ENTRAL RESEARVE LN | ; DOCUMEN? COLLECTION | ORNL 1634 S0 4 Al Reactors=Research and Power TINMARIETTA ENERAGY S5YSTEMS LIBRARIES AR 3 4456 0349588 5 ot -t R ‘["’\ N we e e 4 o~ Jfi-rvu IR v " PRELIMINARY CRITICAL ASSEMBLY FOR THE AIRCRAFT REACTOR EXPERIMENT Dixon Callihan Dunlap Scott CENTRAL RESEARCH LIBRARY DOCUMENT COLLECTION LIBRARY LOAN COPY DO NOT TRANSFER TO ANOTHER PERSON 1f you wish someone else to see this document, send in name with document and the library will arrange a loan. s - 0 OAK RIDGE NATIONAL LABORATORY 1 OPERATED BY - - CARBIDE AND CARBON CHEMICALS COMPANY A DIVISION OF Qfll'ph'i‘rc_ugmq‘t’fnu CARBON CORPORATION | . © POST OFFICE BOX P . ' .OAK RIDGE. TENNKSSEE vl ORNL 1634 This doéument contains 59 pages B 1s copy 30 of 155, Series A. CLassI™aTiny CHANGED To- b Ject Category: Reactors-Research o AL [-- Vi el T e S “and Power. £ a&aaaaaaezza___;2:5:él¢24£ZJ, PRELIMINARY CRITICAL ASSEMBLY for the ATRCRAFT REACTOR EXFPERIMENT Work by: Dixon Callihan J. F, Ellis (Now in U. S. Army) E. V. Haake (Now at Consolidated Vultee Aircraft Corp.) J. J. Lynn E. R. Rohrer Dunlap Scott D, V. P, Williams Preparation by: Dunlap Scott Dixon Callihan DATE ISSUED 0CT 238 1953 PHYSICS DIVISION A, H, Snell Director Contract No. W-7k05, Eng. 26 OAK RIDGE NATIONAL LABORATORY Operated by CARBIDE AND CARBON CHEMICALS CCMPANY A Division of Union Carbide and Carbon Corporation Post Office Box P Oak Ridge, Tennessee NAARIET T ENERY SYSTEMS LERARES - AT ‘ 3 445k D3uS95488 5 . C. B. MIL1W\ ORNL 1634 Reactors-Research and Power INTERNAL DISTRIBUTION C. E. Center A C. E. Larson bW. B. Humes (K-25) .. B. Emlet (Y-12) k. M. Welnberg B\l Bengston A, 7. M1V Jd. A. D. V. E. L. E. D. Biology Library h Health Physics Libra¥gk Central Research Librigr Reactor Experimental Engineering Library 3 Laboratory Records Depart Leboratory Records, ORNL R.C. oontiiieen 0 o 5k, 55-59s 60. 61. 62-6h, 65. 66-67. 68-73. Th. 5. T6. T - T8. 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EXTERNAL DISTRIBUTION AF Plant Representative, Burbank WA AF Plant Representative, Seattle L B L2 43-53,- F Plant Representative, Wood-Ridge QP Project Office, Fort Worth gonne National Laboratory AYlR3 Forces Special Weapons Project (Sandia) Atd@Rc Energy Commission, Washington BatBgle Memorial Institute BechiNlk Corporation Brookhien National Laboratory Burecau 4 Ships Californ¥gResearch and Development Company Carbide afikCarbon Chemicals Company (Y-12 Plant) Chicago Pat@at Group Chief of Navigh Research Commonwealth ¥Rison Company Department of ° Detroit Edison duPont Company, duPont Company, Wigmington Foster Wheeler CorpRration General Electric Corfiny (ANPP) Idaho Operations Offic Iowa State College Knolls Atomic Power Labor Los Alamos Scientific Labo Massachusetts Institute of Ti Monsanto Chemical Company Mound Laboratory National Advisory Committee for National Advisory Committee for A Naval Research Laboratory New York Operations Office North American Aviation, Inc. jory (1 copy to H. C. Paxton) hnology (Kaufmann) onautics, Cleveland nautics, Washington Nuclear Development Assoclates, Inc. Patent Branch, Washington Pioneer Service & Engineering Company 121. Powerplant Laboratory (WADC) 3. Pratt and Whitney Aircraft Divisfon (Fox Project) (1 copy A to W. G. Kennedy) Rend Corporation i rancisco Operations Office 126, lach River Operations Office, Augusta 127. g 128. Bl Radiological Defense Laboratory 129-130. Universit*m‘q¥0alifornia Radiation Laboratory, Berkeley 131-132. University oM@iglifornia Redlation Laboratory, Livermore 133. Vitro CorporatMgof America 134, Walter Kidde NuC“.y% 135-140. Westinghouse Elect: 141-155. Technical Information grice, Osk Ridge -l ABSTRACT A zero power mock-up of the Aircraft Reactor Experiment or ARE was constructed in the Oak Ridge National Laboratory Critical Facility. The asgembly was BeO moderated and reflected and used a powder mixture of Zr, Ne.,, and enriched uranium, simulating the reactor fuel, packed In stainless steel tubes as fuel elements. The clean critical mass, without the ARE regulating and safety rods, was 5.8 kg of U-235 to be compared to the predicted value of 5.5 kg. The addition of the regulating rod guide assembly at the center of the reactor increased the critical mass to 6.68 kg U-235. The value of the ARE regulating rod was 125¢ or 1-1/h times the effective delayed neutron fraction for this reactor. The calibration of one ARE safety rod gave a value of approximately 550¢. Neutron flux distributions vwere meagsured by comparing bare indium and cadmium covered indium foll activations at various points in the reactor. The power distribution was found by measuring the fisslon fragment activity on aluminum catcher folls placed in contact with a fused uranium bearing salt contained in an Inconel tube of ARE specifications placed at various polnts in the reactor. The spatial distribution of those neutrons capable of pro- ducing fission was measured in the moderator and reflector using the same catcher foll method with a metallic uranium digk. The reactivity contributlions by a number of materials, primarily con- stituents of the ARE structural componentis, were measured. The rasdial im- portances of fuel and fuel containers were determined by the reactivity coefficient method as was the contribution of a partial mock-up of a steel pressure shell. 1I. IIT. IV, V. vI. V11, Abstract . List of Figures List of Tables - . . o & 9 < INTROTUCTION « = o ° & ¢ OF CONTENTS & o DESCRIPTION OF CRITICAL ASSEMBLY A, B. C. D. FIRST LOADING - A, B. c. D, E, F. G. SECOND LOADING A. B. C. D. B, F. G. BE. I. Jd. K. SUMMARY Moderator and Reflector Fuel and Coolant Mechanical Equipment « « « o Ingtruments and Power Interlocks Critical Mass Control Rod Calibration ¢ & e ¢ ° * a @ ® a e ® o ¢ * ° » e o o o o & e o e ¢ o s v o Reactivity Value of Pressure Shell Radial Importance of Fuel and Fuel Elemant Container Reactivity Value of Reflector Coolant and Tubege .« » Attempt to Calibrate ARE Regulating Rod Meagurement of Neutron Flux Critical Mass ? ¢ © o o o o ¢ * ® & o o o b ¢ o o o s a <@ o &0 o 8 . & . « Regulating Rod Calibration « » « o Fuel Tube Reactivity Coefficients Reactivity Value of Reflector Coolant and Tubes Reactivity Coefficlents Evaluation of Fuel Tube Type Safety Rod o a o ¢ & & B0 ¢ o Reactivity Value of End Reflector - Neutron Flux Distribution ‘a o © o Fission Neutron Flux Distribution Power Distribution é¢ o « oa o o Evaluation of ARE Type Safety Rod - o o ° ° o ACEKNOWLEDGEMENTS APPENDIX Analysis of Materials <& - o @ o L o < o o & L ¢ o - < o & oo o o @ 4 o -6- & @ o ° - * ° o * - ? o 2 2 o+ e @ ® ¢ * . e o ® « - o - ° ° - . - - . . o L] - . . - * - - ® a . . * * @& - - * - > - - - - ® * - L J * - - . ® T & e e - «e & e - - e ¢ & =B & O * & 9 o ® o » - [ ] > - [ 3 - - - ® ® ®» & @ o © © & & = » » * - » - > o o @ - e * - - - * o e * © o o e o o o ° | ] * * . . * * . 5 K ‘l..,.) & MmN - i W o o 32 l.._-i o n o a o - o O o O\ = W o 5 e 18. 