CENTRAL RESEARCH LIBRARY' - n DOCUMENT coux@rx A ' £ . g (&) : | 5 L ‘5*4 s 'l“”l |l||||'l||||'|!'||“l=|‘i‘”|||-"|||‘|i|‘-|||||'| \IH.||I‘|||I‘I| Reucfors-gz;:bnl::?:nd Power .'.. \y 3 445b 0349L8E L (& € ma’{ “j c: 0 me 3 ,r o N E = \%! PRELIMINARY CRITICAL ASSEMBLIES OF THE % IS S tf‘"\’! REFLECTOR MODERATOR REACTOR - o ! R. M. Spencer : ; : f isar s ey % CENTRAL RESEARCH LIBRARY DOCUMENT COLLECTION LIBRARY LOAN COPY DO NOT TRANSFER TO ANOTHER PERSON If you wish someone else to see this document, send in name with document and the library will arrange a loan. OAK RIDGE NATIONAL LABORATORY OPERATED BY CARBIDE AND CARBON CHEMICALS COMPANY A DIVISION OF UNION CARBIDE AND CARBON CORPORATION (=3 POST OFFICE BOX P OAK RIDGE. TENNESSEE L »n ORNL 1770 This document contains 89 pages This is copy ¢ of 197 Series A - Subgject Category Reactors - Research and Power PRELIMINARY CRITICAL ASSEMBLIES of the REFLECTOR MODERATOR REACTCOR Experimentation by D V P Williams R C Keen (Louisiana State University) J J Lynn Dizxon Callihan Reported by R M Spencer, USAF DATE ISSUED NOV 22 1958 PHYSICS DIVISION A H Snell Dairector Contract Ko W-Th05, Eng 26 OAK RIDGE NATIONAL TLABCRATORY Operated by CARBIDE AND CARBON CHEMICALS COMPANY A Division of Union Carbide and Carbon Corporation Post Office Box P Oak Ridge, Tennessee - 3 4456 0349kBL b t "y “fi. 11- ORNL 1770 1 103 10L4-106 107 Reactors-Research and Power M-3679 (14th edition) INTERNAL DISTRIBUTION C E Center 46 ¢ P Keim Biology Library 47 C S Harrill Health Physics Library 418 ¢ E Wainters Central Research Library 49 D S Billington Reactor Experimental 50 D W Cardwell Engineering Library, 51 E M King Laboratory Records Department 50 E O Wollen Laboratory Records, ORNL R C 53 A J Miller C E Larson rqufix\ 54 J A Lane L B Emlet (K-25) N 55 R B Briggs J P Murray (Y-12) \\ L D Roberts A M Weinberg v R N Lyon E H Taylor W C Koehler E D Shipley E P Bfi]}[fzard A H Snell M E ,Rose F C Vonderlage D D# Cowen W H Jordan W M Breazeale (consultant) S J Cromer M J Skinner G E Boyd J E Sherwood R A Charpie Dixon Callihan J A Swartout A P Fraas S C Lind T A Welton F L Culler D K Holmes A Hollaender R R Coveyou J H Frye, Jr E> 8 Bettis W M Good C B Mills M T Kelley R M Spencer G H Clewett Dunlap Scott R S Livingston D V P *Williams K Z Morgen E L Zimmermsn T A Lincoln ORNL Document Reference A S Householder Library, Y¥-12 P%Fnt EXTERNAL DISTRIBUTION AF Plant Representative, Burbank AF Plant Representative, Seattle AF Plant Représentative, Wood-Ridge American Machine and Foundry Company ANP Project Office, Fort Worth Argonne National Laboratory Armed Forces Special Weapons Project (Sandia) Armed Forces Special Weapons Project, Washington Atomic Energy Commission, Washington Babcoek and Wilcox Comp . Battelle Memorial Instif:fl‘% g | Bendix Aviation Corpqratign ... Brookhaven National Igh SR Bureau of Ships i ) T A 108 109 110 111 112 113-117 118 119 120 121-123 124 125-132 133 134 135-138 139-140 141 142 143 144 145 146-147 148 149 150-151 152 153 154-160 161 162 163 164 165 166 167 168 169-1T70 171-172 173 174-179 180-19k 195 196 197 -11i- Chicago Patent Group Chief of NR}val Research Commonwealth Edison Company Department of the Navy - Op-362 Detroit Edisch Company duPont Comp Augusta duPont Compeny s Wilmington Duquesne Light Company Foster Wheeler Coxporation General Electric Company (ANPD) General Electric Company (APS) General Electraic Company, Richland Hanford Operations O %ice Iowa State College % Knolls Atomic Power Laboratory Los Alamos Scientific Laboratory Nuclear Metals, Inc Monsanto Chemical Cbmpanyflfl§( nony Mound Laboratory 3 National Advisory Committee for Aeronautics, Cleveland National Advisory Committee for Aeronautics, Washington Natal Research Laboratoéy | Newport News Shipbuilding and Dry Dock Company New York Operations Office North American Aviation, Inc Nuclear Development Assoclates, Inc Patent Branch, Washington Phillips Petroleum Company (NRTS) Powerplant Laboratory (WADC) Pratt & Whitney Aircraft Division (Fox Project) Rand Corporation San Francisco Field Office . Sylvania Electric Products, Inc \ Tennessee Valley Authority (Dean) \ USAF Headguarters \ U S Naval Radiological Defense Laboratory \ University of California Radiation Laboratory, fibrkeley University of California Radiation Laboratory, Livermore Walter Kidde Nuclear Laboratories, Inc Westinghouse Electric Corporation Technical Information Service, Oak Ridge Division of Research and Medicine, AEC, ORO Pacific Northwest Power Group Materials Laboratory (WADC) ABSTRACT Five preliminary critical assemblies, planned in conjunction with a full scale aircraft type circulating fuel reactor, were constructed at the Oak Ridge Critical Experiment Facility The first assembly was constructed with a beryllium island and reflector, and had enriched uranium disks placed between sodium filled stainless steel cans arranged 1n a fuel annulus The assenmbly was critical with 15 1 kg of U-235 The second assembly was also constructed with a beryllium i1sland and reflector, however, a powdered mixture of sodium, zirconium and uranium fluorides packed in aluminum tubes wes used as a fuel, and fuel end ducts were provided The assembly was critical with 7 7 kg of U-235 The addition of structural poisons (fifth assembly) to this assenbly increased the critical mass to 18 2 kg of U-235 In the third assembly the beryllium island and reflector were replaced by graphite With 17 3 kg of U-235 in the fuel reglon, this assembly was not critical In order to make the assembly critical approximately 80% of the graphite in the island and imner six inches of the reflector were replaced by berylliuwn The fouwrth assembly was constructed as a two region assembly with a 1 to 1 volume ratio of graphite and powdered fuel in the core and a beryllium reflector This assembly was critical wvaith 15 2 kg of U-235 Neutron flux distributions were determined in these assemblies using indium foil activation Power distributions were determined using aluminum catcher foils in contact with wranium fuel disks Measurements were made on the effect od reactivity of placing various materials in the fuel and reflector regions, and an estimate