed to definitely tie down this point. The plot shows that the relaxation length of the “needed to definitely tie down this apparently simple relahonshlp. iiiiee SHIELDING EXPERIMENTS J. E. Yan Hoomissen differential experiments (see preceding discussion, *“The Differential Shielding Experiments at the TSF: Phase I’’) has made possible a more detailed 217 APPL!CATION OF THE DlFFERENTIAL - _The hknowyledge gumed “from the first data of the il R e | S rflro cm 20¢cm __.‘ Kmm = = Ym = X 40 ¢m Y= ———— == X — n 0.38In. STEE 0.50in. STEEL L 0.62in, STEEL FRONT COMPARTMENT A 2.05in. LEAD (3 PIECES) COMPARTMENT B 22.7¢cm Jd2in. STEEL .78in. STEEL 3in. STEEL in. STEEL 3in x 3in. LEAD J——x— XX K e e X Fig. 14.25. 38in. STEEL! C2-01-05%6-~3-T6A 35.25in. 50-in. GASKET 38in. STEEL X o I_ 40 ¢cm ';.] ; _ 20cm r- | TSF ' -~ REAR ——->|<—:—x———— REACTOR COMPARTMENT F 15in. STEEL _1.5in. LEAD COMPARTMENT D 0.62in. STEEL 0.42in. ALUMINUM COMPARTMENT € VoD 0.62in.STEEL COMPARTMENT E i 1 X [ 10cm X {' - 20¢cm _ ; X - POSITIONS OF GEOMETRICAL g | , CENTERS OF COUNTERS Mockup of G-E R-1 Shield. § * - - » 1¥0d3Y SSTUO0U LDIrOYd dNY A " - w j:cm for TSQ(__deg. Sl_n -1816, p 158; ORNL-1864, p 178. analysis of the measurements taken in the divided- shield mockup experiment reported earlier.® It also serves as a basis for a prediction of the results of future mockup experiments. It is the purpose of this paper to correlate the results, to date, of the mockup experiment and the differential experiments and to outline the procedures for future predictions. In the mockup experiment the reactor was encased in the R-T shield (Fig. 14.25), and measurements were made with a fast-neutron dosimeter along the x axis in the detector tank, The experimental ~values obtained in that experiment are compared with values predicted by applying the data differ- ential and the procedures given in the two preceding papers. Since the differential experiment includes x traverses across the detector tank and z traverses up and down in the tank, the extrapolation of the dose rate measurements to the surface of the four faces of the tank _WQUId give dose rates (Fig. 14.18) from a line beam? emitted at a given g as a function of the angle ¢, (Fig. 14.26). The integra- tion of the area under the curve in Fig, 14.26 would then be equivalent to the dose rate at a point on the surface of the cylindrical crew com- partment from a conical shell beam at the same value of 6. A plot of a series of these integrated dose rates as a function of water thickness of a cyhndrlcal crew comportmen‘r side shield can then be made for # = 0, 30, 60, 90, 135, and 180 deg, for which experlmen‘ral data are available. A composite plot (Flg. 14.27) shows the actual attenuation curves that should be used in designing a cylmdrlcal crew compartment, The average relaxation length of the curves in e PERIOD ENDING JUNE 10, 1955 oo Z2-01-056-3-A149 20 2 o FAST-NEUTRON DOSE RATE (mrep/hr/watt x 10°) 10 / 5 \ / 0 0 90 180 270 360 4>2, AZIMUTHAL ANGLE AT CREW COMPARTMENT (deg) Fig. 14.26. Fast-Neutron Dose Rate from a Line Beam as a Function of ¢, for 0 = 60 deg, p = 45 cm, t = 0, Flg- 14,27 varies from 64 7ch for 9 0 deg fo__75 9 oo 3c E. Clifford i"al., ANP Ouar, Prog. Reps., 4The concept of a line beam is discussed in the pre- ceding poper, ‘ Slnce the( dose rate emn‘fed at eoch‘cmgiua 0 from ORNL-" cylmdrlca! crew. comparfment ‘or the equivalent * thickness :n the s:de of fhe defector tank, for a ‘unit dose rate emltfed in a comcal shell beam at - any angle @ can be deduced from Flg. 14,27, A comprehens‘n}e eXper‘lmenft“ is now under way at the TSF to determine this angular distribution, that is, the fast-neutron dose rate leaving the 219 _ANP PROJECT PROGRESS REPORT ~SECRET 1 2-01-056-7-A—{359A 407 10-3 - ZZPS(B) x1.67 ny 2 o- 8=135 deg f=180d 5 2 10-6 52 12 {72 232 292 352 412 472 DISTANCE IN FROM SIDE SURFACE OF A CYLINDRICAL CREW BOX (cm) Fig. 14.27. £2P5(6) for a Conical Shell Beam as a