ORNL 1234 e o | |,'|I|JI|"||| “’ ll l’“”"“'l“”' Pl'”l"l' CENTRAL RESEARCH LIBRARY DOCUMENT COLLECTION Eam 3 445k D3LOSAY & -'*‘""w ‘- = “.;"L. - . E €% |y . z... K:-\ MFABGH !fi”} pc”flf’flflfl.fij’ rrmfirfl, BE — REACTOR PROGRAM P of the *Qg’ & CLABH[F[E’_‘,\ TION CRAMGRR T 7 II"'@\‘ iTy o - 3 AIRCRAFT NU v PROPULSION PROJECT . { .~ b e - By A Br:. CENTRAL RESEARCH LIBRARY DOCUMENT COLLECTION LIBRARY LOAN COPY DO NOT TRANSFER TO ANOTHER PERSON omeone else to sea this document, If you wish s the library will name with document and send in arrange a loan. OAK RIDGE NATIONAL LABORATORY OPERATED BY CARBIDE AND CARBON CHEMICALS COMF’ANY y :/ A DIVISION OF UNION CARBIDE AND CARBON CORPORATION L OCAK RIDGE, IENNESSEE N5 :JP# Vi R ] % N e N : ‘ * Index No. ORNL-1234 | This document contains 138 pages. This is copy f,’g of 216, Series A. Subject Category: Reactors-Research and Power REACTOR PROGRAM of the AIRCRAFT NUCLEAR PROPULSION PROJECT by Members of the - Aircraft Nuclear Propulsion Project Wm. B. Cottrell, Editor June 2, 1952 OAK RIDGE NATIONAL LABORATORY Operated by CARBIDE AND CARBON CHEMICALS COMPANY A Division of Union Carbide and Carbon Corporation Oak Ridge, Tennessee Contract No. W-7405-eng-26 o, i 3 4455 03L058Y L LY Internal Distributiol: 1. G. M. Adams§n 2. C. J. Barton 3. E. S. Bettis 4. D. S. Billingto 5. F. F. Blankensl@p 6. E. P. Blizard 7. A. Brasunas 8. R. C. Briant 9. F. R. Bruce 10. J. H. Buck 11. T. H. J. Burnett 12. A. D. Callihan 13. C. E. Center 14, G. H. Clewett 15, C. E. Clifford 16. W. B. Cottrell 17. W. K. Eister 18. L. B. Emlet 19. W. K. Ergen 20. G. T. Felbeck 21. A. P. Fraas 22. W. R. Grimes J 23. L. F. Hemphillf 24. A. Hollaender f 25. W. B. Humegf 26. R. J. Jonegj 27. G. W. Keilibliz 28. C. P. Keiy 29. F. Kertesj 30. J. A. Lghe 31. C. E. Lfirson 32. R. 8. JNivingston Issuing Offfte Technical Date Issued: J{}i 7 - 857 ormation Departmen Index No. ORNL-1234 Reactors-Research and Power 33. R a/N. Lyon 34. " D. Manly 35. F. B. McDonald 36. fil. L. Meem 37. ‘A, J. Miller 38. C. B. Mills 39. K. Z. Morgan 40, E. J. Murphy 4 H. F. Poppendiek 4 H. W. Savage . R. W. Schroeder 44. E. D. Shipley 45. O. Sisman 46. L. P. Smith 47. A. H. Snell 48. F. L. Steahly 49. C. D. Susano 50. J. A. Swartout 51. E. H. Taylor 52. F. C. Uffelman 53. F. C. Vonderlage 54. A. M. Weinberg 55. E. P. Wigner 56. C. E. Winters 57. Biology Library 58. Chemistry Library 593-60. Health Physics and Metallurgy Library 61. Physics Library 62. Training School Library 63-72. ANP Library 73-77. X-10 Central Files 78-83. Reports Office, TID , ¥Y-12 Area -, g * &t i Index No. ORNL-1234 Reactors-Research and Power (Special) External Distributiorng 84-94. 95. 96-~103. 104. 105-109. 110. 111-112. 113, 114, 115-119. 120-139. 140-143. 144. 145-151. 152. 153-156. 157. 158-160. 161. 162. 163-165, 166-168, 169. 170-171. 172-173. 174. 175, 176. 177. 178. 179. 180-181. 182. 183-186, 187-201. 202-216, ArgonneWational Laboglftory Armed FRrces SpecialfNeapons Project {Sandia) Atomic ERergy CommfEsion, Washington Battelle morial Ingitute Brookhavel} National.aboratory Bureau of SRips California Rysearclf and Development Company Chicago Pate{yt Grgup Chief of NavaRRefearch duPont Compalg General Electr §lfCompany (ANPP) General Electrjil Company, Richland Hanford Operafids Office Idaho Operatighs @ifice Towa State Cgilegd Knolls Atomjif Powgr Laboratory Liockland Agea Offie ‘ Los Alamo '. Massachug®tts Instifite of Technology (Kaufmann) Massachuffetts Instit§fe of Technology (Benedict) Mound Lgporatory Nationalfhdvisory Comittee for Aeronautics, Cleveland NationajAdvisory Co ittee for Aeronautics, Washington New Yk Operations (¥fice North @merican Aviatide, Inc. Nucleffr Development Afsociates (NDA) Patej Branch, Washingfn RangfCorporation Savianah River Operation} Office (Augusta) SagBnnah River OperationWOffice (Wilmington) UA F-Headquarters-Office §f Assistant for Atomic Energy iversity of California RadWgtion L.aboratory tro Corporation of Americd estinghouse Electric Corporgion right Air Development Cente Technical Information Service, Oak Ridge TABLE OF C FOREWORD . . . . PART I. CIRCULATING- L AIRCRAFT REACTORS INTRODUCTION CHARACTERISTICS OF CIRCULANING-FUEL REACTORS uel Reactor Systems eactors Heat Transfer in Circulating Physics of Circulating-Fuel Other Considerations . DESIGN STUDIES OF AIRCRAFF REACTOR SYSTEM Turbojet Radiators Circulation of Fuel Throu . rated Circulating-Fuel Reactor Beryllium Oxide-Mo Engines and Radiato Shielding . Accessory Systems ' Circulation of Fuel Through Intermedlate Heat Exchangers Tandem Heat Exchanger and Shield Design Annular Heat Exchanger and Shield Design Spherical Heat Exchanger and Shield Design . PART II. Page No. 10 13 15 16 17 19 21 23 25 27 29 31 31 33 36 39 THE CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT INTRODUCTION ENGINEERING DESIGN . General Description Core Design . . Primary Coolmg Clrcult Secondary Cooling Circuit Structural Assembly : Core Temperature Distribution | 43 47 47 49 52, 54 54 54 = < € 40 4-ft DIA. CORE | yRANIUM o~ 1 o el ke 20 |+~ ~7 3-ft DIA CORE —— 0 - 0 0.40 0.20 0.30 0.40 0.50 0.60 0.70 - VOLUME FRACTION FUEL-COOLANT IN CORE FIGURE I. URANIUM INVESTMENT AS A FUNCTION OF VARIOUS REACTOR PARAMETERS 8l 19 Figure 1 were prepared for a beryllium oxide-moderated reactor, other calculations indicate that if water were substituted for the beryllium oxide the curves would be shifted less than 10%, and if sodium hydroxide were substituted for the beryllium oxide the curves would be shifted upward a little and considerably to the left. The enormous power density required in an aircraft reactor coupled with the demand for accurate and dependable control essential for aircraft use make it imperative that the reactor possess an inherent stability a- gainst anomalous transients. This requirement is most easily met if the fuel consists of a liquid with a relatively large volume increase with in- creasing temperature. Sincepower is producedin the fuel, this volumetric change should lag the power production very little and the desired self- stabilization should result. The use of circulating liquid fuel (as well as static liquid fuels) results in reactors with self-stabilizing characteristics. On the other hand the loss of delayed neutrons from the active lattice, caused by the circulation of the fuel, had been viewed with concernbecause these neutrons have a damping effect on oscillations in reactor power. It now appears that the loss of these delayed neutrons may be compensated by a damping mechanism associated with the circulation of the fuel. Other Considerations Although the mostvital considerations in the selectionof a nuclear air- craft power plant are weight and frontal area per pound of thrust—whichare both a functionof turbine inlet temperature—other factors are also impor- tant. The amount of fissionable material required per airplane; the time, cost, and inventory of fissionable material involved in reprocessing the fuel; the composition of the fuel; and the ease of maintenance of the com- plete aircraft power plant are also pertinent factors. Fortunately, the proposed circulating fluoride fuel should be relatively easy to reprocess, and the fissionable material installed in each airplane should run less than 100 pounds. Both of these factors combine to give what appear to be the lowest total fissionable material investment per operational airplane {in- cluding material in reprocessing, inventory, etc.) for any aircraft reactor yet proposed. The possibility of draining the liquid fuel from the reactor after a flight can greatly simplify the airplane maintenance problem. The refuel- ing problem is obviously vastly easier than changing solid fuel elements. In addition, design studies have shown that it is possible to design the re- actor so that specially shielded or remotely controlled ground handling equipment need be provided only for the fuel filling and draining operations 20 and yet keep the weight of a 400, 000-kw reactor, intermediate heat ex- changer, and shield combination under 120, 000 pounds. Even if a lighter, divided shield were used, once the fuel was drained radiation should im- pose practicallyno limitations on maintenance work. The combined bene- fits of these factors should decrease ground time, and very possibly make the in-service time of an airplane using a circulating-fuel reactor a high multiple of that of one with a solid fuel. | ko 21 DESIGN STUDIES OF AIRCRAFT REACTOR SYSTEMS The circulation of fissionable material external to the core of the re- actor directs attention to design studies of components of the entire system, ‘such as the reactor, radiator, and engine. Although the circulating-fuel system has not been studied