of the neutron leaksge was obtalned for two of the assemblies II ITI VIII -vi-~ TABLE OF CONTENTS ABSTRACT LIST OF FIGURES LIST OF TABLES . INTRODUCTION DESCRIPTION OF ASSEMBLY AND MATERIALS CA-10 Assembly Loading Control Rod Calibrations Temperature Effects Neutron Flux and Power Distribution Danger Coefficient Measurements =11 Assenmbly Loading Evaluation of Control Rod D Neutron Flux Distribution Power Distributions Neutron lLeakage Measurements Evaluation of Boral Around Fuel End Ducts Stainless Steel and Boron Poison Rods Cadmium Importance Function Danger Coefficients Effect of Stainless Steel Around Fuel Annulus GHHbOQEEBEOQE > g HOawr CA-12 CA-13 A Assembly Loading B RNeutron Flux Distribution C Power Distributions CA=14 A Critical Loadings B Importance Function of Three Stainless Steel Rods C Thermal Neutron Leskage D Power Distribution APPENDICES A Summry of Materials in Reactor Assemblies B Analyses of Reactor Materials Page vii 1ixX -vii- LIST OF FIGURES Page II-1 Photograph of Two Halves of Aluminum Matrix 3 III-1 Assembly Loading at Mid-plane > ITI-2 Axie)l Loading of Assenbly in Vertical Plane Containing Reactor Axis ITI-3 Photograph of Typical Fuel Shish 7 III-% Calibration Curve for Control Rod D 9 III-5 Average Sensitivities of Control Rod D 10 IIT-6 Change in Reactivity vs Temperature 12 III-7 Bare and Cadmium Covered Indium Traverses Through Fuel Core 13 III-8 Bare and Cadmium Covered lndium Traverses Parallel to Axis Through Cell 0-12 14 ITI-9 Bare and Cadmium Covered Indium Traverses Parallel to Axis Through Beryllium Island and Fuel 15 IIT-10 Indium Praverse Radially in Mid-plane of Reactor, Cell 0-12 to X-12 18 I1I-11 Bare and Cadmium Covere d Indium Traverses Across Fuel Annulus 19 III-12 Power Distribution in Cell Q-13 21 II1-13 Power Distribution Awross Fuel Annulus 23 ITI-14 Power Distribution Through 20 Mil Uranium Disk 2k III-15 Effect of Stainless Steel on Power Distribution 27 ITI-16 Toss in Reactivity Due to Addition of Stainless Steel Core Shells 28 IV-1 Assembly Loading at Mid-plane 31 IV-2 Axial loading of Assembly in Vertical Plane Containing Reactor Axis o o 32 IV-3 Fuel Tube Arrangement 33 V-5 V-6 Iv-7 -8 Iv-9 IV-10 Iv-11 Iv-12 vi-2 VI-3 VI-b vI-5 VII-1 VII-2 VII-3 VII-4 VII-5 -viij- Calibration Curve for Control Rod D Radial Flux Distraibution at the Mid-plane o Rad1al Flux Distribution 6-1/4" From Mid-plane Cadmium Fractions in CA-10 and CA-ll Position of Boral Inserted Around End Ducts Axial Power Distributions in Call N-1k Power Distribution Across Fuel in Mid-plane Reactivity Loss due to Poison Rods Inserted in 0-12 Cadmium Importance Function Assenbly Loading at Mid-plane Assenbly Loading at Mid-plane Neutron Flux Distribution in Mid-plane . Assembly Loading at Mid-plane Radisl FNeutron Flux Distribution in Mid-plane Radial Power Distribution in Mid-plane . Axial Power Distribution Through M-12 Radial Fuel Cadmium Fractions Assembly Loading at Mid-plane Axial Assembly Loading in Vertical Plane Containing Reactor Axais ( v Position of Nickel Sheets Thermel Flux Leakage vs. Boral Thickness . Power Distribution Across Fuel Annulus at Mid-plane Page 35 36 . 37 39 40 41 43 46 b7 52 53 Sh o7 58 60 61 62 64 65 67 69 70 LIST OF TABLES Page III-1. Summery of Calibration Data for Rod D . 8 ITI-2 Summary of Calibration Data for Rods B and C 8 III-3 FNeutron Flux Traverses in Cell Q-13 16 IIT-4 Neutron Flux Traverses Through Cells 0-12 and P-12 16 II1-5 Radial Flux Traverse in Mid-plane o 17 III-6 Neutron Flux Distribution Across Fuel Annulus 20 I1T-7 ©Power Distribution in Cell Q-13 20 III-8 Power Distribution Across Fuel Annulus 22 ITI-9 Power Distribution Through 0 02" Fuel Disk 25 III-10 TFuel Cadmium Fractions for Cell Q=13 25 III-11 Effect of Stainless Steel on Power Distribution 26 IV-1 Rod D Calibration . . . 3k IV-2 Redial Flux Distribution in Mid-plane 3k IV-3. Radisel Flux Distribution 6-1/4" from Mid-plane 38 Iv-k Axial Power Distributions in N-1k 42 IV-5 Power Distribution Across Fuel in Mid-plane L2 Iv-6. Sumary of Neutron Leakage Measurements Wy IV-7T Reactivity Loss Due to Poison Rods Inserted in 0-12 45 Iv-8 Cadmium Importance Function . 48 IV~9 Danger Coefficient Measurements . 49 V-1 Radial Neutron Flux Traverse in Mid-plane 55 VI-l. Radial Flux Distribution in Mid-plane 56 VI-2 Radial Power Distribution in Mid-plane 59 VI-3. Axiel Power Distribution Through M-12 29 VI-k. Fuel Cadmium Fractions in the Mid-plane . 63 -X - Page VIiI-1 JImportance Function of Three Stainless Steel Rods 68 ViI-2 Thermal Neutron Leakage T1 ViI-3 ©Power Distribution Across Fuel Annulus at Mid-plane 71 I INTRODUCTION The Reflector Moderated Reactor (RMR) as presently conceived is a high power, high temperature, circulating fuel reactor, which 1is charscterized by a central beryllium i1sland surrounded by a uranium bearing fluoride fuel annulus and a beryllium reflectorl Faive critical sssemblies have been constructed in which the essential features of the RMR were mocked-up The series of experiments, which were carried out at room tempersture and essentially zero power, provide experimental data on the criticel mass snd neutron flux and power distri- butions that cen be compared to those predicted by multigroup calculation for the assembly in question In addition, danger coefficients were obtained for a large number of materials placed at various points in the reactors The first assembly (CA-10) consisted essentially of a central 12" cube of beryllium partially surrounded by a 3" fuel annulus and a 12" thick beryllium reflector backed by graphite The fuel annulus contained sodium filled stainless steel cans between which were placed 0 01" thick uranivm disks The majority of the measurements were teken with 15 1 kg of U-235 in the reactor, which provided an excess reactivity of 1 9% in Kepr The second assembly (CA-11) differed from the first in that, 1) the fuel region contained an homogeneous mixture of Zr0Op, NaF, C, and UF} packed 1n aluminum tubes, 2) the thickness of the fuel annulus was increased to 4-1/2", and end fuel ducts were provided, 3) the central island region was arranged essentially as a 9" cube, and 4) the regions were assembled to approximate