19, y 20. 21. 22, 23. 2k, 25, 26, LIST OF FIGURES Top Surface of BeQ Columms o « o - Fuel and Coolant Tubes with Be(Q Blocks Critical Agsembly Structure . . . . Details of Critical Assembly . . . Loading Chart, First Loading . . . Control Rod A Calibration » « » « » Control Rod A Sengitivity . » « « o Radial Importance of Fuel and Fuel Tube o * Indiu.m Tr&verse, Radial s 8 + ece @ O & @ Indim TI‘&'VGI‘SG, Axial * o « @ ¢ 8 0 ¢ . * Indium Traverse, Longitudinal at 8.38" Radius e« ¢ « « o o Indium Traverse, Longitudinal at 15.88" Radius Loading Ch&r‘b, Second. ].-Oad.ing ® & o e o e o . ¢ ¢« & & B8 ARE Regulating Rod Calibration e« o o o o « ¢« o o ¢« o o » ARE Regulating Rod Sensitivity . . Radial Importance, Stainless Steel Fuel Tube vs Void Radial Importance, Low Density Fuel and Inconel Fuel Tube vs s & © @ & & s 8 8 & Stainless Fuel Tube o ¢ o ¢ o o o o o ¢ o« . o * * . » - - Indium Traverse, Longitudinal‘at 12.06" Radiug « » o o o » o Indium Traverse, Radial at Midplane (Regulating Rod In) =« « Indium Traverse, Radisl at Midplane (Regulating Rod ‘Out) « . Fission Flux Distributlon, Radial at Midplane « + « « ¢ o o & Fission Flux Distribution, Longitudinal at 10.09" Radius Power Traverse, Longltudinal at Three Radll « o ¢ o o o o o o Power Traverse, Radial at Five Elevations - Indium Traverse, Longitudinal in Fuel « o o ¢ ¢ o ¢ o s o o @ Fuel Self-Shielding e e o % © © o ® & s 6 e ¢ 8 8 o o @ & @ -7- Page 11 15 16 17 19 .21 22 2k 27 28 30 31 33 35 36 37 39 Ly 45 b1 49 50 52 54 25 56 iI. II1. 1V, V. VI. VII. VIIT. I1X. XL, XII. XIII. XIvV. XVII. XVIIT. Composition of Core and Reflector . . « ¢« v o &« &« + & LIST OF TABLES Reactivity Value of Fuel Tubes « + ¢« ¢ ¢ o ¢ o o & o Radial Importance of Fuel and Fuel Container . . . . . Neutron Flux Traverse Radial at Reactor Midplane . . Neutron Flux Traverse, Axial .« + ¢« ¢ o 4 ¢ ¢ o & « & Neutron Flux Traverse, Longitudinal . . » « + « ¢ + & Radial Importance of Stainless Steel Fuel Tube - - - Radial Importance of Fuel and Container Material . . Reactivity Value of Fuel Tubes « ¢ o « ¢ o« ¢« o ¢ « &« ReactiVity CoeffiCientS * ¢ 4 & e & o ® € s & e = e » Summary of Fuel and Tube Reactivity Coefficients.. . . . Neutron Neutron Neutron Neutron Fission Fiesion Flux Traverse, Longitudinal at 12.06" Radius. . Flux Traverse, Redial at Midplane (Regulating Rod In) Flux Traverse, Radial at Midplane (Regulating Rod Out) Flux Traverse, longitudinal in Fuel . . . . .« . « . . Neutron Flux Distribution, Radial at Midplane Neutron Flux Distribution, Longitudinal at 10.08" Radius Power Traverse, Longitudinal at Three Radii . . . . « . « . . . 8- Page 10 18 23 26 29 29 38 38 40 41 k2 143 46 46 48 51 51 53 1. INTRODUCTION The Aircraft Reactor Experiment being planned at Oak Ridge National Laboratory is a high temperature, intermediate power reactor having a beryllium oxide moderator and reflector and a liquid fuel-coolant. This fuel is designed as a mixture of the fluorides of zirconium, sodium and enriched uranium in the proportion required to meet the requirements at the operating temperature and to include sufficient U-235 to insure a nuclear chain reacting systeml. A series of experiments has been performed at room temperature and essentially zero powsr on a mock-up of the reactor constructed in the Oak Ridge Critical Experiments Facility. The main purpose of the ex- periment was to provide a comparison of the experimental value of the critical mass and of the power and neutron flux distributions with those predicted for this mock-up. In addition, an evaluation was medé of the reactivity coeffi- clents of the regulating and safety rods designed for the ARE and of samples of certain ARE structural materials. The reflector and moderator used in this critical assembly were the BeO blocks which had been prepared for the ARE., The fuel was a mixture of powdered fluorides having essentially the nuclear properties of the ARE fuel except those dependent upon density and was packed in stainless steel tubes., A similar powder, but without uranium, was contained 1n tubes located in the BeQ reflector to simulate the fluid reflector coolant of the ARE. The range of reactivity required to first build an essentially "clean" Just critical system and then to evaluate the rather large amount of poison in the centrally located ARE safety rod necessitated two fuel loadings of significantly different uranium concentration. The experimental results, for purposes of reporting, are divided between these two loadings. 1t was necessary to release the BeO for preparation of the ARE reactor before the experiments reported here could be logically concluded. It was not possible, for instance, to measure all of the ARE regulating and safety systems simultaneously. Nor was it possible to evaluate the data before disassembly of the equipment so doubtful results could not be re- investigated. There were, for example, several neutron flux Vvalues which were considered not representative and are not reported. The construction and operating temperatures of the ARE will not permit a further study of the microscopic nuclear characteristics of the reactor for which these experi- ments were degigned. No attempt at theoretical analysis has been made in this report. It ig intended to present a description of the mock-up and the experlimental procedure, where necessary, with a presentation of the data obtained. II. DESCRIPTION OF CRITICAL ASSEMBLY A. Moderator afid.Reflector The core was a right cylinder, with ite axis vertical, 32.8" in diameter and 35.6" in length and consisted of fuel tubes and hexagonal faced BeO blocks. These blocks were approximately 6" long and 3-3/4" across the hexagonal flats with a 1-1/4" diameter hole, parallel to the long axis, through the center for the fuel tube. The average density of the BeO in the 1 "Reactor Program of the Aircraft Nuclear Propulsion Project”, ORNL 123k June 2, 1952. 