sufficiently to define the performance of the resulting aircraft, enough work has been done to appreciate the more out- standing problems. The preliminary design studies—both with and without an intermediate heat exchanger—of a supersonic airplane application of a circulating-fuel reactor are encouraging. These studies were not intended to establish the finaldesign for such an aircraft but to resolve some gen- eral questions regarding the feasibility of supersonic aircraft with circulat- ing-fuel reactor power plants. The decision to employ a circulating-fuel type of reactor did not nec- essarily resolve the alternative choices that exist in the several other features of the reactor system, namely, the type of moderator, the type of heat exchanger and shield system, or even the nature of the fuel. At this time, however, it.is considered that the fuel will be a mixture of metal fluorides, including uranium fluoride. Certain of these salt mixtures have particularly advantageous combinations of desirable physical properties, including high heat capacity, good thermal conductivity, adequate thermal expansion, satisfactory viscosity, sufficiently low melting point, low nu- clear cross sections, excellent corrosioncharacteristics, and satisfactory radiation stability. It has not been possible, however, to make an irrevo- cable decision at the present time regarding either the type of moderator or the type of heat exchanger and shield system that will ultimately prove most desirable for supersonic aircraft application. Several broad classes of moderator were reviewed for applicability to the circulating-fuel concept, including: 1. low-temperature hydrogenous liquids (such as water), 2. intermediate-temperature hydrogenous liquids (such as hy- droxides), 3. high-temperature solid moderators (such as beryllium oxide). Each of these possible moderator arrangements appears to offer some unique advantages and some disadvantages. Consequently, each of the reactor-shield combinations presented below employs a different type of moderator. No attempt is made in this report to evaluate the relative merits of the various moderator arrangements, nor is it clear that any one is superior to the others. 22 There exists a fundamental choice regarding the heat transfer system in the airplane, i. e., whether to circulate the fuel directlythrough the en- gine radiator or toemploy an intermediate heat exchanger and asecondary heat transfer fluid. The former technique would appear ito require the simplest mechanical system, but it is resiricted to the use of a divided shield and only certain special types of airplane configuration because of the intense source of radiation external to the core; the latter technique would be readily adaptable to a wide range of airplane configurations and would permit the use of near-unit shields, which would greatly facilitate servicing and maintenance of the aircraft. At this stage in the design of a supersonic nuclear aircraft the system that willultimately prove superior is largely a matter of opinion. A fairly complete outline of a design of a supersonic airplane in which the fuel is circulated to the turbojet radiator - is presented together with severalreactor and intermediate heat exchanger arrangements employing various approximations to a unit shield to give some notion of the possibilities of liquid-fuel aircraft reactors. In the design of an aircraft reactor it is most important to know the effects of various parameters on the weight of the aircraft reactor and shield, but it is very difficult to determine these effects quantitatively be- cause a shield having anything like the minimum weight is a very closely integrated, complex unit. The only course seems to be to make quite a number of different detail designs on essentially the same bases and then compare them. Unfortunately, the bases cannot be identical because any engineered shield involves many compromises and, at present, many es- timates and assumptions. As a result, even for comparative purposes weight data on engineered shields probably cannot be estimated closer than + 5% if a wide variety of shield types is considered. This follows from the fact that there is of necessity a considerable variation in conditions, e. g., activation of sodium in a secondary circuitis a muchmore serious problem with a circulating fuelthan with sodium as a primary coolant. These con- ditions should be kept in mind in comparing the shield weights presented in this report. In all cases the shielding was designed to give 1 r/hr in the crew compartment at full power output. ’£hese studies have been re- ported in greater detail in separate reports. 3 One of these reports con- cerns the design in which the fuel is circulated through the engine radiator; the other, the reactor and shield design studies in which the fuel is circu- lated through an intermediate heat exchanger. 5. R. W. Schroeder, gg_eil:,, . "_Preliminary Investigation of a Circulating- Fuel Reactor System for Supersonic Aircraft Propulsion, " ORNL-1287 (to be issued). 6. A. P. Fraas and G, F. Wislicenus, "Several Circulating-Fuel Re- actor —Heat Exchanger—Shield Designs for Supersonic Aircraft," Y- F15-10 (to be issued). Circulation of Fuel Through Turbojet Radiators The circulation of fuel through the turbojet radiator permits the direct transfer of nuclear heat to the air stream. Thus, as in the air-cycle re- actor, there is no temperature loss associated with an intermediate heat exchanger. The circulating-fuel reactor, however, can have a less com- plex design and more rugged structure than an air-cycle reactor because it does not also function as a heat exchanger. The shielding problem changes from one of shielding a large reactor with air ducts to one of shield- ing the crew from the radioactive fuel in the radiator and other plumbing external to the core. This, of course, necessitates the use of a divided shield, with the major portion of the shield around the crew because the size and number of the air radiators makes it quite impractical to shield them. An airplane withadesignpoint of Mach 1. 5 at 45, 000 ft, incorporating a circulating-fuel reactor in which the fuel is circulated through the engine radiators has been outlined. This aircraft, Figure 2, has a gross weight of approximately 350, 000 1b, a L./D ratio of approximately 6. 5, and a wing loading of approximately 70 1b/ft2. A divided shield is employed with hydrogenous shielding about the reactor and with hydrogenous and lead shielding about the crew compartment. By confining the six turbojet en- gines to the space in the fuselage behind the reactor, by using a 120-ft reactor to crew separation distance, and by employing a small crew com- partment (4.5 by 5.5 by 12 ft) it appears possible to keep the weight of the reactor-shield-crew combination down to around 105, 000 pounds. The weights of various components of the aircraft are given in Table 1. TABLE 1 WEIGHT BREAKDOWN OF THE MACH 1.5 AIRPLANE Component Weight (1b) Airframe 105,700 Engines 59,900 Radiators 23,800 Auxiliary equipment (pumps, ducting, etc.) 15,300 Reactor and reactor shield 48,400 Crew shield 56,600 Payload 20,000 Contingencies 20,300 Total 350,000 FIGURE 2. MAGH 1.5 AIRCRAFT WITH A CIRCULATING-FUEL REACTOR pbll=Ld4-A il vt Tz 25 The reactor and fuel system has sufficient overload capacity to pro- vide approximately 155,000 pounds thrust for take-off and landing, with approximately double design point reactor power. The radiation dose re- ceived by the crew could be permitted to double {from 1 r/hr to 2 r/hr) for the short periods involved in these operations as well as during the initial climb. As will all divided shield reactor arrangements, special provision would have to be made for ground personnel. Beryllium Oxide-Moderated Circulating-Fuel Reactor The reactor designed for use withthe airplane outlined above employed beryllium oxide for both moderator and reflector. The core consisis of a number of parallel tubes, arranged in two-series passes, to convey circu- lating fuel through the beryllium oxide lattice. The beryllium oxide reflector is provided adjacent to all core surfaces except the fluid inlet and outletend. Moderator heat is removed by the circulating fuel and is employed usefully in the propulsion cycle. By proper allocation of coolant flows, coolant- tube wall temperatures should run only slightly inexcess of the circulating- fuel maximum temperature. To permit removing reflector heat with a high radiator temperature difference, the design under study postulates cooling the reflector with a non-fuel-bearing fluoride mixture. Thisreactor, Figure 3, conforms to the specifications given in Table 2. TABLE 2 DESIGN SPECIFICATIONS OF THE