concentric spheres somewhat more closely This assembly was critical with 7 7 kg of U=235 1 The third assembly (CA-12) was originally designed to evaluste the characteristics of a graphite moderated and reflected assembly The fuel region was built up as a cylindrical shell 21" OD, 3" thick and 36" long The fuel used was similar to that used in CA-11 with the U~-235 concentration approximately doubled This arrsy was not critical Approximately 80% of the graphite in the island and the inner 6" of the reflector was then replaced by beryllium in order to meke the reactor critical The fuel region contalned 17 3 kg of U-235 1In the second modification of this reactor, the 17 3 kg of fuel were rearranged as an ellipsoid with the beryllium and graphite surrounding the central fuel region This two region reactor was not critical e 1 Fraes, A P and Mills, C B , "The Fireball, A Reflector Moderated Circulating - Fuel Reactor" Y-F10-104, June 20, 1952 The fourth assembly (CA-13) was composed of a central ellipsoid contalning a 1 to 1 volume ratio of graphite and the fuel used 1n CA-12 This core was surrounded by a beryllium reflector 12" thick The reactor was critical with 15 2 kg of U-235 The last assenmbly (CA-14) was similar to CA-1l except that the fuel with the higher uranium concentration was used and an effort was made to simulate the poisons (core shells, coolants, coolant tubes, etc ) that will be present in a full scale reactor It was critical with 18 2 kg of U-235 ITI DESCRIPTION OF ASSEMBLY AND MATERIALS The critical experiment assembly has been described in other reports, and only a brief description of the components that are necessary for an wmderstanding of the physical make up of the reactors will be given The assembly apparatus consists basically of a matrix of square 2S aluminum tubes, which form a 6' cube and into which may be placed the reactor meterials The 6' cube 18 divided into two identical halves, one of which is stationary and the other movable by remote control Each half consists of 576 tubes, stacked in a 24 x 24k cell array These tubes are 36" in length with a 3" x 3" outside cross-section and have O O47" thick walls Part of the reactor materials are placed 1n each half and the assembly made critical by control rod adjustment after the two halves have been brought together A reference system 1s used to designate each cell A letter desig- nation is given each column of cells and a number for each row This system 18 evident from any of the loading diagrams The assembly, with the two halves separated, can be seen in Fig II-1 (The materials visible in the upper portion of the movable half have no relation to the subject experiments ) The basic materials present in the reactors beside the fuel and aluminum matrix were beryllium and graphite These materisls were in the form of square blocks, either 2-7/8" x 2-7/8" or 1-7/16" x 1-7/16" and of various thicknesses Each block had a 0 196" hole drilled through its center, normal to the square cross-section When required, ,a skewer rod was inserted through these holes to form a "shish-kabob™ The skewer rods had a 3/16" diameter and were composed of aluminum excépt for those "shishes” used as safety or control rods in which case a steel skewer was used All safety and control rods were normal reflector or island shishes 1 Bly, F T et al, NEPA Critical Experiment Facility, NEPA-1769, April 15, 1951 Fig. ll-1 Table Assembly. B Tfi%\éfi§15h354(93*§% U-235) uranium metal disks used were approximately O O1" thick with diameters of 2 860" and 1 430" and weighed 18 0 £+ 0 1 and 4 5 + 0 1 grams respectively The weight tolerance was obtained by punching small holes in some of the disks Each disk had a 0 196" hole drilled through the center The powdered fuel mixtures are described under the appropriate reactor designation The appendix contains quantitative information on the materials located in each region of all the assemblies, and also spectrographic analyses of some of the materials used IITI CA-10 A Assembly Loading The initial loading for CA=10 1s shown in Figs III=l and III-2, which are cross-sections of the assembly at the mid-=plane of the reactor and the wvertical plane containing the central axis of the reactor, respectively It is seen that the assembly consists of a 12" central cube of beryllium, partially surrounded by a 3" fuel annulus composed of sodium filled stainless steel cans (2-7/8" x 2-T/8" x 1") between which were placed the large sized O OL" thick uranium disks The components and arrangement of a typical fuel shish are shown in Filg III-3 As can be noted from Fig III-2, a section 6" square at both ends of the fuel amnulus was filled with beryllium rather than the uranium disks and sodium The fuel region was loaded with two fuel disks between each sodium filled can except for the 3" end fuel regions behind the central beryllium cube, which had only one disk between each can The fuel annulus was surrounded by a 12" thick beryllium reflector backed by 6" of graphite on the sides and 8" on the ends of the reactor With the loading described above the reactor contained 15 0 kg of U=235 which at a room temperature of 7h°F provided an excess reactivity of 1 9% in k.py The reactor was made gritical by removing the reflector from cells J=13 and 14 and I-13 (1 &%) and by adjustment of the control rods Measurements that required more reactivity than was available on the control rods were accomplished by imserting one or more reflector shishes in the above cells All foll and danger coefficient measurements were taken on the opposite side of the reactdr where the effect of the voids on the measurements was negligible B Control Rod Calibrations The change in reactivity introduced by the linear displacement of each of three control rods was determined Rod D was cali- ORNL—LR—DWG 2562 J K L M N 0 P Q R S T U \ W H Fig III-1 Assembly Loading at Mid-Plane L IIIMITIMIINNN NN\ 7NN NN\ 7777 NN N7 NN N\ 0077 N\GZHENNNNNGHZHN wx/ N3 N 32, NV AN AN NN 2N 2NN NNVZNCLGGNZZZNN NN ZANNNINZZ AN NN NN \\ 227N\ N\ 7722\ NN NN Mk . MmN 444444444444444444 ~r— t A ORNIH—DG 2563 . RN RN RN R R, BN NN A AN NN RN N NN\ NN D NN NN 00 NN NN Ay o NN NN N e