9- blocks was 2.76 gm/cc. The BeO side reflector was 47.5" in diameter outside and also 35.6" in length. Theae blocks had the same dimensions as the core blocks except that the central hole was 1/2" in diamster. Some of the periph- eral reflector blocks were cut to approximate a cylindrical outer surface. There was no Be0O end reflector. The BeO blocks had been hot pressed to shape with a rather close limit on the hexagonal dimensions but a large permissible variation (X 0.1") in the length. Although it was possible to make the top and bottom surfaces of the asgsembly plane by selective stacking, the ends of the blocks, in succeggive layers, were not, in general, coplanar. The uniform top sur- face of the BeO and the ends of the fuel tubes are shown in Fig. 1. The large hole, 2" in diameter, in the BeO block in the left foreground is for one of the ARE fission chambers. Kerosene was used in the machining of the blocks and all of it was not removed by a final wash with trichlorethylene ag evidenced by the strong odor remaining. A quantitative measure of the residue was not made. A chemical analysis of the BeO powder (taken prior to pressing) is given in the appendix as sample #19. A summary of the core constituents 1g given in Table I. The two welghts of fuel mixture given are the contents of 70 fuel tubes in the first (#I) and second (#II) loadings, respectively, with the second loading being used to caelculate the weight percent compositiion. TABLE I COMPOSITION OF CORE AND REFLECTOR Core Volume Volums Weight Welght ££3 % 1b Be0 15.50 89.05 2660 89.26 Stainless Steel 0.32 1.84 153.45 5.15 Fuel #I 165.79 (70 tubes) 1.L42 8.16 #IT 166.68 5.59 Voild 0.17 0.95 - - Total 17.41 100,00 2980.1 lOO.bO Reflector Volume Volume Weight Weight £43 % 1b % BeO 18.09 Ok, 70 3104, 98.36 Inconel 0.0k43 0.23 22.86 0.72 Coolant 0.24 1.26 29.17 0.92 Void 0,60 3.13 - - Void 2" Instrument Hole 0.13 0.68 -- — Total 19.10 100,00 3156.0° 100..00 -10- oL W ! S i £ iz B Fig. 1. Top Surface of BeO Columns L1 The BeO blocks were stacked on a 1" thick Inconel plate which, in turn, was supported by elghteen Inconel "legs", 1" in diameter. The support plate was machined to match the fusl and reflector coolant tube matrix. The blocks were contained by a 47-1/2" diameter, 1/16" thick, open end Inconel can. Since the reactor design prescribes a small clearance between adjacent BeO block colums the can did not fit snugly and consequently the blocks were rather loose. Subsequent tightening by shimming with aluminum strips between the can and blocks gave only a small change in reactivity. (The dimensional change was not measurable since the block shift was not uniform but the effect on the re- activity was only about 9¢.) This tightening was done to reduce day to day shifts in reactivity. : B. PFuel and Coolant The fuel mixture used in the critical assembly wag patterned after the ARE fuel mix #27 which prescribed 50 mol % ZrF), L6 mol % NaF ~4 mol % UF), Since insufficient ZrF) was available at the time of the experiments, ZrO, wag substituted and sufficient carbon, as powdered graphite, was added to compensate for the difference between the thermal neutron scattering by the mixtures. The mixture used in the critical experimente consisted of 65 .6 wt % ZI‘OE 23.7 wt 9 NaF 10.5 wt % C 0.3 wt % Hpp with UF) added to give the specified uranium density, For the initial loading this density was prescribed by Mills® to be 0.1632 gm U-235/cc or a fuel composition of: 12.4 wt % UF), (U enriched to 93,2% in U-235) 57.4 wt % Zroo 20.8 wt % HNaF 9.2 wt % C 0.26 wt % HpO The average packed density of the fuel mixture was 1.87 gm/cc. Since no theoretical prediction was made of the amount of uranium required to provide sufficient reactivity to evaluate the ARE regulating and safety rods, the fuel for the second loading was empirically selected to have a U-235 density about 25% greater than the filrst one. The compositlon of the mixture prepared was 2 Mills, C. B. and D. Scott, "The ARE Critical Experiment", Y-F10-108, August 8, 1952, -12. 16.2 wt % UF), 55.0 wt % ZrOp 19.9 wt % NaF 8.8 wt % C 0.25 wt % HyO The uranium density was 0.2138 gm'U-235/cc and the average packed density of the mixture was 1.88 gm/cc. The ARE reflector coolant, at the time of the experiment, was defined asg the same base mixture as the fuel but without the UFy,. The blend actually used was 69 2 Wh % ZrOE 22,7 wt % NaF 9.1 wt % C which differs from that above because of batching irregularities. A specto- graphlc analysis of the materials used for these blends is given in the Appendix, samples 10, 20, and 21, A small quantity of material having properties more like those of the liquid fuel was prepared from a ligquid mixture of ZrFu, NaF, and UF) by casting it into cylindrical slugs. Its composition was Th.0 wt % 2rF) 17.2 wt % NaF 8.8 wt % UFA The bulk density of the solid was 3.75 gm/cc and the uranium density was 0.235 gm U-235/cc. These slugs were 1.06" in diameter, varied in length between 1" and 5" and were of sufficient quantity to fill one fuel tube when packed end to end. There were small volds between adjacent slugs which reduced the loaded density to 3.54 gm/cc or 0.224 gm U-235/cc. Loaded in this manner there was 114.2 gm U-235 in the 35.6" core length to be compared with the average 123.6 gm U-235 in the packed fuel powder of the second loading. The ARE fuel tubes are to be of Inconel 1.235" OD and 0.060" wall and are connected above and below the Be0O to provide flow channels. Inconel was not available at the time of this experiment, so type 302 stainless ateel was substituted. Spectrographic analyeis of this stainless steel is given in the appendix, sample 17. The tubes were 41-1/4" long overall with a fuel filling length of 39.86". This provided 2.1/8" of fuel both above and below the core to simulate the tube bends of the ARE. The tubes were sealed with threaded stainless gteel caps 1-3/16" long, the threads being coated with Glyptal paint to prevent powder leakage and water absorption. The tubes were packed by vibrat- ing longitudinally as the specified amount of fuel mixture was added. X-ray photographs taken of the filled tubes and examined on a densitometer indicated a variation in the linear uranium density near the tube ends which was not measured quantitatively. The method of loading gave less than 1% variation in the overall density from tube to tubs. The reflector coolant tubes were of Inconel, 1/2" outside diameter, 0.020" wall thickness and 36" long, and were packed to an average density of 1.95 an/cc by the same method as the fuel. They were sealed by welding small Inconel plugs -13- into the ends. These tubes occupied 66 holes in the Be0 reflector and were centered vertically. The neutron source was located in the 67th hole in the reflector. The two types of BeO blocks with the fuel and reflector tubes are shown In Fig. 2 C. Mechanical Equipment Figure 3 shows the reactor table, safety rod, and source drive support structure. The control rod drives are shown mounted below the reactor table. The location of the safety and control elements are indicated in Fig. 4. The control elemenis were fuel tubes mounted on a lead screw driven by a three-phase motor. The rod position indication was transmitted to the panel in the control room by selsyn motors for the intermediate positions and llmit switches for the extremes of travel. Each safety rod consisted of seven fuel tubes arranged in a hexagonal pattern mounted on a common base plate and supported by a cadmium lined rod connected above the center tube. This array was drawn into the reactor by a magnet mounted on a drive similar to that of the control rod, with selsyns and limit switches for remote indication, the latter also beilng integral with the Interlock system. Panel lights indicated the "in" position &f the safety rods, the "out" position of the magnets, and contact of the magnets and rods; selasyn-driven Veeder Root counters indicated intermediate