CIRCULATING-FUEL AIRCRAFT REACTOR Size 3.5-ft sphere Uranium Investment (Total System ) 120 1b Core Constituents Moderator BeO Fuel Fluoride Fuel Mixture Structure Inconel Core Volume Fractions Moderator ~) .65 Fuel ~0.33 Structure ~0.02 Design Point Characteristics Power ~320,000 Btu/sec Fuel inlet temperature ~1000°F Fuel outlet temperature ~1500°F 26 FUEL OQUT ===t FUEL OUT SALT ouUT =~S5ALT. IN -G PLACES * 5 FIGURE 3. Be O-MODERATED CIRCULATING-FUEL AIRCRAFT REACTOR i’ 27 Engines and Radiators A circuMating-fuel type of reactor should be inherently capable of sup- plying propulsive power by means of a turbojet cycle or any of several vapor cycles. The relative performance of these cycles has not yet been clearly established, but, pending comparisons, the use of the turbojet cycle has been presumed. Preliminary optimization studies have indicated that a compressionratio in the regionof 6:1 with aturbine air inlet'temperature of approximately 1250°F is desirable. Further increases inthe compression ratio would diminish the radiator weight at the expense of engine weight, and further increases in turbine inlet temperature (with fixed reactor con- ditions) would favor turbojet weight to the detriment of the radiator log mean temperature difference and radiator weight. Having established the total air flow requirements, the compressor inlet area needed was established on the basis of NACA developmental ex- perience. The number of engines necessaryto accomodate the total air flow (or to provide the total inlet area needed) will depend on the size engine that can be made available when an airplane of the type described will re- quire engines. Consequently the specification of the actual number of en- gines at this time would be very arbritary. The use of any number of en- gines, within reason, would have only secondary effects on the overall airplane weight. However, since the use of six engines appears to be con- venient from an installation standpoint, the use of six engines, Figure 4, conforming to the general specifications listed in Table 3, has been assumed. TABLE 3 ENGINE PERFORMANCE AT MACH 1.5 AND 45,000 FT Thrust per engine 8100 1b Compression ratio 6:1 Turbine inlet temperature 1250°F Equivalent sea-level airflow 665 1b/sec The total thrust provided by the six engines plus that from the auxil- iary systems was found to be approximately 54, 000 lbs. Calculations have indicated a specific impulse of approximately 28 and an over-all thermal efficiency of 29%. This latter is defined as the ratio of net thrust horse- power to reactor thermal horsepower. The radiator design involves a radial grouping of rectangular-shaped banks about the engine centerline between the compressor and turbine. Liquid-conveying tubes are passed through, and normal to, closely spaced §—4 REACTOR SHIELD ASSEMBLY MAIN TURBOJET ENGINE MAIN ENGINE COMPRESSOR FUEL-TO—-AIR RADIATORS MAIN ENGINE TURBINE SHIELD WATER RADIATOR REFLEGCTOR GCOLANT RADIATOR REFLECTOR COOLANT PUMP WATER PUMP AIR TURBINE ELECTRIGC GENERATOR ELECTRIC GENERATOR AIR TURBINE L M Qi A B. G D. E. F. G. H. FUEL PUMP \ll WATER PUMP K. L. :,- REACTOR FIGURE 4 POWER PLANT ARRANGEMENT vgsl-~-24A (ENEUR 8¢ 29 sheet fins. The flow pattern contemplated is countercurrent, as dictated by the temperatures of the two fluids. The arrangement entails the use of dividing baffles between adjacent banks, and the installation of by-pass valves in these baffles will permit controlling turbine inlet temperature while substantially constant liquid mean temperature is maintained. Shielding The circulating fuel introduces new problems in shielding, due to the delayed neutrons and fission product gamma rays which must be shiélded out at the crew position. These radiations specify the crew shield thick- nesses within rather narrow limits, thus removing much of the freedom usually enjoyed in the design of a divided shield. Furthermore, this shield differs from other divided shields in that the reactor shield is of uniform thickness throughout, and all the lead shielding is located at the crew com- partment inside the plastic layer. The method of design is much like that of ANP-53, and is described in another report.” However, considerable uncertainty exists in the strength of the fission product gamma source, which is the dominant component. As a consequence, it was necessary to incorporate considerable conservatism in the design. The total tolerance dose was taken to be 1 r/hr, divided as shown in Table 4. The resulting principle shield dimensions are shown in Table 5. 7.R. W. Schroeder, Preliminary Investigation of A Circulating Fuel Re- actor System for Supersonic Aircraft Propulsion,’ ORNL-1287, to be issued. - 30 TABLE 4 ALLOWED CONTRIBUTIONS TO THE TOTAL RADIA'I"ION, DOSE IN CREW COMPARTMENT Component A. Neutrons O A0 O . Radiators to rear . Radiators to sides . Radiators to front . Reactor to rear . Reactor to sides Reactor to front Total B.. Gamma Rays - m 0 a0 T Radiators to rear Radiators to sides Radiators to front Reactor to rear - Reactor to sides Reactor to front Total Radiation Dose rem/hr rep/hr Flux 0,002 29.4/cm? sec _ 0.002 - 1/4(29.4/cm?2 sec) per side 0.250 5 x 104 hard gamma/cm? sec 0.025 1.38 x 10% mev/cm? sec TABLE 5 0.02 0.02 0.005 0.0005 7.25/cm? sec 0.10 0.010 0.10 0,010 0.005 0.005 0.25 0,025 0.250 0.300 0.300 See ANP-53 0.025 0.100 0,100 0,06 0.05 0.025 0.025 0.75 0.75 THE PRINCIPLE SHIELD DIMENSIONS Shield Reactor (all sides) Crew Shield Rear "~ Sides ‘Vertical Front Slanting Front Material Water Lead and Plastic Lead and Plastic Lead and Plastic Lead and Plastic Thickness (cm) 78 22.4 and 33.4 2.75 and 39.0 0.62 and 37.3 1,50 and 38.0 31 Accessory Systems Accessory circuits would be required to permit cooling the reflector and shield and to provide power for pumps and general aircraft accessory demands. Since it appears that many of the accessories are favored by the use of variable-speed drives, the over-all accessory system comtemplated involves the use of individual pneumatic turbines for power supply. The energy is supplied, and reflector cooling is achieved, by passing com- pressor bleed-off air through the reflector radiators and then to the ac- cessory turbines and to parallel propulsive nozzles. The accessory tur- bines are controlled by means of variable-area discharge nozzles, which provide some net thrust after imparting energy to the accessory turbines. Preliminary analysis of this system indicates that it not only provides re- flector cooling and accessorypower adequately, but that the reflector cool- ing system, when considered as an cpen, Brayton-cycle power plant, has a favorable cycle efficiency and power-to-weight ratio. The use of individual accessory-drive turbines permits supplying the accessories with power and speed control without special mechanical, hydraulic, or electrical trans- mission systems. The radiator for the shield water is cooled by low-pressure compressor bleed-off air, and the air is then discharged through a propulsive nozzle. Circulation of Fuel Through Intermediate Heat Exchangers The limitations imposed by partially shielded nuclear power plants on ground handling, servicing, and maintenance of nuclear aircraft are quite serious. The circulation of the reactor fuel to the turbojet radiators im- poses a further constraint on possible aircraft serviceability, In addition to serviceability, however, significant limitations are also imposed on crew compartment size, airplane configuration, and possibly on aerody- ramic stability and control in nuclear aircraft that employ divided shields. It is therefore apparent that strong incentives exist for developing reactor arrangements which are amenable tc a i iype of shield. - It is notable that for the range of sizes of reactor and crew compart- ments under consideration, and with the separation distances involved, the divided shield results in the lowest shield weight; unit shields are appreci- ably heavier.