Function of Water Thickness at the Side of Cylindrical Crew Compartment for Yarious Angles of 0 and for p = 45 deg and £ = 64 ft. shield as a function of angle 6; in the meantime, a preiiminory measurement of this distribution (Fig. 14.28) is used in thls analysis as the best available data, Ata given water thickness the values of dose “rate as a function of 6 can be read from Fig. 14.27. . -__;These pomts — an example for a given water thick- ~niess is shown in Fig. 14.29 — are plotted, and a “smooth curve is drawn between the points. This shows the ‘dos’e_rrafe_fo be expected at agiven water : _,220. 2-01~056-7-Al46A » o W o o ™ N\ [\ 0o @ —— ] FAST-NEUTRON DQSE RATE AT A DISTANCE OF 64 ft FROM REACTOR (mrep/hr/waft x10% rn ~ ’ O £ \ -—/ o o 0 P ] / o J o TN 0 20 40 60 80 100 120 140 160 180 REACTOR ANGLE & {deq) o © Fig. 14.28. Estimation of Fast-Neutron Dose Rate at 64 ft from Reactor Encased in R-1 Mockup as a Function of Reactor Angle 0. thickness from a conical shell beam at any angle 6. This distribution is weighted by the angular dis- tribution of the mockup and integrated to determine the total dose rate to be expected at a given water thickness in from the side of a detector tank. A plot of the total dose rate at a given water thick- ness as a function of @ for the R-1 mockup is given in Fig. 14.30. This process was repeated for several different water thicknesses at the detector tank, and the results, along with the experimentally measured values, are shown in Fig. 14,31, |In general, the agreement is good. The apparent dis- agreement can be attributed to three factors, the first of which is the uncertainty of the angular distribution of the fast-neutron dose rate emitted from the R-1 mockup. This uncertainty, possibly the largest contributor to the disagreement, will be cleared up by future experimentation, The second factor contributing to the d:sagreement is the low count rate encountered in the actual R-1 experi- ment. The third factor is the lack of complete collimation of the direct or unsccttered becm in the ] differential experiment, as pointed out in ‘the preceding paper. L FOFPERIPny A g i, R A M » " coss dlstrlbuhon. _,;‘The dnfferem‘lalfi"exp’eraments in whlch ‘the TSF reactor was placed in the reactor 2%P°(0) x 167 O 30 60 - PERIOD ENDING JUNE 10, 1955 eFereT 2-04-056-7~A-140A 90 120 150 180 8, ANGLE BETWEEN CONICAL SURFACE BEAM AND REACTOR-CREW COMPARTMENT AXIS (deg) Fig. 14.29. 12P%(6) ot a Point 37.5 cm in from Surfucekof Cylinarical Crew Compartment as a Function of Reactor Angle 0 for p 45 deg und 13 = 64 ft The method to be used in reachmg a predlcf:on of the dose rate to be expected in the crew com- partment with the reactor in the R-1 mockup can now be illustrated. Up to this pomt the dISCUS- snon has’ beenulrlmited to measurements made in the 3). E. Faulkner, Focusing of Radzatzon ina Cylmdrzcal Crew Compartment, ORNL. CF-54-8-100 (Aug. 18, 1954) the _dose rate ’measureé_ gf th_e center of the cclylfy' 1 trengfh c’r the surface of th _ cavutym _at the ang|e 9 hove /been determme&? The lcmo of ) these values for different ‘angles of 6 (Table 14.1) tank and measurements were made in the cavity of the GE-ANP crew compartment mockup allow an estimate of this ratio to be made experimentaily. The dose rate at a given water thickness at the s:de of fhe cy||ndrlco[ crew compartment (Fig. ed and IS qssumed to be the source From the experlment (F:g. 14.12) the ‘f"‘dose rafe at the center ‘of the cov|ty from a line _,_\_.beam qt ‘the angle 6 is known. This experimental " value |'s 'cfllawsted to the correspondlng value (that s, it is multlplled by 27) at ‘the center of the _cavity for a comcal shell beam based on the pro- ) v'_':__‘";cedure Voutllned in the precedmg paper. Ncnw, both is shown to be effectively a constant with an average value of 5,1. This indicates that the 221 28) from a comcal shell ‘beam at the angle 6 o FAST~NEUTRON DOSE RATE SCATTERED IN SIDE OF DETECTOR TANK . ANP PROJECT PROGRESS REPORT SECRET 2—04—-056-7-A147A " 150 .. n o / o S — o ~ a 0.50 \ 0.25 | 0 20 40 &0 80 100 {20 {40 160 180 ‘ REACTOR ANGLE 8 {deq) Fig. 14.30. Fast-Neutron Dose Rate Scattered in Side of Detector Tank as a Function of Reactor Angle 6 (Reactor in R-1 Shield, ¢t = 10.2 ¢m). angular distribution of the neutrons leaving the inside surface of the crew compartment can be assumed to be the cos3 of the angle measured from the normal to the surface. This picture, along with Faulkner’s work,® neglects any neutrons that might go across the cavity and be reflected back into it, This should be a small contribution to the source strength, on the order of 10%, if an albedo of ]/10 is assumed, - The dose rate at the center of the cavity of the ‘cylindrical crew compartment with the reactor en- cased in the R-1 shield mockup can now be pre- dicted. The source strength at the surface of the cavity is found by making a calculation as out- lined above except that a water thickness equal to the side shielding of the crew compartment mockup . is chosen. This source strength is integrated (multiplied by a factor of 5.1) to give the con- tribution to the dose rate at the center of the crew 222 TABLE 14.1. RATIO OF DOSE RATE AT CENTER OF CREW COMPARTMENT CAVITY TO DOSE RATE AT INSIDE SURFACE AS A FUNCTION OF @ Dose Rate at 9 Corrected Dose Rate Surface of (deg) at Center of Cavity Cavity Ratio™* (mrep/ hr/w) (mrep/he/w) 0 1.55 % 1079 3.14x 1078 485 30 .08 x 1073 2.10x10°% 5.14 60 5.30x 1078 9.70 x 1077 5.46 90 1.86 x 1076 3.80 x 107 4.89 180 3,15 % 1077 6.00x10°8 5.25 * Average value of ratio = 5,12, compartment mockup from neutrons scattered through the side of the compartment. The dose rate contribution at the center of the cavity resulting from neutrons entering the rear and the front of the crew compartment can also be found from the differential data by using the y traverses for a line beam at a given @ (Figs. 14.3 and 14.4). This is converted to data for a conical “shell beam by multiplying by 2m. This gives the dose rate at a given water thickness in the front or the rear of the crew compartment as a function of the angle 6. Again, by weighting these distribu- tions by the angular distribution of the R-1 mockup, the source strength at the inner surface at the front and rear of the crew compartment can be found. An angular distribution of emission from the inner surface must be assumed and an integro- tion performed to find the dose rate at the center of the cavity, A good approximation is to again use a cos® distribution for the front of the crew compartment, but for the rear the distribution will be closer to a cosine distribution because of the existence of thick lead. The three component parts of the dose rate at the center of the cavity (the neutrons scattered in the side, the rear, and the front) are then added. This prediction will be completed when a more exact measurement of the angular distribution of neutrons leaving the R-1 mockup has been made, N 1. e e Y £ [T - o ki QQAI » » FAST-NEUTRON DOSE (mr‘ep/hr/wafi) PERIOD ENDING JUNE 10, 1955 SELRET -5 : ’ ' 2-04{~056-3-22-85R1A hH=1495 f+ d= 64 ft a= 90 de CALCULATED CALCULATED MEASURED MEASURED CENTER OF TRIPLET FAST-NEUTRCN DOSIMETER AT: x = VARIABLE y=90c¢m 2= 0Ocm ~-78 —-66 -—54 -42 -30 -8 -6 0 6 18 30 42 54 ‘ x, HORIZONTAL DISTANCE FROM & AXIS TO DETECTOR CENTER (cm) Fig. 14.31. Fast-Neutron Dose Rate Along x Axis of Detector Tank. 66 78 223 3 3 " T Part IV APPENDIX g 'Lyl () > o - © CFi55-2-93 CF-55-484 15. LIST OF REPORTS ISSUED FROM FEBRUARY 1955 THROUGH MAY 11, 1955 : :_. Crn‘:cal Assembly wnh /8 in. Incone! Core Shells e IS1’IC$ of Confrol Rods and Mcferlals o Potenhally Suncble for Use in fhe o ART F’arf || | o CF-55.4-137 tio ontrol Rod terials. Part 11]: The Effect of Neutron frradiation on Some Rare Earth Samples bLEvdluotion' of ART ControlRod Ma- S AUTHOR(s) H. C. Hopkins L.. T. Anderson W, T. Furgerson, H. C. Hopkins C. S. Burtnette