NSANARNAANS Wy NN NN SN 22NN NN o o 7777 A P S UYL )R J i Fig IlI-2 Axial Loading of Assembly in Vertical Plane Containin g Reactor Axis Fig. IlI=3. Typical Fuel Shish. -8- o Y brated by the integrated period method after Whl%h rods B and ; C were evalunted by direct compmrason with Rod D°. Detailed < data for Rod D are given in Table III~1 A graph of reactivity ! of this rod in cents, versus its linemr displacement i1s shown in Fig III-k, The average change in resctivity per wat dis- placement for each inerement of displacement is plotted for the same rod in Fig. III-5, Siftilayr data for rods B and C are shown in Teble III-2 The delayed neutron fraction has been taken as O 0073 TABLE IIT - 1 SUMMARY OF CALITBRATION DATA FOR ROD D Pogition of Rod Total Change Average Incremental From Mid-Plane 1n Resctivity Sensitivaity in Cents in Inches o Value in Cents per Inch 00 0.0 - 2+555 2 01 0 79 7« Blils 8,98 1 %2 104245 14,62 2435 11 987 19 80 2497 13 560 24 Th 3 14 14 g21 28 98 3 12 16 290 32 50 2.57 17 623 35,02 1.89 19 19% 37 T9 1 76 25 697 42 69 075 TABLE III=-2 SUMMARY CALIERATION DATA FOR RODS B AND C Distance of Rod Reactivity in Cents Average Sensitavity From the Mid-plane in Cents per Inch in Inches Rod B Rod C , Rod B Rod C 00 00 00 - - 50 11 3 35 2 26 0 70 10 O 20,9 12,2 192 1 7h 1540 28 5 28 2 3 20 320 20.0 33,0 39 3 2.22 2 27 25 0 35.0 43,2 0 78 0 78 2 Callihan, D. "Critical Experiments on Direct Cycle Aircraft Reactor" ORNL~1615, October 22, 1953, p 22 REACTIVITY CHANGE (cents) 60 50 40 30 20 10 ORNL-LR-DWG 2564 LINEAR DISPLACEMENT OF ROD (in) Fig II-4 Cahibration Curve For Control Rod D _— il .// ~ o /. S /. / / /O o / / " IN REACTIVITY PER UNIT LENGTH (cents /in } CHANGE n ORNL LR DWG 2565 10 15 20 LINEAR DISPLACEMENT OF ROD {imn) Fig Il 5 Average Sensitivities of Control Rod D 25 30 v D C. ~11- Temperature Effects The changes in reactivity produced by changes in temperature over a limited range were determined by changing the demand temperature of the assembly room Several days were allowed for the reactor temperature to reach equilibrium The reactor was made critical before and after the temperature changes, and the reactivity differences determined in terms of the calibrated control rod settings The temperature of the reactor was determined from two iron-constanteh thermocouples Figure ITI-6 shows the reactivity changes plotted as a function of the assembly temperature A systematic study was not undertaken to determine the sources of this temperature effect, it was determined primerily to correct all reactivity measurements to a common reactor temperature Neutron Flux and Power Distraibution Neutron flux distributions were obtained using bare and cadmium covered indium foils The techniques used for these measurements have been ade%uately covered in previous reports and will not be repeated here Power distributions were obtained using the aluminum catcher foil technique which is also described in the same reference l TPlux Distributions Bare and cadmium covered indium traverses were made through the fuel core, starting at the mid-plane and extending to the edge of the graphite reflector and parallel to the axis of the reactor through cell Q-13 Table III-3 contains data showing the positions of the bare and cadmium covered foils and thelr relstive activities Figure III-T7 shows the relative activities plotted as a function of the distances of the foils from the mid-plane Bare and cadmium covered indium traverses were also made through the beryllium island parallel to the axis of the reactor, starting at the mid-plane and extending to the edge of the graphite reflector, through cell 0-12 and P-12 The traverses made through cell P-12 pass through the fuel core while those through 0-12 do not Table III-4 contains data showing the distances of the foils from the mid-plane in the respective cells and also the relative activity obtained for each fo1l Figures IIT-8 and III-9 show the graph of the 3 TIbid, p 36 IN REACTIVITY {cents) CHANGE 20 20 Lo d ORNL LR DWG 2566 55 65 75 8BS ASSEMBLY TEMPERATURE ( F) Fig Il 6 Change in Reactivity vs Temperature 95 105 _gt_ 28 24 20 16 12 RELATIVE ACTIVITY ORNL LR-DW! 2567 DISTANCE FROM MIDPLANE OF REACTOR (in) Fig III-7 Bare and Cadmium-Covered Indium Traverses Through Fuel Core - FUEL CORE — BERYLLIUM | e GRAPHITE — = m, BARE INDIUM ACTIVATION X N CADMIUM FRACTION CADMIUM COVERED ] T T~ | . ./ INDIUM ACTIVATION ..I ™ T e —— o ° 0/! \ \ —"-—--._————_ . 0 5 10 15 20 25 30 10 05 CADMIUM FRACTION _-bl_ RELATIVE ACTIVITY 45 40 35 30 25 20 10 ORNL-LR-DWG 2568 -BARE INDIUM ACTIVATION Fig lI-8 Bare and Cadmium Covered \\ o CADMIUM-COVERED INDIUM ACTIVATION-/ \\\( | - —— — 10 15 20 25 30 . A DISTANCE FROM MIDPLANE OF REACTOR {(in) Indium Troverses Parallel to Axis Through Cell O0—-12 05 CADMIUM FRACTION _gt_ RELATIVE ACTIVITY 32 28 24 20 16 12 ORNL LR DWG 2569 BERYLLIUM _ | | FUEL _ | _ -l {SLAND T AVERTTT BERYLLIUM GRAPHITE —— BARE INDIUM ACTIVATION o——@ "\ CADMIUM FRACTION . —— N\ " . \ / | CADMIUM-COVERED INDIUM ACTIVATION //\.\ \ 0 5 10 15 20 25 30 DISTANCE FROM MIDPLANE OF REACTOR (in) Fig III—9 Bare and Cadmium-Covered Indium Traverses Parallel To Axis Through Beryllium Island and Fuel 05 CADMIUM FRACTION -16- relative acticity of the various foils snd the corresponding cd fraction®* as a function of their distances from the mid-plane of the reactor for the traverses through cells 0-12 and P-l2 respectively TABLE III-3 Neutron Flux Traverses in Q-13 Distance of Foll from Relative Activity Mid-plane in Inches Bare Foils Cadmium Covered Foils 00 10 338 8 08L 05 9 118 7 889 15 9 396 T 959 25 8 899 7T 663 L5 8 302 T 393 65 8 068 6 910 85 8 532 7 160 95 9 533 T 370 10 0 13 370 8 569 11 0 17 815 8 580 15 0 15 502 4 578 Z1L 5 182 0 46k 23 0 2 97k 0 341 28 0 o0 767 0 078 TABLE III-4 Neutranr Flux Traverses Through Cells 0«12 apd P 5 2563 75 2253 95 2806 95 5318 _LZ_ RELATIVE FISSION RATE < TN CADMIUM COVERED INDIUM ACTIVATION-"] H\Q 0 5 10 15 20 25 30 RADIAL DISTANCE FROM REACTOR AXIS (in) Fig IV 5 Radial Flux Distnbution at the Mid plane 05 CADMIUM FRACTION RELATIVE ACTIVITY ORNL LR DWG 2581 ~————— BERYLLIUM DISKS - BERYLLIUM ———A = - A~ T —————_ BARE INDIUM