® Obviously, a type of shield can exist that is intermediate in weight and concept, between these two extremes which would permit a 8 'Report of the Shielding Board for the Aircraft Nuclear Propulsion Program,' ANP-53, October 16, 1950, 32 mechanic without special shielding to service the engine and reactor—al- beit for short periods of time. An upper limit on the radiation does from such an intermediate shield was established as 1 r/hr at 50 ft from the reactor operating at 5% full power. A circulating-fuel reactor system in which the ground handling and servicing are greatly facilitated may be effected by incorporating a com- pact reactor design that permits easy and complete fuel drainage with a quasi-unit shield such as that defined above. The reactor heat would then have to be transferred from the circulating fuel to a secondary fluid in an intermediate heat exchanger located within the reactor shield. Since the use of intermediate heat exchangers enclosed within the re- actor shield substantially increases the size of the shielded volume, con- siderable effort has beendevoted todetermining the optimum configuration and location of the heat exchangers. Three heat exchanger and reactor arrangements have been investigated; each was somewhat better than its predecessor. Thethree arrangements, in chronological order of analysis, are: tandem, annular, and spherical. - Each arrangement is distinguished by the location of the heat exchanger withrespect to the reactor core. Iden- tical assumptions were made in the design of the shield for each arrange- ment. In the arrangements discussed, the use of both low-temperature (water) and high-temperature (sodium hydroxide) moderators inthe reactor core are described. No attempt has been made at this time to select the optimum moderator from among these and the previously discussed beryl- lium oxide moderator. In all cases the heat exchanger was separated from the reactor only by enough material to reducethe reactor neutron-leakage flux entering the heat exchanger to the level of the delayed-neutron flux. This required the equivalent of about 10 in. of water between the active lattice and the heat exchanger matrix. It was found that by lacing the intermediate heat ex- changer with 8.5 vol % boron carbide it should be possible to utilize prac- tically any material in the secondary circuit without encountering an intol- erable amount of activation, because most of the neutrons absorbed in the heat exchanger would be absorbed by the boron. Probably no coolant likely to be used would be inclined to become much more radioactive than NakK, yet evenwith NaK thisheat exchanger arrangement should make it possible to keep the maximum radiation dose down to less than about 20 r/hr at 5 ft from the engine radiators immediately after a long flight. With fresh NaK (or NaK that had been allowed to decay for 5 days) the corresponding dose after 1.5 hr of warm-up at 5% power wouldbe only .1 r/hr. Estimates of the activation of other coolants such as natural lithium, water, a eutectic melt of LiCi-MgCl,-KCl, and a noninflammable sodium-lead alloy were found to give activities from 1 to 50% of that of NaK. 33 It is gratifying to find that it is possible to design something like a unit shield of about 120, 000 1b that makes possible servicing and tune-up work with circulating-fuel reactors even with the reactor running and permits ground personnel to approach a nuclear airplane for short periods without special shielding in the event of a minor landing accident or the like. A summary of the various arrangement and design conditions together with the resulting shield weights are given in Table 6, Tandem Heat Exchanger and Shield Design The first arrangement considered was one in which the heat exchangers were grouped behind the reactor core (Figure 5). This reactor design was ‘based on the use of water as a moderator although it is readily adaptable to sodium hydroxide or some other liquid moderator. The reactor core is composed of tubes through which the liquid fluoride fuel flows at about 10 ft/sec for operation at 400, 000 kw with a 400°F temperature rise through the reactor. The fuel enters at the top aft end of the reactor, makes a complete loop through the fuel tubes in the core, and discharges to the heat exchanger. A divided shield and two quasi-unit shields, as summarized in Table 6, were designed for this heat exchanger arrangement. The shield design features and specifications are discussed for each case. The first shield designed for the tandem heat exchanger arrangement was a divided shield somewhat similar to the NEPA divided shield.} The reactor shield was designedto provide no heavy material except in its structure. For simplicity, the shielding material was considered to be principally borated water containing two atoms of boron for every 100 atoms of hydrogen. The thickness of the reactor shield was taken as being such as to give approximately the same dose at the rear of the crew shield as in the NEPA shield, except that the lead shadow-shield disks at the reactor in the NEPA shield were moved to the rear of the crew compartment. The thick layer of steel (3. 75 in) in the form of the core pressure shell and shield inner shells coupled with a layer of lead shielding around the heat exchanger made it possible to remove the lead from the sides of the NEPA crew compartment. The total weight of reactor, intermediate heat ex- changer, reactor shield, and crew shield was found to be 111, 000 pounds. A second shield for the same reactor was designed to provide some- thing as nearly like a unit shield as possible. In this design enough shield- ing was provided at the reactor to bring the radiation leakage from the reactor shield down to the same level as that from the sodium in the sec- ondary circuit. This was done by introducing a layer of lead, 5.4 in. in thickness, around the reactor and a layer, 10.4 in. in thickness around the heat exchanger. The thicknesses of the water layers around both the TABLE 6 PARTIALLY DIVIDED SHIELDS FOR CIRCULATING-FUEL REACTORS WITH VARIOUS INTERMEDIATE HEAT EXCHANGERS Tandem Heat Exchanger Annular Heat Exchanger Spherical Heat Exchanger 1 2 3 4 1 2 3 1 2 3 Reactor shield diametar {in.} 150 150 121 121 148 118 118 148 118 118 Crew shield weight (1b} 5,000 11,000 36,000 14, 000 5,000 36,000 14,000 5,000 36,000 14,000 Weight of reactor, intermedi- ate heat exchanger, and re- actor shield {lb) 151, 000 130, 000 75,000 75,000 123,000 62,000 62,000 115,000 54,000 54,000 Total weight of reactor, in- termediate heat exchanger, and all shielding (includ- ing crew shield) (1b) 156, 000 141, 000 111,000 89, 00C 128,000 98,000 76,000 120,000 90,000 68,000 Reactor power (kw) 4000, 000 400, 000 400, 000 400, 000 400, 000 400, 000 400,000 400, 000 400, 000 400, 600 Diameter of reactor core (in.) 32 32 32 32 32 32 32 32 32 3z Liquids in primary and sec- ondary circuits Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Fluoride-NaK Temperature loss in inter- mediate heat exchanger(0F) 100 100 100 100 100 100 100 100 100 100 Pressure loss in intermedi- ate heat exchanger (psi) 160 100 100 100 100 100 100 100 100 100 i i 1 1 1 i 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 Crew shield size (ft) 6z x7 xlz2 6-2'):7235122 6 x?lez 5x5x122 62x72x122 6Zx'llez2 5x5x122 62:':7'2);122 6Zx72x122 5x5xl123 Reactor-crew separation distance (ft) 50 50 50 120 50 50 120 50 50 50 Radiation inside crew com- partment (r/hr) 1 1 1 1 1 1 1 1 1 I Radiation 5 ft from center of reactor (r/hr) 300 2,400 380,000 380,000 300 380,000 330,000 300 380,000 380, 000 Radiation 50 ft from center of reactor (r/hr) 7 36 5,600 5,600 7 5,600 5,600 7 5,500 5,600 Radiation 300 ft from center of reactor (r/hr) 1 6 156 156 1 156 156 1 156 156 Ve U DWG. E-Y-F 7-474R{ APPROX. 12 in. H,0 AROUND HEAT EXCHANGER AND PUMP HOUSINGS H,0 MODERATOR INLET PIPE Y, -in. STEEL PLATE FOR REACTOR AND SHIELD SUPPORT, 2 REQUIRED HEAT EXCHANGER H,0 MODERATOR l PIPE OUTLET ACTIVE 1 LATTICE SECONDARY COOLANT LINES FUEL PUMP {-in. THICK SHELL HOUSING CIRCULATING AIR AND THERMAL EXPANSION GAP { . FUEL EXPANSION /4-in. THICK TANK, 4 REQ'D 3g-in. THICK ALUMINUM ¥ -in. MINIMUM H,O LAYER AT 300-350 psi AROUND ENTIRE REACTOR AND HEAT EXCHANGER CASTINGS FIGURE 5. H,0—-MODERATED CIRCULATING —-FUEL REACTOR WITH A TANDEM HEAT EXCHANGER Gt 36 reactor and heat exchanger were also increased. A light shield at the crew was then used o bring the total dose in the crew compartment down to thetolerance of 1 r/hr at full power output. The total weight of reactor, heat exchanger, and shield was found to be 156, 000 pounds. Since the above shield seemed to be excessively heavy, an attempt was made to reduce the thickness of lead in the reactor shield and increase the amount of shielding at the crew. It was felt that even if a mechanic could not remain in the vicinity of the reactor for an extended length of time it would be a tremendous advantage to be able to carry out a limited amount of airplane maintenance work. The bulk of the maintenance work would then be carried out after the fuel and secondary coolant had been drained, and yet there would be an opportunity for checks, adjustments, and other tune-up work as the power plant was being warmed up. Accord- ingly, a shield was designed that permitted a dose of something like 1 r/hr at 50 ft from the center of the reactor at 5% of full power output. In this design some of the lead and water was removed from the reactor and heat exchanger shield evolved for the preceding design and the shield thicknesses at the sides andrear of the crew compartmentwere increased accordingly. The total weight of this reactor, heat exchanger, and shield was found to be 141, 000 pounds. The design study covering an airplane in which the circulating fluoride fuel was carried directly from the reactor to the engine air radiators uti- lized a reactor-to-crew separation distance of 120 ft and a smaller crew compartment than that employed in the shields of this study. To get some idea of these effects a new crew shield was designed to go with the first reactor shield design. A greater reactor-to-crew separationdistance re- duced the dose at the crew by a factor of about 6, and the smaller size of the crew compartment gave another factor of nearly 2 in crew-shield sur- face area. This made possible a saving of 22,000 lb in the weight of the crew shield. Annular Heat Exchanger and Shield Design Since much of the weight in the tandem heat exchanger arrangement seemed to arise from the large surface area of the long pressure shell of the reactor-heat exchanger combination, the more compact arrangement inherent in the annular heat exchanger arrangement seemed promising. 