positions of the magnets. The two safety rods are shown in the "out" or least reactive position in Fig. 4. In the normal or "in" position during operation the fuel tubes of the safety rods were centered vertically in the reactor. The Po-Be neutron -source was located in the reflector and was moved by a remotely controlled motor between the midplane of the reactor and & paraffin- boron carbide shileld mounted above the Be0. The source strength was about b x 10 neutrons per second. D. Instruments and Power Interlocks The reactivity of the agsembly was regulated by positioning the control and safety rods from the control panel and certaln interlocks were provided In the system to avoid lmproper sequence of operationsg during start-up. It was necegsary that the source be all the way in the reactor and the control rods In their lesast reactive position before the safety rods could be raisged. The control rods could be moved only after both safety rods were in contact with their magnets, and one was completely raised and the other partially raised. In the event that the BeO blocks of the array were dieplaced vertically the up- ward motion of rods was stopped by interlocks. Neutron and gamma sensitive detecting devices were placed about the agsembly for measurement of the reactor power level with assoclated amplifiers, scalers and recorders located in the control room. Two BF; proportional counters supplied information to scalers through linear ambPlifiers for the determination of neutron multiplication during the initial approach to critical. A record of the relative power and period of the assembly was provided by a log N amplifier. One D-C amplifier and two vibrating reed electrometers fed by BFy ionization chambers and a ¥=sensitive scintillation detector provided additionel relative power level Information. ZXach of these last four instru- ment channels contained relays which were used to actuate the safety system. The purpose of the safety devices was to shut down the reactor quickly in the event of unusual rises in reactivity. Actuation of the safety system. ~1h- - RESTRICTED Y=12 PHOTO 11700 Fig. 2. Fuel and Coolant Tubes with BeO Blocks Gl e —— - — Fig. 3. Critical Assembly Structure 16 TABLE TOP 1" IRON o - - x o o ..l_ 5 ~ 0 o { w z3 -y [ — ST T T RIS NN W T SSSSSSSSIS TR UTS SSUITUNSESSSS g B8 WP PPT 7777 R 7PN 7T N 777 z S CONTROL ROD DRIVE o d N/ 2 N I w A-"—'v A—"-v AA“m—v@ A—"-v T A7 LTI L AL L L L ALLLL 3R ORO2OROS OO SRS -5 z0 co o [ o A O RS T R S SN @ PN ALY S LA L 290000092020 b S e 5 %<0 »(0 y~0 »<® )0 )~ > o e e 1 w g 2 K SOENISSUNSISISSSSOSNANSN . M a W_M_ Sl XL L LY LA gy (0 y<0)y~§y—<0)~0)<=) 2255 ¢ e ———— T / &, R O S SR RSN SRS S P IINIIIIIN I A T . - 47.5 T L7 Ll XL L L L LY L LLL AL SO R R RSO RN S SR N ERNENAN 1 S S e N e N S S NN NI N RN N a = xS b N\ ——— « BARE INDIUM (BRI ACTIVATION MINUS >\\ Jo(BR U Cd COVERED INDIUM BR 10|~ ACTIVATION N 2 ruet o ONOE [ TUBESSf | ; | ! ' ~ _ i | | ~R | ¢ 6 12 18 24 DISTANCE FROM AXIS OF REACTOR (INCHES) % FIGURE 9 50 ACTIVITY (ARBITRARY UNITS) - DWG. 21521 INDIUM TRAVERSE, AXIAL - <« BARE INDIUM ACTIVATION B / ] Cd COVERED INDIUM ACTIVATION \ \ INGONEL SUPPORT PLATE / N Cd FRACTION Cd FRACTION / ACTIVATION BARE INDIUM ACTIVATION MINUS Cd COVERED INDIUM”/,\\ / 6 12 18 24 30 DISTANCE FROM BOTTOM OF REACTOR (INCHES) FIGURE i0 36 8¢ TABILE V Neutron Flux Traverse Axial T Distance from Indium Activity in A: rbitrary Units ‘Bottom of cd Bare-Cd Cadmium Reactor Bare Covered Covered Fraction o" 4,870 3,670 1,200 0.2L5 6 29,400 18,930 11,470 0.356 12 4l ,580 28,600 15,980 0358 18 48,500 30,950 17,550 0.362 2k 42,100 30,070 12,030 0.286 30 26,420 15,970 10,450 0.396 36 2,440 1,480 960 0.396 . The low cadmium fraction and the tendency for a non-symmetrical flux at the bottom of the reactor are to be noted, a result probably due to the fast neutron reflection by the Inconel support plate. Similar longitudinal traverges were made at radial distances of 8,38" and 15.88" (Cells 10-8 and 12-8) but in the upper half of the reactor only. Table VI and Figs. 11 and 12 show these data. TABLE VI Neutron Flux Traverse Longitudinal Distance from Indium Activity In Arbitrary Units Bottom of Cd Bare-Cd Cadmium Reactor Bare Covered Covered Fraction 8.30" Radiuse 18" 39,200 24,800 14,400 0.367 2k 34,100 20,890 13,210 0.387 30 20,540 13,710 6,&30 0.332 36 1,920 1,280 640 0.330 15.88" Radius 18 20, 600 11,300 9,300 0.450 24 18,290 10,250 8,040 0.439 30 11,210 6,040 5,170 0.462 36 1,030 540 490 0.473 -29. Ty DWG., 21522 DISTANCE FROM BOTTOM OF REACTOR (INCHES) FIGURE 11 50 i l INDIUM TRAVERSE LONGITUDINAL AT 81.38 INCHES RADIUS —40 BARE INDIUM ACTIVATION -8 = > > x __le S0 | 2 o Cd COVERED INDIUM |ACTIVATION: = x ' : < O : - l E BARE INDIUM ACTIVATION Q Cd. COVERED INDIUM AGTIVATION~| N / ' : 1O ) 'BARE INDIUM ACTIVATION \ MINUS Cd. COVERED INDIUM | ACTIVATKTN l l %_\ 0 6 12 18 24 30 36 DISTANCE FROM BOTTOM OF REACTOR (INCHES 12 - FIGURE T lE IV, SECGNQ“LQADING A. Critical Mass Provision was made to increase the reactivity in the available tube positions by increasing the uranium concentration in the powder mixture, thereby allowing further investigation of the poison regulating and safety rods of the ARE. The U-235 content of the fuel was increased from 0,1632 gm/cc to 0.2138 gm/cc by adding UF,, the relative proportions of the other salts remaining the same. The details of the composition are given in Section II-B. The "clean" assembly was again made critical requiring 5.19 kg U-235 in the 35.6" length of the core and contained in 42 fuel tubes with Rod A removed an equivalent of 15¢.%¥ The values are to be compared with 5.85 kg in 62 tubes which were initially critical with the fuel of lower uranium density. The lower critical mass is a consequence of geveral factors other than the change in uraniwi density. The stainless steel loading was significantly less since the twenty positions formerly filled with fuel tubes were left empty in thie experiment. The reflector wag also effectively thicker since the overall dimensions of the BeO were the same in both cases. The critical mass of th%s loading had not been pre- dicted theoretically but subsequent calculations™ give a multiplication constant of 1.009. No additional experiments were done with this clean agsembly. Immediately after determining the critical mass, the regulating rod aggembly with its assoclated components were mounted at the center of the reactor requiring the removal of the center fuel element, the center 36" column of BeO, and the insertion of three Inconel tubes. The three tubes extended completely through the reactor and were 3.,75" OD by 0.052" wall thickness, 2-31/32" OD by 0.042" wall thickness, and 2-15/32" OD by 0,039" wall thickness. The tubes were approximately concentric. This unit arrange- ment was exactly the same as that in the ARE except the insulation in the annulus between the inner two tubes was omitted*¥, A step-wise evaluation of this change was not done but the mass required to keep the system critical was determined. The required loading was 55 fuel tubes or 6.80 kg U-235 with the regulating rod inserted a distance corres- ponding to a reactivity value of 54.5¢ which was approximately equal to that of one fuel tube. The critical mass therefore, with the regulating rod assembly mounted but with the rod removed was 6.68 kg U-235 in 5k fuel tubes. Figure 13 glves the arrangement of the 55 fuel tubes. This repre- sents an increase of 1.5 kg U-235, or 12 fuel tubes over the unpoisoned gystem, The large change in critical mass shows the Inconel at the center of the core is a strong poison which greatly reduced the total rod effectiveness. B. Regulating Rod Calibration The weak ARE regulating rod was evaluated in the central position, 8.7, ¥ The loading pattern consisted of a hexagon of 4 columns on the slde, plus the following tubes: 10-10, 10-L, 6-k, 6-10, 8-3. 