ACTIVATION CADMIUM COVERED INDIUM ACTIVATION _— _ Kcnowum FRACTION \__________/ 4 6 RADIAL DISTANCE FROM AXIS ( ) 8 10 Fig IV 6 Radial Flux Distribution 6% in from Mid plane 05 CADMIUM FRACTION -38- TABLE IV~3 Radiel Flux Distribution 6-1/4" from Mid-Plane Distance from Axis Relative Indium Activity in Inches Bare Cd Covered o7 Te523 5 LaL 23 6.579 L 679 33 L, 563 3 589 L 3 L 1k 3 292 4 9 3 882 3 27h 5T 3 949 3.202 63 3 953 5 120 71 L 366 3 364 76 L 866 3 346 87 6 294 3 634 97 T 229 3 532 113 T.389 3 076 The associated radiasl cadmium fractions for indium obtained in the mid-plane from CA~10 and CA-1ll are shown in Figure IV=-T. D Power Distributions Power distributions were obtained by placing O 002" thick quarter size uranium disks wlth sluminum catcher foils adjacent to fuel tubes in the deslred location. The first was measured parallel to the reactor axis in cell N-1L The 0.002" thick uranium disks gnd associated catcher folls were placed adjacent to the fuel tube between the fuel tubes and the beryllium island. A second power distribution wes taken i1n the same location with Boral#® strips (9" x 2-7/8" x 1/4") placed sround the end fuel ducts an the stationsxy table, replacing some beryllium and graphite as shown in Figure IV~8 The data have been arbitrarily normalized at a point 14 5" from the interface In order to maintain the reactor craitical with the Boral inserted, additiomal 0.01" fuel disks were added to the beryllium island on the movable table The data obtained from both rums is shown in Table IV-4 and Figure IV-9 % The Boral wis prepared fram z mixture of 35 wt % BEE and slumimam, This mixture is held between two aluminum sheets, each about 0.04" thick forming a 1/L" sandwich comtaining 250 mg of borom per squsre centimeter. The boral strips wereé wrapped in tape to minimize FPawen contamination. CADMIUM FRACTION to 05 ORNL—LR-DWG 2582 r=— CA—-1{0— BERYLLIU | | | M—————-J'--—FU -‘—CA—H—BERYLLIUM—’J‘—- FU | | | | EL BERYLL UM ——————— e EL——T-(——— BERYLLIUM—————t—— DISTANCE FROM CENTER OF FUEL REGION {(in) Fig IV=7 Cadmium Fractions in CA—10 and CA—44 o w— Y A ‘é)\< CA—1O\ cA—1 6 4 0 2 4 6 8 10 12 14 16 18 et NN \\\\\\ \\\W TVl s i NUNIINUINIUNIN \\ GRAPHITE BERYLLIUM FUEL BORAL v ol VA Y N7 / N% N \// NN L /% 11 12 13 NN \\\\7///////\\\\\ 14 Fig IV-8 Position of Boral Inserted Around End Ducts on Stationary Table _40_ RELATIVE ACTIVITY ORNL—LR-DWG 2584 BERYLLIUM REFLECTOR FUEL ‘ } | l ‘ | BERYLLIUM ISLAND R = == = — = — 7 E+ ' \/TRAVERSE WITHOUT BORAL AROUND END DUCTS N TRAVERSE WITH BORAL ARQUND END DUCTS —A— 0 10 15 20 25 30 35 DISTANCE FROM INTERFACE (in) Fig IV—-9 Axial Power Distributions in Cell N—1{4 Axial Power Distribution in N-1k ~L4o. TABLE IV-L Distance from Interface Relative Actaivity* in Inches Without Borsal With Boral 14 5 6 902 6 902 17 3 5 896 5 647 20 O 6 312 5 469 22 7 6 027 L hés 25 5 5 205 1 160 28 3 3 W7 0 248 31 0 2 602 0 098 A power distribution was also measured across the fuel annulus in the mid-plane 6f the reactor The 0 002" thick wranium disks were placed in the horizontel gep (approximately 1/4") between the fuel tubes of adjacent vertical cells, with eight 5/16" catcher foils distributed across the fuel disks The data obtained is shown in Table IV=5 and Figure IV-10 TABLE IV~ Power Distribution Across Fuel in Mid-Plane Distance From Reactor Axls O3\ & £ oW oW o\ M Relative Activity 6 955 5 021 L 213 3 163 3 500 3 880 5 17k T 549 ¥ Dats normalized at the point 1% 5" from the interface RELATIVE ACTIVITY 10 ORNL —-LR—-DWG 2585 I | | i | | | | g L | l VOID- |t VOID—=| }-— | o BERYLLIUM BERYLLIUM ISLAND > I FUEL > REFLECTOR | | / \M 2 3 4 5 6 9 10 DISTANCE FROM CENTER OF ISLAND (in) Fig IV—-10 Power Distribution Across Fuel in Mid-plane E Neutron leakage Meagsurements Measurements were obtained in an effort to estimate the ratio of the end to side neutron leakage from the reactor In the first set of measurements six bare and cadmium covered indium foils were placed on the back swrface of the reactor extending from the reactor axis outward 7" Three sets of folls were placed on the outer swrface of the graphite side reflector at the mid-plane These foil measurements were taken without any Boral sroumd the fuel end ducts Side and end fast neutron leakage meaSurements were also cbtained using a fast fisgsion chamber Measurements were taken both with and without Boral around the fuel end ducts, and with a 1/4" of Boral around the fission chamber¥#. Slde leskage was determined by placing the chamber in cell X-12 with the center of the chamber gt the mid-plane The axls of the fisslon chamber was approximately 1-1/2" from the graphite side reflector The end leskage wWas obtained by placing the chamber approximately 7" behind the station- ary table with its axis parallel to the mid-plane with the center of the chamber on the projected exig of the reasctor The dats obtained are sumarized in Table IV-6 TABLE IV-6 Summaxy of Neutron leskage Measurements Ratic of End to Side Ratio of End to Side Teaksge With Boral Ieekage Without Boral Arommd End Ducts Arocund End Ducts Fission Chamber 6 28 Bare Indium = 23 Cd Covered In = 12 Bare-0d Covered - 11 % Slebs of Boral, 18" x 2-7/8" x 1/4", were placed around the chamber, which was 10" lomg, thus approximstely 4" of Boral extended beyond either end. -45- F Evaluation of Boral around Fuel End Ducts The Boral sreets that were placed around the fusl end ducts as shown in Figure IV-7 were evaluated The gain in reactivity due to their removal was compensated for by removing fusl disks from +he 1sland region on the movable table The loss in reactivity due to the replacement of beryllium and graphite by Boral around the fuel end ducts amounted to 1 &% in k oo G Stainless Steel and Boron Poison Rods Two 3/16" diameter rods were evaluated in the center of cell 0-12 1n the 1sland The first rod was a stainless steel tube 40" long filled with natural boron to a linear density of O 1 gm B/in The second was a stainless steel rod 38' long The loss in reactivity in cents for each rod i1s shown as a function of the distance of the end of the rod from the mid-plane in Table IV-T and Figure IV-1ll TABLE IV-7 Reactivity Loss Due to Poison