'The layout shown in Figure 6 was prepared for a sodium hydroxide moder- ated and reflected reactor. Critical mass calculations indicated that even with only 25 vol % sodium hydroxide in the core the critical mass of an unpoisoned 25 in. dia sphere would be about 35 pounds. In the design shown, a series of concentric helical coils of tubing were provided to carry the sodium hydroxide through the reactor core with a circulation NaK INLET NaOH -NaK HEAT EXCHANGER Y5 0D TUBES ON % G-¢C SQ PITCH HEAT EXCHANGER INLET MANIFOLD NaK INLET FIGURE 6. NaOH- MODERATED (L L LLLL e e SRR o f Yo NaOH L LS S LN AR, Y» B4C GANNED WITH 0080 STEEL (*/g STEEL s Z /j_ InaRERe DWG. E-Y-F7-213aRl NaK OQUTLET HEAT EXCHANGER QUTLET MANIFOLD HELICAL COIL COOLANT TUBES o 1.250 OD TUBING i ) sreel 27000 —— - B e i TUR / y 0080 STEEL ™ 1'% ¥ %“ J STEEL ) OO0 000 N g\‘ Y» B4C GANNED IN h AN ooeo sTeEL R 5 U\ /] seee OO0 2 a! - boPDPPDPOEPOOO ;\\ 2 PRESSURE SHELL : P‘P?p E‘ = ) | =% i\ NaOH h W —f ™ ‘ g [T g qqaq B N - R loQaaQaddaaaq ‘é ?\ . . ol ’ Q0 UQ 1N 1% B,C CANNED IN QC Q O \ N 0060 STEEL o~ o b b, /2 NaOH ) N Y, STEEL 1N 2 NaOH QUTLET TO fi~ NaOH-NaK HEAT EXCHANGER -/ AN 10Y% CIRCULATING { AN J [ NaOH REFLECTOR % et 6 ANNULAR TYPE HEAT EXCHANGER Y% OD TUBES ON ¥3p €.-¢ SQ PITCH AN I T I I I T 77 ==t ~ NOTE: DIMENSIONS ARE IN INCHES /"’/' 2, HEAT EXCHANGER OUTLET LINE, 6 REQ'D ~/ NaK OUTLET CIRCULATING—FUEL REACTOR WITH AN ANULAR HEAT EXCHANGER g 38 rate high enough through both core and reflector to keep the outlet temper- ature to 12500F with an inlet temperature of 1050°F. A special small heat exchanger was placed in the inlet manifold for the NaK of the sécond- ary circuit to provide cooling of the sodium hydroxide, The reflector was made 11 in thick and was separated from the heat exchanger by a 1 in. thick layer of boron carbide to attenuate the neutron leakage flux from the reactor to the level of the delayed-neutron flux from the fuel in the heat exchanger. As in all these heat exchanger designs, the center tube of every square cluster of nine was a '"dead' tube filled with boron carbide and sealed off at the ends. The shield designs were similar to those prepared for the tandem heat exchanger except for their shape and reduced volume. It was found that both the delayed neutrons and the decay gammas from the fuel in the heat exchanger were attenuated more rapidly in passing through the shield than the prompt neutrons and gammas released in the core. Thus, the bulk of the radiation leakage from the shield surface came from the core instead of from the fuel in the heat exchanger. In fact, the material in the heat ex- changer was moderately effective as shielding material (roughly 60% as effective as water); hence, the heat exchanger actually at least partly ''paid its own way'' instead of being completely parasitic. Two by-products of the design study with the 11 in,thick reflector fol- lowed by 1 in. of boron carbide appear to be so important that they alone would justify the use of a thick reflector. If a considerable amount of boron is provided in the shield water to suppress secondary gammas from neutron captures inhydrogen, the indications are that with a thin reflector the steel pressure shell becomes a more important source of gammas than the re- actor core, so far as leakage from the outer surface of the shield is con- cerned. Closely associated with this is the fact that with a thin reflector the problem of gamma heating of the pressure shell is quite serious, where- as with the arrangement used it was not. Calculations indicate that a re- flector thickness equivalent to 10 in. of water followed by 1 in.of boron car- bide is the minimum that should be provided from these two standpoints. The layer of boron carbide would, of course, require careful and thorough cooling by passing sodium hydroxide or some other coolant through tubes imbedded in the canned boron carbide. The weights of the reactor-annular heat exchanger-shield combinations were foundtobe 98, 000 and 128, 000 1b for the divided and near -unit shields, respectively, a saving of 13, 000 lb over the tandem arrangement for the divided shield and 28, 000 1b for the near-unit shield. Fairly complete specifications are presented in Table 6. 39 v, Spherical Heat Exchanger and Shield Design The large saving in shield weight effected by going from the tandem to the annular heat exchanger arrangement was most encouraging. It im- mediately suggested that a further substantial reduction in weight might be realized by going to a more nearly spherical configuration. A novel heat exchanger in which the tube matrix was laid out in the form of a spher- ical shell was devised toform a basis for further designs. The basic con- figuration envisioned is shown in Figure 7. The heat exchanger tubes were grouped in bundles. Each bundle had a rectangular cross section except where the bundles terminated in circular disk headers. If the two pumps are taken as being at the North and South poles respectively, each tube bundle would lie on a variable pitch helix running from approximately the Arctic to the Antarctic Circle. The angle between tube centerlines and the Equator would be about 40 deg, whereas the corresponding angle at the Arctic circle would be about 90 degrees. This would make it possible to keep the center-to-center spacing of the tubes independent of latitude. Although the tube bundle shape is quite unusual, fabrication of the heat exchanger should not be too difficult. After jigs had been prepared for building up a tube bundle, one tube bundle after another could be assembled, brazed and/or welded into its header disks, pressure tested, inspected, and assembled in the spherical steel shell. A simple fillet weld between the header disk and the pressure shellwould suf- fice to complete the installation. In other respects the heat exchanger and shield design of Figure 7 is essentially similar to the tandem and annular arrangements described pre- viously. The 11 in. thick reflector of the annular heat exchanger design was used again, as was the same type of shield construction employed in both of the other designs. One difference was that boron carbide was placed in the baffles between tube bundles instead of in ''dead' tubes. The reactor design is suggestive of a homogeneous reactor, which it really approaches quite closely. The calculatedcritical mass proved to be reasonably low—only 20 pounds. Credit for this must go to the thick beryllium oxide reflector. The latter would be cooled by a nonuranium bearing fluoride salt. It should be possible through careful design of the moderator cooling passages and the diffusing vanes at the reactor core inlet to keep the metal temperature at all points below the average temperature of the fuel leaving the reactor. Thus, the hottest metal anywhere in the system would be the tube walls at the fuel inlet end of the heat exchanger. As shown by the data in Table 6, the sphericalheat exchanger designre- duced the over-all weight of the reactor-heat exchanger-shield combination for the design conditions to only 90, 000 1b for the combination using a divided shield and 120, 000 1b for the near-unit shield, a savings of roughly 8000 lb in both instances over the annular heat exchanger arrangement. Pt Lo Saarr HEAT ExcpancEr TUBE Bunois NN N N ~ U ™. S ~ Bowsaren AH20 LRESSLRE SHELL SELF- SEALNG SUBBER INNVER FRESSURE SHbke s OF Smrer o FIGURE 7 REACTOR WITH A SPHERICAL HEAT EXCHANGER oy 182~Ld-A AEQUT PART 11 THE CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT 43 INTRODUCTION The ARE is areactor experiment whichhas as its fundamental objective the acquisition of useful and reliable data that cannot be obtained faster and cheaper by any other means. Accordingly, these data must be of such nature that they will expedite the development and fabrication of the aircraft reactor in such a manner as toreduce the aircraft reactor cost, in time and money, below whatthis cost would be without the ARE. It is the opinion of the members of the ANP Project of the ORNL that the aircraft reactor will - most economically be attained through the information to be obtained from this Aircraft Reactor Experiment. Indeed, as the task of flying a super- sonic airplane on nuclear energy is so exceedingly complex and the nec- essary data are so meager, more than one experimental reactor may be required before sufficient information and experience are obtained to per- mit the design and construction of a reactor for a useful airplane. The experimental reactor is, as nearly as is practical, a prototype of 350-megawatt, BeO-moderated, circulating-fuel aircraft reactor..The core is approximately a 3-ft-square beryllium oxide cylinder in an Inconel shell. The fuel is circulated through the central portion of the beryllium oxide moderator and thence to anexternal heat exchanger, pump, and back to the core. While in the core, the fissionprocesses raise the temperature of the fuel by several hundred degrees to 1500°F. The reflector and pres- sure shell are cooled by a secondary system independent of the circulating fuel. This basic nuclear design of the ARE is capable of developing many times the power output at which it will be run in the experiment. As the primarypurpose of the ARE is to obtain experience in the high- temperature range with a nuclear power source, the reactor will be oper- ated at design-point power for as much of its lifetime as operating diffi- culties permit. However, it is planned to limit the operational life of the reactor to about 1000 hours, but with as much time at design power of 3 megawatts as possible., Other critical parameters of the experimental reactor are as nearly in one-to-one correspondence with aircraft reactor specifications, as practical considerations permit. For example, the maximum temperature in each is 1500°F and the temperature drop across the ARE is 350°F, while that in the aircraft reactor would be around 500°0F. No attempt, however, will be made to make ausefuldisposition of the output power which will be dumped directly as heat energy. It is ultimately car- ried away by a waste water system. Although most of the thermodynamics of the fuel can be evaluated with- out the use of a chain-reacting system, there is no way of simulating the heat production of the fission process. Therefore, the power-production 44 and heat-removal functions of the experimental reactor will be as thor- oughly studied as the operation time of the experiment permits. Another most important aspect of the ARE is the information to be obtained on the control and stability of such circulating-fuel reactors. Answers tosuch questions as effecton reactivity of temperature fluctuations in the fuel and effect on stability of the "lost' delayed neutrons may then be ascertained. Obvicusly, to build a circulating-fuel reactor requires the existence ~of a liguid fuel and structural metal, both capable of prolonged operation in a mutual environment at high temperatures and radiation fluxes. In addition, these materials must both possess certain ratherrestrictive physical, chemical, and nuclear characteristics. In particular, the fuel must have a low viscosity and a reasonably high heat capacity to effect the desired heat removal from the cores. A fuel which seems to possess the minimum requirements in these categories is the system of fluorides of sodium, potassium, zirconium, and uranium. This fuel has been studied to a considerable exient, and the data on the fuel characteristics indicate that it can be satisfactorily employed in the ARE. The fuel tubes and structural members of the reactor will be fabri- cated of Inconel, as this metal appears to possess the best combination of the required characteristics. Even Inconel, however, does not possess entirely satisfactory corrosion resistance, although there is every indica- tion that the dynamic corrosion of this metal by the fluoride fuel may be improved. Otherwise, the metal, the fuel, and the beryllium oxide mod- erator possess the desired high~temperature characteristics. Experiments have shown that the proposed materials suffer no radiation instability at the intended operating flux levels of the ARE. However, there is some uncertainty regarding the existence and extent of the radiation- induced corrosion at neutron fluxes approaching that to be expected in the aircraft reactor. | Alrcraft reactors would require minimum structural thicknesses within the reactor, as well as componcnts with the minimum size and weight. At the present time, however, information on corrosion rates and properties ~of structural materials under the proposed operating conditions, is too meager to permit the incorporation of aircraft-type mechanical features into the ARE without jeopardizing the entire experiment. Accordingly, in instances where the relevance of the experiment is riot unduly compromised, the design of ARE components has emphasized durability and reliability. In general, therefore, the design of the ARE is analogous to that of an engine dynamometer test cell, in that in both instances the relevance of the item under test is preserved, whereas the power absorption equipment and other accessories are not intended to simulate those of the ultimate applica- 45 tion. Thus, insofar as possible, the experimental reactor is a low-power 3-megawatt prototype of the 350-megawatt circulating-fuel reactor proposed for supersonic aircraft propulsion. Summarily, the ARE project stands at a crucial point in the ANP pro- gram where some five years of analytical work have failed to establish the impossibility of achieving nuclear-powered flight, but where gaps in avail- able data do not permit the analytical proof that success can be attained, The purpose of the ARE is tofurnish some of the crucial data that are lack- ing so that with these data the solution of the problem can go forward, If efficient use is to be made of the substantial effort being expended in the ANP program, the expeditious completion of the ARE program is impera- tive. It is the purpose of this report to describe the design and operation of the facility and as accurately as possible to establish the time schedule for accomplishing the Aircraft Reactor Experiment. The design data for this reactor are given in Table 7. | TABLE 1 AIRCRAFT REAC TOR EXPERIMENT DESIGN DATA Heat flux (Btu/hr/sq ft) Power (max/avg) Power density, max (kw/liter of core) Specific power (kw/kg of fissionable material) ~ Materials and Amounts - Fuel Uranium enrichment (% U235) Critical mass (kg) Total uranium inventory (kg) Fuel elements Fuel-element jacket Moderator General Location Oak Ridge Operator ORNL Purpose Experimental Neutron Energy Epithermal Status Design Heat, Max (kw) 3000 Heat transported out by circulating fuel ~~2:1 5 < 400 NaF-KF-ZrF4-UF4 93 .4 ~12.5 ~65 66 parallel tubes (each 3 ft long, 1.235 in., OD, 60-mil wall) containing the circulating liquid fuel Inconel BeO 46 (Table 7 contd.) o - Reflector BeO Shield Concrete . Primary coolant The circulating fuel Reflector coolant NaF-KF-ZrFy Circulating Fuel-Coclant Maximum fuel-coolant temperature (OF) 1500 Maximum sheath temperature (CF) 1510 Maximum Inconel temperature (OF) 1510 Maximum moderator temperature (OF) 1675 Consumption at max power (g/day) 3.0 Design lifetime (hr) 1000 Burnup in 1000 hr at max power (%) 0.25 Inlet temperature (°F) 1150 Outlet temperature (°F) ' 1500 - Flow velocity (ft/sec) 4 Flow velocity (gpm) 83 Pumping power (hp) 10 Neutron Flux Density (Avg) Thermal-max (n/cm2/sec) 3 x 1013 . Thermal-av (n/cm2/sec) 1.5 x 1013 " Fast-max (n/cm2/sec) 7T x 1_013 ' Fast-av (n/cm?