6 Mills, C. B. ANP Quarterly Progress Report June 10, 1953, ORNL 1556, p.28 *% A subsequent evaluation of 5505, a rough equivalent to thisg insulation, mede independently of the Inconel tubes, showed it to effect only a small change in reactivity. The result is given in Table X, and an analysis in the Appendix, semple 8. -32. and subsequently used for operational control. The method of calibration was the same as that described in the Section III-B for the fuel tube type rod and ylelded a total reactivity value of about 125¢. At the "zero" rod position all of the B)C of the rod was above the reactor and the bottom of the stainless steel section of the rod was flush with the top of the BeO. Figure 1bh gives the calibration curve of the regulating rod. It is of partic- ular interest that the minimum and maximum values of the rod as a reactor poison occurred when it was somewhat displaced from the limits of its travel, being a minimum when inserted approximately 3" and a maximum when inserted 32", The reactivity Increase in the interval from zero to 3" can probably be attributed to a contribution of the rod to the neutron reflection or to the reduction in neutron channeling in this bare ended assembly. When the rod was at the 32" position, the top of the 30" section containing boron was at the level of the top of the BeO blocks. Further downward displacement, therefore, did not introduce addltional poison and redistributed that already present only slightly. It d4id, however, bring the top 2-3/16" long steel section into the core, providing some neutron reflection and reducing the channeling. The reactivity would have increased upon further lowering of the rod as boron was removed at the bottom of the core. In a second installation of the rod, to be described later, at a radial distance of T.5" from the center, the above be- havoir at the 3" displacement was verified. It is to be noted in Fig. 15 that the net result of these end effects is to give the rod a negative sensitivity in these ranges. The irregularities in the data near the peak of the sensitivity curve has been attributed to the interaction with an adjacent fuel type control rod which was being Ingerted with the poison rod. At this position their ends were at the same level. These effects were not investigated further. Since the maximum reactivity of this asgembly did not occur when the rod wag at the upper limit of its travel it is important that some consideration be given to the operating position of the ARE gafety rods. The data of Figs. 1lb and 15, or similar results from the ARE, should be used to ascertain the appro- priate ralsed position of the ARE rods to insure that the reactivity is always decreased as they are insgerted, without significant loss in total rod value. A pogitive mechanical stop should be provided to limit the downward travel of the rods at the points where the reactivity is a minimum. It is believed, however, that the neutron reflection provided is the ARE by the pressure shell and the moderator coolant will reduce these effects. C. Fuel Tube Reactivity Coefficients In an effort to provide additional data for comparison with the clean reactor of the first loading, the experiment in which the importance of the stainless steel fuel tube was measured as a function of radius was repeated and the data are shown in Table VII and Fig. 16. Since the holes in the BeO were 3-3/4" apart, no intermediate points have & real meaning and therefore no curve has been drawn through the measured values. 3k ROD VALUE ¢ DWG. !l!! 120 v - / CONTROL ROD B CALIBRATION 100 / ARE REGULATING ROD (WEAK) 90 80_ / 70 / 60 40 / / —t | A ] 1 ] l ] i i 1 ] | | 1 0 2 4 6 € o 2 4 16 8 20 22 24 2 28 30 32 4 ROD POSITION (INCHES) FIGURE 14 2/INCH ROD SENSITIVITY DWG., 21526 I ! ! 0 ARE REGULATING ROD SENSITIVITY | A N\ N \l | ] 18 20 22 24 ROD POSITION ( INCHES) FIGURE 15 26 28 1~ 30 100 GAIN IN REACTIVITY ¢ 20 @® O N O H O 20 e DWG. 21527 RADIAL POSITION (INCHES) FIGURE 16 T T [ ] | ! | RADIAL |IMPORTANCE o STAINLESS STEEL FUEL TUBE VS. VOID 0 o o REGULATING ROD GUIDE ‘/— TUBES (INCONEL) | 1 : | : 1 [ 2 4 6 8 0 12 4 6 L€ TABLE VII Radial Importance of Stainless Steel Fuel Tube , Reactivity Posltion Value in ¢ 9-8 131.0 10-8 9k, 1 11-8 69.9 12-8 5743 A comparison of the reactivity coefficients of an Inconel fuel tube with the stalnless steel fuel tube was made. An Inconel tube of ARE dimensiong and specifications was filled with powdered fuel having the same density and U-235 concentration as that contained in the stainless steel tube. This tube then replaced the stainless steel one in successive positions along a radius. In another experiment some evaluation of the radial effect of a variation in U-235 concentration was indicated by substituting a stainless steel fuel tube with the first loading (0.163 gm of U-235/cc) for one with the higher concentration (0.21% gm of U-235/cc). A summary of these data is given in Table VIII and Fig. 17. TABLE VIII Radial Importance of Fuel and Container Material Compared to Normal Fuel Tube Reactivity Value in ¢ Low U-235 Density! Inconel Pogition Fuel Tube Fuel Tube 5.8 -29.0 -27.2 10-8 -21.1 -24.9 11-8 -1k ,5 -17.2 12.8 - 9.0 ~-10.1 The reduction of the relative value of the Inconel fuel tube at the 3.75" radius (position on 9-8) might be explained by its proximity to the large quantity of Inconel in the regulating rod guide tubes. In general, the stainless steel fuel tube values and the low density fuel tube values are consistent but do not compare very well with the flux and power distri- butions described later in this report. D. Reactivity Value of Reflector Coolant and Tubes Since the center of the reactor was strongly poisoned by the Inconel tubes of the regulating rod assembly which might cause a change in the total reactivity value of the filled reflector coolant tubes, it was re- measured for this asgembly. The removal of 56 of 66 tubes gave an increase -38- LOSS IN REACTIVITY ¢ 30 N o o DWG, 2]525 T ! i ! 