Rods Inserted in 0O-12 Distance From Mid-Plane Loss in Resctivity in Cents in Inches Boron Filled Rod*¥ Stainless Steel Rod 00 W 6 50 L5 29 9 30 90 17 1 - 15 C L 8 - 21 0 0k - H Cadmium Importance Function The importance of cadmium as a neutron absorber when placed in the mid-plane of the reactor was obtained by taping a piece of cadmium 2=-3/L" square and O 02" thick weighing 19 09 gm, to either s block of beryllium or a fuel tube, and recording the loss in reactivity for each radial position The data obtained are shown in Table IV-8 and Figure IV-12 ¥Further insertion of the Boron filled rod from the mid-plane to the assenmbly interface (9") resulted in an additional reactivity loss of 30 2 cents _9-b_ LOSS IN REACTIVITY (cents) [\ o 75 18 O ORNL LR DOWG 2586 DISTANCE FROM MIDPLANE (in) Fig IV-11 Reactivity Loss Due to Poison Rods Inserted in O-12 / BORON-FILLED ROD / STAINLESS STEEL ROD 0 5 10 15 20 25 75 LOSS IN REACTIVITY (cents) o o N 14! ORNL-LR —DEG 2587 T ———— BERYLLIUM —-—| ORNL—-LR—DWG 2591 b E F G H J K L M N O P Q@ R s T NUMBERS IN REFLECTOR 1313 |13 113 113 SQUARES INDICATE NUMBER OF INCHES OF Be IN CELL 13 | 13 | 18 |18 | 18 |18 | 18 | 13 |13 BEHIND WHICH IS ENOUGH GRAPHITE TO FILL 301n TA 13| 18 | 18 | 18 | 24 | 24 | 24 |18 | 18 | 18 | 13 OF THE TABLE 13 | 1318 | 24| 36|27 |36 |27 |27 |24 |18 |13 |13 13 | 18 | 24 | 24 36 | 24 | 24 | 18 | 13 13 | 18 | 18 | 24 | 21 21 | 24 |18 [ 18 | 13 HoLE | 13|18 | 24| 27 27 | 24 |18 | 13 T I & T4 & | | 1318|2427 27 |24 |18 | 13 w o c | € o |3 | 13]18|24]27 27 | 24 |18 | 13 13| 18 | 18 | 24 | 21 21 124 |18 |18 | 13 13 | 18 | 24 | 24 |2f ST 24| 24 18 |13 13 | 13| 18| 24 27 |27 |27 |27 36 |24 |18 | 13 | 13 13| 18 | 18 | 18 |24 |36 |24 | 18 | 18 | 18 | 13 9in FUEL TUBE + 21 in Be 13 | 13| 18 |18 | 18 | 18 | 18 | 13 | 13 o FUEL TUBE | + 5in GRAPHITE GRAPHITE (SAME LENGTH AS LONGEST FUEL TUBE IN THE SAME CELL) Fig VI—1 Assembly Loading at Mid-plane 86 RELATIVE ACTIVITY ORNL LR DWG 2592 BERYLLIUM A | | — e GRAPHITE ! | _ ey I BARE INDIUM ———_ | / fi/ - \ CADMIUM FRACTION 05 .-—-—'. / / CADMIUM COVERED INDIUM S~ wn Q 15 RADIAL DISTANCE FROM AXIS { } 20 25 Fig VI 2 Radial Neutron Ftux Distribution in Mid plane 30 CADMIUM FRACTION -59- ¢ Power Distribution Power distributions were taken radially in the mid-plare from M-12 to P-12 and axially back through M-12 The traverses were cbtained by placing & 0.002" uranium fuel disk and its associated aluminum catcher foil adjacent to a fuel tube The data obtained for both traverses are shown in Tables VI-Z and VI-3 and Figures VI-3 and VI=k TABLE VI-2_ Radial Power Distribution in Mid-Plane Distance From Axis Relative in Inches Activaty 00 9 66 30 9 9k 60 13 09 9.0 2l 23 10 3 Ly 14 TABLE VI=3 Axial Power Distribution Through M-12 Distance From Mid-Plene Relative in Inches Activity 10 9 66 51 8 95 90 9 18 13 0 10 3k 17 O 25 31 Three uranium foils were covered with O 020" of cadmium and placed in radial positions in the mid-pleane The fuel cadmium fraction, as determined from the activity of these foils together with the activities of similarly placed foils for the radial power distribution, are plotted as a function of the square of the radial distance from the reactor axis 1in Figure VI-5 Teble VI-lb shows the data from which the curve was constructed This gives an average of very nearly 50% thermal fissions in the mid-plane of the reactor. 09 RELATIVE ACTIVITY OR LR DWG 2593 50 40 30 20 40( N 4 G RADIAL DISTANCE FROM AXIS ( ) Fg VI 3 Radal Power Dstribution in Mid plane ¥9 RELATIVE ACTIVITY 50 40 30 20 ORNL DWG 2594 6 8 10 DISTANCE FROM MID PLANE { ) Fig VI 4 Axial Power Distribution Through M 42 29 FUEL CADMIUM FRACTION R D!I! 2595 o8 AVERAGE =50/ o6 020 20 40 60 80 (DISTANCE FROM aXIS ( )2 Fg VI 5 Radial Fuel Cadmium F actions 120 -63- TABLE VI-4 Fuel Cadmivm Fractions in the Mad-Plane Distance From Cadmium Axis in Inches Fraction 00 0 200 60 0 399 10 3 0 813 ViI CA=-14 A Criticel Loadings The final assembly was bullt up with e geometry similar to that used in CA-1l1l The aim of the experiment was to insert materiels in the reactor that would simulate the structural polsons that would be present in & full scale reactor in the power range of 50 to 200 megawatts ‘The assembly was loaded as shown In Figures VII-1 and VII-2 and conteined 18 2 kg of U-235 in the 6.42 gnm U-235/cc fuel mixture described in Section V In general the reactor was maintained criticel with two dlfferent sets of poilsons The first set of polsons included 1 Boral sheets (9" x 2-7/8" x 1/4") which surrounded the fuel end ducts on the stationary table Thas Boral extended from the end of the assembly to a point 9" from the central fuel region, thus 4-1/2" of the end ducts were covered Boral sheets (1-1/4" x 2-7/8" x 1/4") were placed around the fuel end ducts on the moveble table These sheets extended from the end of the fuel ducts to a point 7-3/4" from the central fuel region 2 Stainless steel sheets (0 146" thick) covered 96% of the inside fuel surface The outslide fuel surface was uncovered from the 1nterface of the assembly back 4-1/2" on the stationsry teble The fuel surface exposed at the assenmbly interface and the corresponding surface at the opposite end of the central fuel region were left uncovered The remaining outside fuel surface was 96% covered with O 118" thick stainless steel sheets The total weight of the stainless steel was 25 5 kg _bg_ oD\~ L1 \\i\\ SN . SN o @//////%%///// ////Z// . / Fig VII-1 Assembly Loading at Mid-plane (No Structure Shown) _99_ GRAPHITE BERYLLIUM NICKEL SHEET NICKEL SHEET NICKEL SHEET NICKEL SHEET NICKEL SHEET MOVABLE TABLE STATIONARY TABLE Fig VII-2 Axial Assembly Loading in Vertical Plane Containing Reactor Axis L ORNL-LR DWG 2597 -66- 3 TNickel sheets 2-7/8" wide and 0 01¥ thick were in the positions indicated in Figures VII=-2 ahd VII-3 In addition to the above poisons, 177 1n5 of beryllium were added to the fuel region Of this total, 150 inJ were distributed symmetrically throughout the outer three fuel layers frq% 0 to 18 inches back on the stationary teble An additional 6 in” were distributed symmetrically throughout the inner 4-1/2" long fuel layer adjacent to the assembly interface on