/sec) 3 x 1013 , ] Intermediate -av (n/cm2/sec) 4 x 1013 L : Dimensions Core 33 in. dia, 35-1/4 in. high Reflector thickness 7-1/2 in. on side, ends open Shield thickness ' Approx, 7-1./2 ft concrete Over-all (reactor and heat-exchanger pits) 42 ftwx85ft 1x28 fth Contol Shim control Increase UF4 concentration in fuel Regulation ' One B4C absorber rod (2 in. OD 1-1/4 in. ID) Safety Three B4C absorber rods (2 in. OD, 1-1/4 in. ID) 47 ENGINEERING DESIGN General Description A briefdescription of the over-all ARE system is given to facilitate an understanding of the relationship between the various components of the system, and the components are described indetail. A schematic arrange- ment of the reactor system is shown in Figure 8. Heat is generated within the circulating fuel as it is pumped through the beryllium oxide-moderated core. The core is provided with a beryllium oxide side reflector that is housed, with the core, within an Inconel pres- sure shell. The reflector and pressure shell are cooled by a separate cool- ing circuitemploying a mixture of fluorides containing no uranium. Circu- lation of both the primaryand secondary circuit fluid mixtures is obtained by sump-type centrifugal pumps. Theheatis abstracted from the circulat- ing-fuel outside the core by means of four fuel-to-helium heat exchangers through which the fuel is circulated. The helium is then cooled by passing it through four helium-to-water heat exchangers, and the hot water is dis- charged. Helium flow rates in the fuel-to-helium heat exchanger are controlled by variable-speed hydraulic systems that drive the helium blowers. Con- trol of the helium flow rate in this manner permits smooth control.of re- actor power at any reactor temperature for which the nuclear controls are set, within the capacity of the heat removal system. At very low powers the temperature of the helium passing through the fuel heat exchanger close- ly approaches the fuel temperature. The system is designed so that when the rate of power generation is equalto or lessthan the power loss through insulation, electrically heated helium passes through the fuel-to-helium heat exchanger. Leakage monitoring is achieved bythe use of double-walled piping that provides an annulus through which helium is pumped outside the fuel. The helium is bledfrom the annulus at established stations and passed through sensory equipment. Several detectors, including the GE type-H leak de- tector, are under investigation, but the final decision regarding the type to be employed has not been made. PUMP HELIUM BLOWER =) HEAT EXCH ROD SURGE TANK ACTUATOR SURGE TANK PUMP - % - 2 o o O HELIUM Sl BLOWER 2 S 3 S 5 = 2 WATER o % he e 2 —_— L 18] ] T T T t HELIUM ] -— ’ F UEL — / REFLECTOR COOLANT FIGURE8. SCHEMATIC ARRANGEMENT OF THE AIRCRAFT REACTOR !t I m C c Z - EXPERIMENT 4 WATER 296¥1 "OMA 8y 49 In accordance with current concepts, preheating, as well as the addi- tion of heat during low power operation, will be accomplished by means of electrical heaters attached to the core pressure shelland to the outer pipe of the double-walled fuel piping. In the piping, the helium flow inthe annu~> lus acts as a heat conveyer, and transports heat tfo sections that are not, directly in contact with heaters. Fill and dump tanks for fuel are provided, and will be installed in a shielded pit adjacent to the reactor and heat exchanger pits. The tanks have not yet been designed in detail, but provision is being made for heat addition to maintain the fuel mixture in a molten condition and for removing from the fuel dump tank the afterheat evolved within dumped fuel after power operation. Control of the reactor is obtained by the following features: 1. Regulation-oneboronrod passesthroughthe corecen- terline with approximately 0. 75% Ak/k. 2. Safety-threeboronrods passthroughthe inner portion of the core with approximately 5. 0% Ak/k per rod. 3. Shim-shimming is achieved by varying the uranium tetrafluoride concentration in the circulating fuel. The external fluid circuit equipment is located in shielded pits adjacent to thereacfor pit. Two reactor pits are furnished to permit installation of a second reactor prior to complete decontamination of the pit used for the first reactor. Theheat disposal equipment is placed in a room outside the well-shielded reactor pit in order to allow servicing after draining and flushing fuel. The heat exchanger room is shielded, of course, because of the activity of the circulating fuel during and after power operation. Core Design The ARE reactor assembly consists of an Inconel pressure shell in which beryllium oxide moderator and reflector blocks are stacked and through which pass fuel tubes, reflector cooling tubes, and control assem- blies. Elevation and plan sections of this reactor are shown in Figures 9 and 10, respectively. The innermost region of the lattice is the core, which is a cylinder approximately 3 ft in dia and 3 ft long. The core is divided into six 60 deg sectors, each of which includes one serpentine fuel tube coil that passes through 11 stacks in series, as illustrated. The six serpentine coils are connected in parallel by means of external manifolds. DWG,%OZ Ri WELD BACK-UP RING BOTTOM HEAD CORE SUPPORT STRUCTURE CORE AND REFECTOR COOLANT INLET {SHOWN ROTATED 30 DEGREES FROM TRUE POSITION) REINFORCING PLATES FUEL OUTLET MANIFOLD ASSEMBLY CONTROL ROD ASSEMBLY THERMAL INSULATION CORE AND REFECTOR COOLANT QUTLET PIPE ( ROTATED 30 DEGREES FROM TRUE POSITION FUEL INLET MANIFOLD 1Y FUEL TUBE WELDING JOINT FUEL TUBES 1.235 0D, 0.060 WALL BOTTOM TUBE SHEET, 1 THICK TOP HEAD TOP TUBE SHEET COOLANT ANNULUS, Y46 THICK CORE AND REFLECTOR CAN, Y46 THICK COOLANT TUBE 03500 OD, 0.020 WALL NOTE : ALL DIMENSIONS IN INCHES 2 7.500 34.500 35.250 ACTIVE LATTICE FIGURE 9. ELEVATION SECTION OF THE EXPERIMENTAL REACTOR 09 51 Y-F7-203¢ SRene 3 REQ'D EQUALLY SPACED SUPPORT 8 LIFTING LUG ALL DIMENSIONS IN INCHES WIDE FOR i 6 SHELL COOLANT NOTE : BeO BLOCKS THICK £ ez, S A AR ST BeC CAN WALL A8 % AR e 0 N S FLOW ANNULUS PRESSURE > . IR \\\\\/ O SO T S VUSSP << <5 , D oSy \ REACTOR TUBES COOLANT 0500 OD x 0020 WALL REFLECTOR ROD ASS'Y 4 REQ'D, | AT CENTER & 3 EQUALLY SPACED ON 15" DIA. ¢ CONTROL 20 FUEL CIRCUIT, 6 REQUIRED IN PARALLEL, EACH CIRCUIT CONTAINS || PASSES THRU LATTICE IN SERIES. ACTIVE INSTRUMENTATION HOLE 2 REQ'D SECTION OF THE EXPERIMENTAL PLAN FIGURE IO 52 A reflector with a nominal thickness of 7.5 in. is located between the pressure shell and the cylindrical surface of the core. The reflector con- sists of beryllium oxide blocks similar to the moderator blocks but with 0.5 in. holes for the passage of thereflector coolant. Thereflector coolant is admitted at qne_eyd of the pressure shell, passes through the reflector, bathes the pressure shell, fills the moderator interstices, and exits at the other end of the pressure shell. Primary Cooling Circuit Hydrodynamic and thermodynamic considerations have influenced the ARE core design to such an extent that an appreciation of these problems appears tobe desirable priorto description of the primary cooling circuit. In acirculating-fuel reactor in whichthe heat is generated, and transported by the fuel, there may be a significant temperature gradient in thefuel and tube wall. In order toprevent anexcessive fuel tube wall temperature, the fuel flow in the tubes throughout the core must be turbulent. Analyses in- dicate that laminar flow would result infuel wall temperatures of the order of several hundred degrees F higher than the mixed mean outlet fuel tem- perature. The transitionbetween laminar flow and turbulent flow occurs in the Reynolds number range 2000 to 5000. Maintaining turbulent flow is so important that it was thought best to design for a Reynolds number of 10,000 and thereby allow for some deviation in flow rate or fluid physical properties. It is inherent in the nature of the circulating-fuel reactors and the fluids involved that the necessity of turbulent flow imposes more stringent design requirements on the Aircraft Reactor Experiment than onthe actual aircraft reactor. The power tobe abstracted from a circulating-fuelreac- tor is a function of the fuel volume flow—a higher powered reactor results if higher flow rates are employed. The viscosity of the fuel is such that the flow rates required by the aircraft reactor result in fluid flow well into the turbulent flow region. However, since flow rates are reduced to achieve the desired power level for the ARE, the fluid flow—if straight-through flow is employed—falls well into the non-turbulent flow region. In order to maintain turbulent flow, a multipassflow arrangement, as well as a low viscosity fuel, must be employed. 