1 l ! RADIAL IMPORTANCE | l FIGURE 17 e “OW DENSITY FUEL TUBE — ® VS. NORMAL FUEL TUBE o o NCONEL FUEL TUBE vs. O STAINLESS STEEL FUEL TUBE ® o) o 0O o REGULATING ROD GUIDE /TUBES (INCONEL) | 1 | | [ [ 1 L 1 2 4 6 8 10 12 14 16 RADIAL POSITION (INCHES) in reactivity of 78.2¢. When this is extrapolated linearly, the value of 66 tubes becomes 92¢, an increase from the 82¢ extrapolated in the first loading. There was not enough additional Inconel tubing available to evaluate the containers alone but, considering the small value of the coolant and contalners combined, this did not appear necessary. It should be noted that the ARE has a similar coolant in the interstices of the BeQO blocks. Because of the difficulties 1n handling the hygroscopic material, a total evaluation of this poison was impractical. As reported below, however, constituents of this coolant were evaluated at one point in the reactor. E. Reactivity Coefficlents The effects on the reactor of a number of different materials were evaluated at a single point as follows: The fuel tube of position 10-8 (7%" from the center) was removed and, with the reactor critical, the control rod poesition was noted as the base point. Thls base point was used ag reference for the subsequent runs in which the material to be evaluated was placed in this same position. In the case of a large change in reactivity, additional fuel tubes were added at selected points on the core periphery. The tubes were evaluated as TABIE TX added and the data are given Reactivity Value of Fuel Tubes in Table IX. Posltion Valvue in ¢ ~ The sample materials were 11-3 46,4 used in the form of rods, tubes, L5 60.5 chips, or powder the latter two 5-10 39.3 being packed into aluminum tubes 5-G 51.8 and sealed to prevent leakage and 5-h.} 99 water absorption. The reactivity .6 773 coefficlients of the aluminum tubes I k8.7 (semples 1 and 2 in Table X) were 7-4 56.5 measured and used to obtain the net 5-10} ’ result for the samples contained. 6-11 83.5 Two pamples (chromium and manganese) 5-3 ' were packed into both an aluminum h-B.} 149,7 tube and into an annulus formed by haob two concentric aluminum tubes while others were run twice in a single tube using different quantities of materials, Comparisons gave some measures of the self-.shielding. For cobalt in particalar there is & marked change in the loss in reactivity per mole be- tween the two runs. Table X glves a complete summary of these data. There were several reactivity coefficients measured in the same position (10-8) which pertained more directly to the fuel and contalners. Some of these values have been given previously but for each comparison a summary is given in Table XI. The aluminum fuel tube was packed with fuel to the game density (0.2138 gm U-235/cc) as the stainless steel fuel tube. All reactivity values are referred to a vold in the test cell. =40~ L s ———— . — —— W' TABLE X Metl & Dimensions (Inches) Volume Sample Form U. D. “Wall Thick. Length .3 oo 1 Al Tube 1,001 0.034 40.0 4,17 68 .33 2 Al Tubes 0.545 0.092 39,05 152.1 (Annulus )* 3 Iron Rod 1.248 Solid 44.5 54443 891,95 4 Iron Rod 0.50 Solid 44.95 8.83 144.70 5 Iron Tube 0¢543 0.091 40.12 5.18 84.88 6 Nickel Rod 0.999 Solid 42-1/16 32,97 540.3 7 Nickel Tube 1,004 0.036 48.0 5.28 786f52 8 8i02(Sand) - 0.930 Solid Pack 39.0 26,49 434.1 9 KF Powder 0.930 Solid Pack 39.0 26449 434,41 10 NaF Powder 0,930 Solid Pagk 39,0 26?49 434,1 11 Cr Powder 0,930 Solid Pack 3940 26.49 434.1 12 Cr Powder 0.930 0.1925 39,0 17.39 285.0 13 Mn (Chips) 0.930 Solid Pack 39.0 26.49 434.1 14 Mn (Chins) 0,930 0.1925 39,0 17.39 285.0 15 Cobalt (Slugs) 0.930 Random Filled 39-1/8 26.58 435.6 16 Cobalt (Slugs) 0.408 Solid 39.0 5.1 83,57 17 S. Steel Tube 1.240 0.06 40.0 8,92 146.2 1e Inconel Tube 1.235 0,06 | 40,0 3487 145.3 Total Weight Density gms gm/ce 184.5 2.70 411 2.70 6995 7.84 1129 7.80 652 7.68 478932 8.86 7705 8490 1.76 1,43 1.08 1801.6 4,15 1193.1 4,19 1273.0 2,93 769 2470 1966.5 4.51 717.3 8.58 1119.6 7,66 1200.2 B.26 Weight in Core Moles in gm Core 164.3 6,09 370 13,73 5599 100.25 894 16.2 578 10.35 4056 63,1 571 9.73 697 11.6 566 3 975 427. 7 10.18 1645 31,63 1089.8 20095 1162 21.15 702 12.78 1790 30.37 655" 11,11 997.1 18.01 106849 18,72 flyhole g - in Core ~ 2.0 = 0.33 - 4,2 = 0.31 -196.7 = 1.96 - 0.8 = 3.14 - 33.8 = 3.26 ~20643 =~ 2.98 - 83.9 - b5.54 + 0.9 $0.08 = 27.6 = 2.83 - 4.5 - 0c44 =106.6 =~ 3.37 = 76.3 =~ 3.64 -308,8 - 9,87 -153.8 -12.03 =325.3 =10,T71 -19034 ~17.14 - 70,0 - 3.89 = 93,5 = 4.99 # Sample #2 consisted of two tubes assembled to form the annulus used for samples 12 & 13. The outer tube was sample 1 and the inmner was as described under dimensions of sample 2 above, ‘ —l]w TABLE XI Sumary of Fuel and Tube Reactivity Coefficlents Material Reactivity Change Aluminum Fuel Tube + 155.5¢ Low U-235 Density Stain- less Steel Fuel Tube + 73.0 Stainless Steel Fuel Tube + 94.1 Stainless Steel Tube (Empty) - 70.0 Inconel Fuel Tube 4+ 69.2 Inconel Tube (empty) - 93.5 Stainless Steel Tube | ~and Cast Fuel ‘ + 87.b Inconel Tube and Cast . Fuel + 62,2 F. Evaluatlon of Fuel Tube Type Safety Rod A reference check of the value of the fuel tube type safety rod was made by the rod drop method”?., This method, briefly, involves making the reactor critical with the safety rod in the most reactive position and then, as the rod is actuated, using a high speed recorder to observe the fast tran- sient of neutron level decay. The relation which applies in this case i=s No - Nl Ny x 100¢ = the reactivity change in cents where: NO ig the neutron level at critical Ny 1s the extrapolated level at the time the rod 1s dropped. The value of one of the rods, #2, was 480¢ and that of both when removed gimultaneously was 900¢. G. Reactivity Value of End Reflector To obtain some measure of the contribution to the reactivity of the reflector coolant and pressure shell at the ends of the ARE, the following measurements were made. A one-inch thick 6" x 9" layer of stainless steel was placed on the top of the fuel tubes in a locatlon shown by the dotted lines in the vicinity of position 10-8 of Fig. 13. This gave an increase in reactivity of approximately 6¢. A like area and thickness of sodium gave a 2.9¢ increase in reactivity. This sodium was contained in 8 mil wall stainless gteel cans. H. Neutron Flux Distribution The neutron flux distribution was measured with indium foils by the method . previously described except that a remotely placed uranium metal disk and alum. inum catcher foil were used to normalize the power level.from run to run. The 5 E. L. Zimmerman, "A Graphite Moderated Critical Assembly-CA-4", Y-881, December T, 1952, ~hoo urenium disk was approximately 1- 7/16" in dlameter and 0.0l" thick, weighed 4.5 gm and was enriched to 93% in U-235., The Al foil was against the uranium dise during the foll exposure and was counted in an in-chamber type pro- porticnal counter. The activities were empirically corrected