A. S. Thompson A. S. Thompson V. J. Kelleghan E. 5. Bettis B. Cottrell A. S. Thompson A. S, Thompsen L. T. Anderson Tc Jo B(:I”es L. A. Mann REPORT NO. TITLE OF REPORT . Reflector-Moderated Reactor CF-55-2-142 Vertical Component of Fuel Forces on Reflector and Island CF-55-3-161 Gamma and Neutron Heating of the ART Fue! Pump Assembly CF-55-3-167 Circulating-Fuel-Reactor-Powered Ramjet CF-55-3-191 Fission Product Heating in the Off-Gas System of the ART CF-55-4-34 Empirical Correlation for Fatigue Stresses CF-55-4-44 Allowable Operating Conditions ~ CF-55-4-83 High Temperature Valve Information Summary CF-55-4-8.7 ART Design Data CF-55-4-116 The ART Off-Gas System CF-55-4-124 Flexible Mounting Systems CF-55-4-159 Thermal Stresses in Tube-Header Joints CF-55-5-76 Calculation of the Beryllium Contri- bution to the ART Temperature Coefficient of Reactivity CF-55-5-93 Surface-Yolume Ratios for Five Different Fluoride Fuel Systems H. Ex_prerrimeptal Engineering CF-55-2-100 ART Reactor Accidents Hazards Tests " Critical Experiments Three Region 'Ref‘lecfa{Mader&éJ T R. M. Spencer ,Eyaluohon of Reachvny Chdracfer_ AT " J. W, Noaks J. W. Noaks DATE OF DOCUMENT 3-2-55 3-28-55 3-24-55 3-28-55 4-5-55 4-11-55 4-5-55 4-18-55 4-21-55 4-11-55 4-25-55 5-11-55 5-12-55 2-11-55 2-14-55 4-13-55 4-25-55 227 ANP PROJECT PROGRESS REPORT REPORT NO, CF-55-4-178 CF-55-4-18 - CF-55-2-79 CF-55-3-157 CF.55.3-179 - CF-55-4-167 CF-55-4-181 CF-55-2-89 CF-55-2-148 CF-55-3-15 CF-55-3-47 CF-55-3-61 CF-55-3-62 : C.F-55-3-137 CF-55-3-174 ' .CF-55-4-32 8 Examination of First Three Large TITLE OF REPORT Evaluation of ART Control Rod Ma- terials, Part IV: The Variation of Reactivity with Control Rod Diameter AUTHOR(s} J. W, Noaks IV. Chemistry Solubility of Composition 30 in Water V. Corrosion Data and Results of ARE Corrosion Capsules Fluoride Pump Loops Examination of Inconel-Flueride 30-D 'Pu_mp Loop Number 4695-1 Examination of Sedium-Incone! Pump Loop 4689-4 Examination of Fluoride Pump Loops 4930-A and 4935-1 J. C. White R Baldock | G. M. Adamson, R. S. Crouse G. M. Adamson, R. §. Crouse, P. G. Smith G. M. Adamson, R. S, Crouse G. M, Adamseon, R. S. Crouse Vl. Heat Transfer and Physical Properties Measurement of the Viscosities of Composition 35 and Composition 74 Electrical Heating and Flow in Tube Bends Qualitative Velocity Profiles with Rotation in 18 Inch ART Core Heat Capacity of Lithium Hydride Measurement of the Viscosity of Composition 72 Measurement of the Viscosity of Composition 30 Measurement of the Viscosity of ' Composition 43 Status Report on Forced Convection Experimental Work in Converging and Diverging Channels with Volume Heat Sources in the Fluids Measurement of the Viscosity of Composition 2 S. I. Cohen, T. N, Jones H. W. Hoffman, L. D. Palmer, N, D, Greene G. L. Muller, J. 0. Bradfute W. D. Powers, G. C. Blalock S. l. Cohen, T. N. Jones $. I, Cohen, T. N. Jones 3. I, Cohen, T. N. Jones H. F. Poppendiék, N. D. Greene S. l. Cohen, T. N. Jones DATE OF DOCUMENT | 4-29-55 4-4+55 2-15-55 3-24-55 3-28-55 4-21-55 4-26-55 2-15-55 2-22-55 3-1-55 3f7'55 . 3'8'55_ 39:55 _3,-] 6-55 3-24-55 4-1-55 e ) ot s it %uh%n%.m‘ i nbi ol o ¥, o PERIOD ENDING JUNE 10, 1955 REPORT NO, TITLE OF REPORT AUTHOR(s) DATE OF DOCUMENT Yili. Radiation Damage _t CF-55-2-36 The Fate of Certain Fission Products M. T. Robinson, in the ARE S. A. Reynolds, _ H. W. Wright 2-7-55 F CF-55-4-16 Ru Deposition in In-Pile Loop M. T. Robinson 4-5-55 CF-55-5-22 A Theoretical Treatment of Xe 135 . Poisoning in the ARE and the ART M. T. Robinson 5-2-55 Vill. Shielding CF-55-2-111 Calibration of the Revalet, a Remotely Variable Lead-Trans- mission Gamma-Ray Dosimeter D. L. Gilliland CF-55-4-122 Spectrometer Measurements of Fission - Product Gamma Rays for the CFR E. P. Blizard 4-21-55 IX. Miscellaneous ORNL-1864 Aircraft Nuclear Propulsion Project Quarterly Progress Report for Period Ending March 10, 1955 A. W, Savolainen (ed.) 4-13-55 - " 229 T Y (o ni...,. a s ey ey