the stationary table The remaining 24 inJ of beryllium were symmetrically distributed throughout the fuel region on the movable table During this series of experiments reactivity wvalues were obtained for one sheet of nickel (24" x 2-7/8" x 0 010") placed in a horizontal position across the top of cell K-12, and for the extension of the reflector from 9" to 32" of beryllium on the sides of the reactor The insertion of the nickel sheet resulted in a loss in reactivity of 8 cents The replacement of graphite by beryllimm in cells G-10 through 15 and cells V-10 through 15 and the addition of graphite to cells E-1C through 15 and X-10 through 15 resulted in a gain in reactivity of 9 6 cents The second set of poison distributions involved the following changes 1 The entire fuel surface was 95% covered by 27 8 kg of stainless steel 2 All the Boral around the end ducts was removed 3 The rest of the assembly was as described previously The remaining measurements described were obtained with the reactor containing this second set of poison distributions Importance Function of Three Stainless Steel Rods The loss in reactivity due to three 316 stainless steel rods vas determined at three radial positions in the reflector and at one position in the island The rods were 3/16" in diameter, and 36" long and each weighed 132 gn They were inserted as skewers in the various cells and extended through the entire shish on the stationary table The date obtained are shown in Table VII-1 10 114 12 13 14 15 ORNL-LR-DWG 2598 —— NICKEL = SHEET NICKEL SHEET —] Fig VII-3 Position of Nickel Sheets on the Stationary and Movable Tables -67- -68- TABLE VII-1 Tmportance Function of Three Stainless Steel Rods Average Distance of Rods Loss 1n Reactivity From Axis in Inches in Cents 23 23 1 10 8 17 6 13 7 12 O 16 7 60 C Thermal Neutron lLeakage Thermal neutron leakage was determined at the end and side of the reactor, both with and without Boral shielding sround the end fuel ducts These meagurements were taken with 0, 1/8", 1/4", and 3/8" thicknesses of Boral swrrounding the U-235 fission chamber that was used The side leakage was measured with the U-2%5 fission chamber located in cell D-12 The geometrical center of the chamber rested 12" back of the interface on the stationary table or 3" back of the mid-plane of the reactor The axlis of the chamber was approximately 1-1/2" from the edge of the graphite side reflector and was parallel to the reactor axis The Boral layers surrounding the chamber were in the form of concentric cylindrical sleeves that extended from the interface of the stationary table back 24" The inside diameter of the inner most layer was 2", and the outside diameter of the outer most layer, using 3 layers, was about 2-3/4" The ends of the Boral sleeves were not capped The end leakage was determined with the fission chamber placed 5—1/2" back of the end graphite reflector The axis of the chamber was parallel to the floor and perpendicular to the axis of the reactor, which passed through the geometrical center of the chanmber The data obtained are shown in Table VII-2 and Figure VII-% All measurements are shown relative to the side leakage without any Boral shielding around the chamber or in the reactor D Power Distribution A radial power distribution was taken in the mid-plane across the fuel annulus (cells P-12 and Q-12) The traverse was obtained by Placing uranium and aluminum foils between the vertically adjacent fuel tubes in the cells The data obtained are shown in Table VII-3 and Figure VII-5 THERMAL FLUX (arbitrary units) 01 0 Ot O o1 0 0001 S ORNL-LR—0WG 2599 BACK LEAKAGE BACK (WITH BORAL IN TABLE) SIDE LEAKAGE Ya s ¥ % BORAL THICKNESS (in) Fig VII—4 Thermal Flux Leakage vs Boral Thickness 5/8 L ORNL-LR—-DWG 2600 N o —0)— RELATIVE ACTIVITY | I i | | \ | | 1 O O 0\—/ e VOID VOID - BERYLLIUM ==J| =< FUEL TUBES ——————| |~————BERYLLIUM————3= | | | | | ] . | | | || | 2 3 4 5 6 7 8 9 10 T 12 DISTANCE FROM AXIS (in) Fig VII-5 Power Distribution across Fuel Annulus at Mid-plane -T1- TABLE ViI-2 Thermal Neutron Leakage Layers of 1/8" Boral¥* Counter Boral Around Relative Arownd Counter Position End Ducts Actlvity 0 Side No 1 000 1 Side No 0 00635 2 Side No 0 00103 3 Side No 0 000676 0 End No o 777 1 End No 0 0466 2 End No 0,0332 3 End No 0 0301 b End No 0,0235 0o End Yes 0 352 1 End Yes 0 0197 2 End Yes 0 0160 3 End Yes 0 0125 4 End Yes 0 00992 %0 161 gm B/cm?/Layer of Boral present TABLE VII-3 Power Distribution Across Fuel Annulus at Mid-plane Distance From Axis in Inches Relative Activity kL 8 25 30 5 2 17 86 5 6 14 24 6.2 10 97 65 12 %0 T1 12 35 TP 12 63 7,8 14 87 81 19 32 8 L 28 3k -T2- APPENDIX "A" Summery of Materiasls in Reactor Assemblies CA-10 Average¥* Mass Volume Mass Volume Density Island kg em3 x 10™° Fraction Fraction gm/cm Beryllium 48 1 26 0 911 918 1 699 2S Al Matrix L 5 16 085 058 0 158 24Ls Al Skewers 02 01 003 002 0 008 Stainless Steel skewer - - - - 0 o2k Void - 06 - 021 - The island contained eleven aluminum and one steel skewer They were placed in the outer island cells Fuel Annulus (Includes beryllium region required to close annulus) Sodium 52 4 52 5 486 T34 0 805 Uranium 16 1 09 149 012 0 265 0 161 304 Stainless Steel Boxes 14 3 18 1%2 025 0 220 28 Al Matrix 11 1 5 4 103 075 0 158 24s Al Skewers 06 02 005 003 0 008 Stainless Steel skewer - - - - 0 024 Al Ceps T4 27 068 038 - Beryllium 60 33 056 oL5 1 699 Void - } 8 - 067 - The lower uranium density reported refers to the fuel region behind the island The higher figure refers to the remainder of the fuel region Reflector (Beryllium) Beryllimm 1521 6 822 5 91k 918 1 699 25 Al Matraix 141 6 52 1 085 058 0 158 24S Al Skewers 09 03 - - 0 008 Stainless Steel skewers 08 01 - - 0 025 Void - 20 5 - 023 - CA-10 = Continued -T3- Average¥ Mass Volume Mass Volume Density kg em’ x 103 Fraction Fraction gm/cm Reflector (Graphite) ] Grephite . ‘1 1673 5 o7k 7 907 898 1 542 2S Al Matrix 171 6 63 1 093 058 0 158 Void - W7 7 - ol - * The average dansities reported are for a unit cell (9"2 cross- sectionsal areg) in which the material was actually present -Th- CA-11 Average Mass Volume Mass Volume Density kg cmd x 1070 Fraction Fraction gm[cm Island (Includes Be Region enclosed by end ducts) —_——— Beryllium 57 1 30 9 912 918 1 699 25 Al matrix 53 20 085 058 0 158 24s Al skewvers 02 01 003 002 0 008 Voad - o7 - 021 - The island contained 16 skewers, which were placed in the outer beryllium shishes (9" length) Fuel Annulus (Includes end ducts) Fuel Mixture Including Usrnium 6L 4 35 0 620 527 - Uranium 8 2 - 079 - 0 123 UF)y, - - - - 0 163 Zr0o - - - - 0 553 NaF - - - - 0 200 C - - - - 0 088 2S Al Matrix 10 5 39 098 058 0 158 Al Fuel Tubes and Spacers 28 6 10 5 275 159 0 443 Stainless Steel Nuts 0 8 01 007 001 - Veoad - 16 9 - 254 - The weight of the fuel mixture includes 360 grams of uranium present as disks Each fuel tube had a stainless steel nut on both ends Reflector (Beryllium) Beryllium 1518 6 820 9 o1 918 1 699 28 Al Matrax 141 3 52 0 085 058 0 158 243 Al Skewers 08 03 - - 0 008 Stainless Steel skewers 08 01 - - 0 024 Void - 20 5 - 023 - Reflector (Graphite) e e Graphite 1626 677 o474 906 898 1 542 2S Al Matrix 166 831 61 3 093 058 0 158 -75- CA-12 Variation I (Graphite island and reflector) Average Mass olume Mass Volume Density kg em”? x 10°2 Fraction Fraction gm/cm) Island (Includes region enclosed by fuel annulus) Graphite 180 1 10k 9 903 898 1 542 2S5 Al Matrix 18 5 6 8 093 058 0 158 Stainless Steel skewers 0 8 01l 00k 001 0 024 Void - 50 - oL3 - Fuel Annulus Fuel Mixture 75 3 40 5 622 545 - (Includang Uranium) Uranium 18 5 - 153 - 0 249 UF) - - - - 0 329 Zr02 - - - - 0 hh9 NaF - - - - 0 163 C - - - - 0 072 Al Fuel Tubes 3% 0 121 273 163 0 M3 2S5 Al Matrix 11 8 L 3 Q97 058 0 158 Stainless Steel Nuts 10 01 008 002 - Void - - - 232 - Reflector ettt Grephite 6738 5 3965 1 906 900 1 542 28 Al Matrix 696 8 256 2 09k 058 0 158 Stainless Steel Skewers 08 01 - - 0 024 Void - 185 &4 - ok2 - The outer 128 cells were filled with AGHT Graphite in all CA-12 configurations All other graphite is AGOT Variation ITI (Combination graph:ite and beryllium 18land and reflector) Island Beryllium 82 2 Ll L 515 L72 1 699 Graphite 61 8 36 1 387 383 1 542 28 Al Matrix 14 g 55 093 058 0 158 Stainless Steel skewers o8 01 005 001 0.024 Void - 80 - 085 - -T6- CA-12 - continued Average Mass Yolume Mass Volunme Density kg em® x 10°7 TFraction Fraction gm/mm3 Fuel Annulus Fuel Mixture 82 0 bWy 1 608 415 - Including Uranium Uranium 19 6 - 146 - - Al Fuel Tubes 35 0 12 9 260 121 0 443 25 Al Matrix 16 8 6 2 124 058 0 158 gtainless Steel Nuts 11 01 008 001 Void - 42 9 - Lol Reflector (Be and Graphite) Beryllium 681 1 368 2 470 450 1 699 Graphite 639 T 372 6 hh] 456 1 542 2S Al Matrix 129 3 W7 5 089 058 0 158 Void - 29 3 - 036 - Reflector (Graphite - includes only the region outside the Be reflector) Graphite 5674 k4 3345 6 906 900 1 542 2S Al Matrix 587 7 216 1 0%k 058 0 158 Void - 154 9 - oh2 - Variation III - (Pwo Region) Core Fuel Mixture 75 3 40 5 622 545 - Including Uranium Uranivm 18 5 - 153 - - Al Fuel Tubes 33 0 12 1 273 163 0 443 Al Matrix 11 8 4 3 097 058 0 158 Stainless Steel Nuts 10 01 008 002 - Vold - 19 % - 232 - Reflector (Be + Graphite) Beryllium 559 3 302 3 676 662 1 699 Al Matrix 72 2 26 5 087 058 0 158 Graphite 196 5 11k 4 237 251 1 542 Void - 13 3 - 029 - CA-12 - Continued -T7- Average Volume Mass Volume Density cm> x 1073 Fraction TFraction g;m./cm3 Mass kg Reflector (Graphite) Graphite 6362 0 Al Matrix 658 2 Void - CA-13 Core Tuel Mixture ™ 3 Including Uranium Uranium 18 5 UFY - Zr02 - NaF - C - Graphite 120 7 Al Fuel Tubes 33 0 Al Metrix 2L 0 Stainless Steel Wuts 10 Void - Reflector (Beryllium) o Beryllium 1654 7 Al Matrix 154 0 Al Skewers o4 Stainless Steel skewers 13 Void - Reflector (Graphite) Graphite 2071 4 Al Matrix 212 L Void - 3Thé6 1 906 900 1 542 2hk2 0 094 058 0 158 174k 5 - oh2 - 4o 5 297 266 - - 073 - - - - - 0 164 - - - 0 224 - - - 0 081 - - - 0 036 70 3 L75 L63 0 771 12 1 130 080 0 221 8 8 094 058 0 158 01 ook 001 - 20 0 - 131 - 8oL L o1k 918 1 699 56 6 085 058 0 158 02 - - 0 008 02 001 - O 024 22 4 - 023 - 1206 4 Q07 898 1 542 78 1 03L 058 0 158 59 0 - oll - ~78- CA-1k4 Island (Includes Be Region enclosed by End Ducts) Average Mass Volume Mass Volume Density kg cmd x 103 Fraction TFraction gm/cm Beryllium 57 1 30 9 905 918 1 699 Al Matrix 5 3 20 084 058 0 158 Nickel 05 01 008 002 0 028 Al Skewers 02 01l 003 002 0 008 Void - 0T - 020 - Fuel Annulus (Including End Ducts) Fuel Mixture 81 3 4z 7 596 Sll - Including Uranium Uranium 19 6 - 143 - 0 249 Al Fuel Tubes 36 0 13 2 264 165 0 k43 Aluminum Matrix 12 7 Y 7 093 058 0 158 Beryllium 5 4 29 039 ,036 - Stainless Steel Tuts 11 01 008 002 - Void - 157 - 195 - Reflector (Beryllium) Beryllium 1559 0 82 7 913 918 1 699 Al Matrix 145 1 53 3 085 058 0 158 Nickel 18 2 001 0002 0 028 Stainless Steel Skewers 0 8 1 000k 0001 0 024 Al Skewers 08 3 0005 0003 0 008 Void - 20 9 - 023 - Reflector (Graphite) Graphite 1255 1 731 0 906 898 1 542 Al Matrix 128 7 W7 3 093 058 0 158 Nickel 12 1 001 0 028 Void - 35 6 - oflu - APPENDIX B Analysis of Materials in Weight Per Cent Material Assembly Ag Al B Ba Be Ca Cd Co Cr Cu Fe Mg Mn Mo Na N Pb Si 3n Sr Ta Ti v W Zn Zr Al Matrix All wt% <0 04 02 <0 01 <004 0 001 0 08 <0 08 015 015 05 02 03 <008 <10 03 <0 08 25 <0 08T <01 02 <0 08 <03 <015 Beryllium All wt % 0 05 <0 02 <01 <005 <005 <0O05 005 <0O0IT <001 <002 <005 <01 <0 05 <0 02 <0 02 <02 <01 Stainless Steel (Core Shells} 10 11 wt % <0 05 <0 02 <0001 <01 05 03 <0 O1 1 < 02 <01 2 < 02 0 05 <02 <01 Nickel Sheet 14 wt % <0 05 <0 02 <0001 <01 02 03 2 01 02 < 02 <0 1 01 <0 02 <0 02 <02 <01 Nickel Sample 11 wt % <0 05 <0 02 <0001 <01 ] 0 05 03 ] 0 05 04 <0 02 <01 02 0 05 <0 02 <02 <01 Inconel Sample 11 wi % 03 <0 02 <0 001 <01 1 03 0 02 1 <0 02 <0 1 1 03 0 05 <02 <01 Bismuth Sample 10 N wt % <0 05 <0 02 <0001 <01 <005 <005 <005 <0O05 005 <001 <0 01 <0 02 <0 05 005 <005 <002 <0 02 <0 02 V2 <01 Lead Sample (1 ) 10 11 wt% P <0 05 <0 02 <0 001 <01 <005 <005 <005 02 01 <0 0] <0 01 <0 02 <0 05 <0 05 <0 02 <0 02 <0 02 <02 <01 Lead Sample (1/4 ) 11 wt% P <0 05 <002 <0001 <01 <005 <005 <005 03 01 <0 01 <0 01 <0 05 <0 05 <0 02 <0 02 <0 02 02 <01 310 Stainless Steel Sample 11 wt% <004 <004 <0004 <002 100 015 >10 <0 02T <0 643 031 <063 >0 <0 08 063 <004T <006 <1 3 <0 005 008 <13 <031 <015 Sodium 10 PPM 80 <25 <10 75 2T <10