for counter dead time, background, and fissgion fragment decay and then used for power normalization Table XII and Fig. 18 give the result of the bare indium and cadmium covered indium longitudinal traverse at a point 12.06" from the center (position 11-8). The zero of the abscissa is the bottom of the BeO column where it rests on the 1" thick Inconel support plate. The scattering by the Inconel possibly accounts for the reduction of the Cd fraction at this point. TABLE XII Neutron Flux Traverse AdJjacent to Tnconel Cast Fuel Tube Longitudinal at 12.06" Radius Distance from Indium Activity in Arbitrary Units Bottom of Cd Bare-Cd Cadmium Reactor Bare Covered Covered Fraction o 8,920 7,025 1,895 0.212 6 37,280 26,450 10,830 0.299 12 55,396 38,630 16,766 0,303 18 59,941 41,019 18,922 0.316 24 51,612 33,65k 17,958 0.348 30 32,202 21,062 11,140 0.346 36 2,858 1,845 1,013 0,35k The radial flux traverse at the midplane of the reactor is given in Table XIII and Fig. 19. The dashed lines on the figure are activities extrapolated from the data obtained in the fine structure measurements taken near the eleven inch position. There were no similar measurements made in the first loading. The wide gap in this traverse was unexplored because of the Iim- portance which was attached to the study of a unit core cell In the time available. This emphasis has been at least partially contradicted by the fission flux traverse described in the following sectlon. The strong effect of the center regulating rod assembly is indicated in these curves. The comparatively small flux and cadmium fractlon depressions at the center of the regulating rod accentuates the relative inefficiency of this type of rod and guide tube arrangament for reactor control -43- (ARBITRARY UNITS) ACTIVITY 4 DWG. 21529 - | | | : l INDIUM TRAVERSE LONGITUDINAL AT [2.06 INCHES RADIUS | I 50 40 30 20 10 A | | | | /BARE INDIUM ACTIVATION /—Q\ )\ Cd COVERED INDIUM ACTIVATION { \ N N\ Cd FRACTION = TOP OF BeO BARE INDIUM ACTIVATION N MINUS Cd COVERED INDIUM ACTIVATION INCONEL SUPPORT PLATE | ] ] i ] ] i i | i ] | 0 6 12 18 24 30 36 DISTANCE FROM BOTTOM OF REACTOR (INCHES) FIGURE I8 Cd FRACTION (ARBITRARY UNITS) ACTIVITY S Dwg. 21530 T T I | | T T T T INDIUM TRAVERSE RADIAL AT MIDPLANE REGULATING ROD IN ] ! , &S BARE INDIUM ACTIVATION 80— INCONEL [ |/ GUIDE Tu7 I I .' / / ol ] Cd FRACTION % Cd FRACTION -~ SR L % I W T R = I \ 7 | l T/ 20 :. "'\\' /M \\ \\ + L M| /] /‘(\ 7! \ 7/ \ 4\ i~ L1 1 I s / /| “ BARE INDIUM ACTIVATION 10 2. c MINUS Cd COVERED y—FUEL—| s 4 ' INDIUM ACTIVATION TUBES - ROD | — LTy ] 0 | | ! I f | l | l [ 2 4 6 8 10 12 14 16 18 20 DISTANCE FROM AXIS OF REACTOR (INCHES) FIGURE 19 Sy e - ' TABIE XIII Neutron Flux Traverse Radial at Midplane (Regulating Rod In) Indium Activity in Arbitrary Units Cd Bare-Cd Cadmium Radius Bare Covered Covered Fraction 0.00" 61,492 46,396 15,096 0.245 1.03 - 61,265 46,508 14,760 0.241 1.28 61,579 46,185 15,329 0.249 1.69 63,181 49,418 13,763 0.218 2.06 68,352 W7,24h 21,108 0.309 2.06 69,858 50,976 18,882 0.270 2.94 68,414 54,196 14,220 0.208 4,56 79,06k 54,666 24,400 0.309 5.4k 81,787 57,679 24,108 0.295 6.69 73,689 54,792 18,898 0.256 9.56 77,086 47,463 29,623 0.384 10.4k4 66,231 43,445 22,786 0.344 10.88 53,510 38,794 14,716 0.275 ©11.25 52,798 39,888 12,910 0.245 11.62 53,367 38,197 15,170 0.284 12.06 57,380 38,441 18,939 0.331 12.94 61,021 37,054 23,967 0.392 22,06 12,818 5,122 7,696 0.600 | 23.56 4,716 - 24 .00 1,998 633 1,335 0.668 Table XIV and Fig 20 give a similar traverse at the same location but with the regulating rod removed. The neutron flux is slightly greater than that observed in the preceding experiment. TABLE XTIV KNeutron Flux Traverse Radial at Midplane (Regulating Rod OQut) Indium Activity in Arbltrary Units Cd Bare-Cd Cadmium Radius Bare Covered Covered Fraction 0.00" 66,596 50,628 15,968 0.240 1.28 68,250 50,233 18,017 0.26k4 1.69 67,604 52,705 14, 899 0.220 2,06 72,671 52,689 19,982 0.275 5 oIl 85,886 59,994 25,892 0.301 6.69 76,695 55,810 23,885 0,300 9,19 81,760 55,818 25,942 0.318 “L6- (ARBITRARY UNITS) ACTIVITY -.Dwo 21531 1 | l n v T A INDIUM TRAVERSE ] 80 | ’/'\ A RADIAL AT MIDPLANE 'gch:IODNEEL A REGULATING ROD oOUuT TUBES \—BARE INDIUM ACTIVATION N 70—-—-~"'_‘7H‘,P 60 — \ - o \.(‘(/—Cd COVERED INDIUM ACTIVATION — 50 A 40 — Cd FRACTION — 30 — 4] ] —fl"‘“‘—- — 2 ///gle INDIUM ACTIVATION > MINUS Cd COVERED — INDIUM ACTIVATION B 10 TUBES | 'j | ’ _ l ’ l [ | | ! | | I 0 2 4 6 8 12 14 16 E 20 22 24 DISTANCE FROM AX1S OF REACTOR (INCHES) FIGURE 20 Cd FRACTION 44 An Inconel tube, loaded with the cast fuel salt described in Section II-B was ingerted in position 11-8 (11.25" from the center) replacing the normal powder packed stainless steel fuel tube., A longitudinal neutron flux traverse wvag made axlally in this tube with the folls placed between the cast slugs. Table XV and Fig. 21 are the result of this traverse. The three points of Fig. 19 at positions 10.88". 11.25", and 11.62" were made in the same Inconel tube -cast fuel arrangement. TABLE XV Reutron Flux Traverse Cast Fuel in Inconel Tube Longitudinal at 11.25 inches Radius - Distance from Indium Activity in Arbitrary Units Bottom of Cd Bare-Cd Cadmium Reactor Bare Covered Covered Fraction 0.18" 9,487 7,317 2,170 0.229 3.7Th . 25,062 19,622 5,440 0.217 6.68 36,026 26,126 9,900 0.275 10.68 17,233 33,638 13,595 0.288 13.7h 50,90k 38,140 12,764 0.251 20.30 52,798 39,868 12,610 0.245 25.80 42,726 30,759 11,967 0,280 29.74 29,709 20,748 8,961 0.302 34,80 7,467 4,978 2,489 0.333 J. Fission Neutron Flux Distribution A measure of the distribution of neutrons causing fisslions wasg obtained by use of the catcher foil method. A 4-1/2 gm enriched U disk 0,01" in thickness end approximately 1-7/16" in diameter, was_exposed, together with en Al catcher foil, at each of the points in the reactor’ indicated on Fig. 13. The exposures were repeated with the disks and catcher foils covered with 0.02" Cd. Table XVI and Fig. 22 give a radlal traverse at the midplane of the reactor. * This experiment is to be distinguished from the usual power digtribution measurements in which the catcher foils are placed adjacent to the tranlum bearing fuel in a critical assembly. This latter type of experiment’is described In the following section. -h8- e DWG. 21532 t T l | | T l T T ! ' INDIUM TRAVERSE LONGITUDINAL AT {1.25 INCHES RADIUS IN INCONEL TUBE WITH CAST FUEL PN ‘ )/ \ BARE INDIUM ACTIVATION / /\ %\ \ 40— |INCONEL RADIUS \ SUPPORT PLATE \ | b / TOP OF BeO 20 /4 \\ \ e0 | ’ & s\\‘l = -4 0 4 8 12 16 20 24 28 32 36 DISTANCE FROM BOTTOM OF REACTOR (INCHES) FIGURE 24 ¥S RELATIVE ACTIVITY - 4 DWG. 21536 POWER TRAVERSE CAST FUEL IN INCONEL TUBE RADIAL AT 5 ELEVATIONS l.o et . \ T—18.24 INCHES FROM BOTTOM . N OF REACTOR — . 23.17 7 \?3286 6 \\ 28.92 .5 ‘ \35 42 .4 3 .2 R 0 8 10 12 14 16 i8 20 DISTANCE FROM AXIS OF REACTOR (INCHES) FIGURE 25 gs The relative fission fragment activities collected on the three folls are noted on the flgure. Since this fission fragment activity ls strongly de- Pendent on the uranium density iIn the slug surface, a measure was made of the homogeniety of the uranium in the second experiment. Three aluminum folls were again located on the split surface of the slug as shown in Fig. 26b. This arrangement, without the second half of the slug, was exposed %o the leakage neutron flux of another critical assembly and the resultant fission fragment activitles measured. Ignoring the differential back scattering across the face of the slug and asguming the flux to which it was exposed to be uniform, these activities, included in Fig. 26b, give a relative distribution of uranium across the slug face used in the self-shielding measurements. Applying this indicated variation as a linear correction to the values in Fig. 26a and normal- l1zing, the values for the relative fission rates across the fuel slug becoms 72.6 and 72.4 at the sides of 65.6 at the center. These measuremsnts and those of the neutron flux shown on Flg. 19 within the fuel tube at an abscissa of 11.5" were made at the same points. — Cast Fuel 76 °7l+ N Aluminum _/ Catcher Folls Fig. 26a Fig. 26b Bd Fuel Self=Shielding K. Evaluation of ARE Type Safety Rod The safety rod of the ARE was mounted in position 10-8 (7%" from the center). The components of this rod assembly are identical with those of the regulating rod except the construction of the poison rod itself. The total length of this rod is 40-3/8" which includes a 2-3/16" stainless steel guide on each end. The polson section consisted of elghteen annular stalnless steel cang, each 2" long and 2" OD, containing a sintered B),C-iron mixture having a density of 2.45 gm/cc and composed of 80% B,C by volume. The dimensions of the sintered slugs were 1.86" 0D, 1.26" ID and 1.88" long. In this case, the step-wise evaluatlon of the reactivity change during the mounting of the equipment for this safety rod was made. The removal of the fuel tube gave a loss of 9h.1l¢. The removal of the 35.6" column of BeO gave a loss of 224¢, the insertion of the two larger Inconel thimble tubes gave a loss of 172¢, the third or central Inconel gulde tube gave a loss of _56-- 31¢ for a total of 521¢. Since there was not sufficient reactivity avail- able to maintain the assembly critical with the safety rod inserted, it was necessary to agaln use the rod drop method of evaluation. It was obgerved that the maximum reactivity of the system with the safety rod agsembly mounted occurred when the bottom of the rod was 3.2" below the top of the BeO, a condition comparable to that found with the regulating rod mounted on the axis. The rod was dropped a distance of 32-1/4" from this position and caused a decrease in reactivity of 550¢. For comparison the two ARE regulating rods were evaluated by the same method in this position, 10-8. Although of somewhat different construction, thelr position of maximum reactivity was also with the lower.end of the rod 3.2" below the top of the BeO., The rods were also "dropped" 32-1/4" from this position. The reactivity value of the "strong" rod wag 160¢ and that of the "weak" rod was 80¢. The "weak" rod referred to here is the one calibrated In the center of the reactor by the period method described in Section IV-B. The value of the rod there was 125¢., V. SUMMARY The preliminary assembly for the ARE contributed information for checking calculational methods and applicable nuclear constants for a BeQ moderated and reflected reactor. The rather close agreement between the pre- dicted and experimentally determined critical mass of the first loading In- creased the confidence in the multigroup calculations. Although the first calculated value of the critical mass was 5% lower than the measured mass, subsequent refinements in the calculations reduced this to about 1%. The step by step transition from the clean, unpoisoned reactor to a low temperature mockup of the ARE was intended to aid in the understanding of the mechanisms Involved for. calculational purposes and did so to the extent that the transition was carried out. The calibration of the "weaker" of the two regulating rods indicated that 1ts value was, for safety reasons, too high for use with an automatic control device of the type designed for the circulating fuel reactor. The total value of the ARE safety rod was found to be sufficient to accomplish 1ts intended purpose. The flux measurements were not as complste as could be desired, although the radial iImportance functions and reactivity coefficients which were measured should aid in the understanding of the flux and spectrum of this reactor. There was essentially no work dcne in determining the neutron leak- age spectrum from the reflector surface. The power distribution measurements acrogs the fuel within a tube suffered from poor resolution but give some indication of the value of the fuel self-shielding. ..5"[ - VI. ACENOWLEDGEMENTS The work of Dr. L. G. Overholser, Messrs. D. R. Cuneo and A, B, Townsend in the preparation of the fuel is gratefully ack- nowledged. Appreciation is expressed to Drs. R. C. Keen, and E, L. Zimmerman for their help in preparation and review of this report. _58- ANALYSIS OF MATERIALS APPENDIX SAMPLE| MATERIAL FORM Ag Al B Ba Be Ca Cd Co Cr Cu Fe Gd K Li Mg Mn Mo Na Ni Pb Si Sn [ Sr| Ta Ti v W | Zn | Zy 1 Al 1in. OD Tube | % wt|< .04 <.004 | <,02|<.001 <.08{<.04 | <.04[<.04]<.08 |[<1 <.04 | <.04 < .04 < .08 <04 <.15]<.04(<.2|<1 <.04 | <.04 |1 <.3<.08 2 Al ]éin. OD Tube | % wt| <.04] <.004 | <.02|<.001 <.08[< .04 | <,04|<.04{<.08 <46 <.04 | <.04 <.04 <.08 < .04 3 <.04)<,2|<1 <.04 | <.04 |<1 <.3|<.08 3 Fe 1]4 in. Rod % wt| <.04| <.04 [<.004 |<.02( .001 <.08(<.04 | <.04|<.04(<.04T <.04 6 <.04 o1 <.04 08| <.04]<.2| <1 <.04 | <.04 |<1 <.3|<.08 4 Fe }é ine Rod % wt| <.04] <.04 |<.004 | <.02( .0O1 <.08(<.04 | <,04|<.04({<.04 < .04 b <.,04 o1 < .04 08| <.04]<.2| <1 <.04 | <04 (<1 <.3|<.08 5 Fe }é ine Tube % wt| < .04| <.,04T|<.004 | <.02| .001 <.08}<.04 | <.04|<.04|<.04 < .04 3 < .04 3 < .04 08| <.,04)<.2| <1 <.04 | <.04 |1 <3< .08 6 Ni 1 in« Rod % wt| < .04 08 | <.004 | <.02[ .001 <.08}<.04 6 | <04 .04 3 < .04 .3 < .04 < .04 .08 <.04|<,2|<1 <.04T| <.04 [ <.3(<.08 7 Ni 1 in. Tube % wt| <.04| <.,04T|[<.004 | <.02] .001 <.08! < .04 15| <.04| .04 .3 < .04 3 < .04 <.04 081 <.04| <.2| <1 <.L.04T| <.04 <1 <.3(<.08 8 $i0, Sand % wit{ <.04| <.04 [<.0004|<.02{<,001 <.,08[<.04 | <.04|<.04|<.04 < .04 <.01|<.01 {<.04 | <.02 <.04| <.01) <.08 < .04 <.02(<.2 <.04 | <.04 <.3]<.08 9 KF Powder % wt| <,04| <.04 [<.03 |<.04{<.001 <.08[<.04 | <.,04|<.,04(<.04T| <.04 04 | <.04T| <.04 <.04| <.1 <.08 <.04] <.08] <.04|<,2| <1 <.04 | <.04 (<1 <.3|<.08 10 NaF Powder % wt|<,04| <.04 |<.004 |<.02|<.001 <.08|<.04 | <.04]|<.,04|<,04 < .04 08[<.01TI< .04 | <.,02 <.04 < .08 < .04 08| <.02|<.1| <1 <.04 | <.04 (<1 <.3{<.08 11&12 Cr Powder % wt| <.04| <.04 | <.,004 | <.04|<.001 <.08(<.04 .04 .08 1