53 If the core isto include fuel tubes parallel to the core axis with a com- bination of series-parallel connections, then: R = RQNDL KV where R = Reynolds Number, p = fluid density, Q = fluid volumetric flow rate, N = number of tubes in series, D = tube diameter, L = reactor core length, pu = viscosity, V = fuel volume within reactor core. Analysis of the relationship between these design variables indicated that attdinment of the desired Reynolds number would require a few series passes, even with maximum feasible exploitation of the other variables. This being the case, and since the design problems involved in the use of many series passes appearedto be similar to those associated with theuse of a lesser number of series passes, it was decided to employ 11 series passes and avoid compromising the volumetric flow rate of the core fuel volume. Several means of achieving multipass series flow in the core were studied. An arrangement consisting of six long tubes, each bent so as to constitute 11 series passes, appearsto offer several important advantages. The six bent tubes require a total of 12 end connections to manifolds out- side the pressure shell. The fuel flow within, or adjacent to, the active lattice is confined to the interior of constant-diameter circular tubes, and: eliminates abrupt flow cross section transitions with indeterminate flow patterns. The six parallel tubes are almost symmetrical relative to the core centerline, and obviate the need for individual orificing. The series fuel tube arrangement with vertical tubes involves problems of gas removal and liquid drainage. Purging of bubbles from the coils has been investigated experimentally and analytically. Both investigations in- dicate that at the design flow velocities bubbles and gas traps should be swept out of the core. 54 Secondary Cooling Circuit The use of aliquidcircuit within the pressure shell, other than the fuel circuit, is desired (1) to maintain the pressure shell essentially isothermal at a temperature of approximately 11500F, (2) to cool the reflector with a fluid other than circulating fuel, and (3) to fill moderator interstices and other internal voids with a liquid maintained at a pressure higher than the fuel pressure to prevent a fuel tube leak from adding reactive material to the core. The liquid to be employed must be compatible with the structural ma- terial under dynamic conditions and should cause no undesirable reaction with the fuel in the event of a leak in the separating wall. The fluoride mixture NaF-KF-ZrF4 {mole % 36-18-46), i.e., the fuel carrier without the UF, is well suited for this purpose. The possibility of cooling the moderator by positive flow of this fluid through the moderator interstices has been studied. It has been found, however, that the mass flow through a gap of constant periphery varies as the third power of the gap width (in the laminar flow region, with fixed pressure drop). Temperature analyses with postulated gap width distribu- tions indicated that gross temperature inequalities could exist throughout the moderator because of tolerance accumulations affecting interstice widths. Accordingly, no attempt has been made to have positive flow through the moderator blocks, but the presence of the 'stagnant' salta- round the moderator permits moderator heat to be conducted to the fuel stream., Structural Assembly The mechanical design of the Inconel pressure shell has been the sub- ject of coordination between ORNL and the fabricator, Lukenweld Co., Inc., and it is expected that the shell as constructed will resemble very closely that illustrated in Figure 9. The serpentine tubes maybe bent in flat planes,. with welded connections employed from plane toplane, or ""U" tubés welded to straight core tubes may be employed throughout. These alternatives have been investigated by commercial, tube-bending concerns, and one con- cern has contracted to furnish the completed coils. Core Temperature Distribution The circulating fuel in the ARE reactor core flows in paraliel through six serpentine tubes, and each tube traverses the core 11 times. In pass- ing through the core, the fuel is heated from 1150 to 1500°F, 55 The two sources of heat for the fuelcirculating throughthe reactor are internalheat generation - resulting from fission and gamma-ray absorption - and heat transferred to the fuel from the remainder of the core. The heat transferred from the remainder of the core is produced both from gamma- ray absorption in the moderator and the parasitic core material and from neutron slowing-down inthe moderator. All thisheat is transferred to the circulating fuel and results in a temperature gradient across the fuel, mod- erator, and parasitic material. The resultant temperature profile across the fuel and adjacent mode- rator block and parasitic material is shown in Figure 11 for both the en- trance and exit condition. There are four distinct media in each tempera- ture profile: the fuel, the fuel tube wall, the stagnant salt between the fuel tube and the moderator block, and the beryllium oxide moderator block. The maximum tube wall temperature in the reactor is about 15000F. The temperature differential in the beryllium oxide block is entirely due to internal heat generation (heat transfer between adjacent beryllium oxide blocks was neglected). The temperature differentials in the stagnant salt and fuel tube wall are almost entirely caused by transferring the heat generated externally, since the effect of internal heat generation is small. The temperature differential across the fuel, which occurs mainly in the boundary and buffer layers is principally caused by the temperature differ- ential associated with transferring into the fuel the heat generated in the moderator and parasitic material. The contribution to the temperature differential resulting from the variation of fuel residence times with radius is only about 5CF. The radial and axial distributions of fission, gamma, and neutron heating used in this analysis were taken from data reported in the section on "Reactor Physics. " External Fluid Circuit The ARE fluid circuit is intended tohandle toxic and corrosive fluorides at 1500°F, and requires a considerable amount of design and developmental effort. Where possible, commercially available components are employed but pumps, heat exchangers, and certain other components are being con- structed especially for use with the ARE. The locations of the major com- ponents of the system, includingthe reactor pits, heat exchanger room, re- actor, and two heat disposal loops are illustrated in Figure 12. The heat exchanger room is shielded from the reactor pits to permit servicing after fuel drainage and flushing. The shield is designed to minimize activation of fluid circuit structure by reactor neutrons during power operation and to attenuate reactor post-shutdown gammas to a level permitting access to fluid circuit components after shutdown. S NN N N N NN 51340 1675 E 7)) w & & P P < - o [+ 4 1 4 Ll w Q. Q. s = W il - - - \ o w - i / 2 = = - O " - | < < 1so0 1500 3 5 5 x o a Q o. g < ~<—FUEL —| -~ o le TUBE STAGNANT _|| WALL SALT - e Beo—> FIGURE 1. TEMPERATURE PROFILE IN FUEL TUBE AND MODERATOR UNCLASSIFIED DWG. D-Y-F7207R{ INERT SALT COCLING SYSTEM NO. A = EXCHANGER He EXCH. PRIMARY COOING SYSTEM NO, 1 SCOURCE TEST PIT L HEAT EXCHANGER ROOM - SURGE TANK COOLANT TO He EXCHANGER w PRIMARY COOLING —— — i —— . e—iints 8. LEGEND SYSTEM NOQ, 2 —— - s == e COQOLANT FROM SOURCE — v sttt = ———— COQLANT TO SCURCE v/ s /= INERT SALT FROM SOURCE - - — / INERT SALT TO SOQURCE co oo~ HELIUM FROM RODS o O == HELIUM TO RODS —_— W == WATER (N —_—g WO = WATER OUT ar—————e L. 1N i e sy e = (L OUT STACKED BLOCKS ROD COOCLING BLOWERS INERT SALT COOLING SYSTEM NOQ. 2 e —— 24 28ft-0in, FIGURE 12. FLUID CIRCUIT EQUIPMENT ARRANGEMENT LS 58 Primary Cooling System Fuelflows from the six core tubes to a common header, and a common trunk line conveys the fuel into the heat exchanger room. The fuel then divides into two completely independent loops, each capable of dissipating 1500 kw. Within each loop the fuel divides into two parallel 750 kw fuel-to- helium heat exchangers, passes through a centrifugal pump, and then to a common return trunk. A line diagram of the primary cooling circuit is shown in Figure 13. In an individual loop, helium flows from a centrifugal blower through a fuel-to-helium exchanger in which its temperature is increased from 250 to 750°F. After the helium is cooled to 2500F inahelium-to-water ex- changer, itpassesthrough another heating and cooling cycle and is returned fo the blower. This so-called 'double-sandwich' arrangement permits two helium heat transports per cycle and thereby halves the helium flow rate required through the blower and ducting for a specified power and temper- ature rise, Secondary Cooling System The reflector andpressure shell cooling system, as in the case of the fuel system, is divided into two heat disposal loops, each capable of han- dling the load associated with 1500 kw reactor power. Heat is conveyed from the primary coolant to helium and then to the water sink. A line diagram of the secondary cooling circuit is shown in Figure 14, Monitoring Circuit Alllines and components containing fuel or reflector coolant are double- jacketed and helium passes throughthe annulus. The helium pumping head is maintained by drawinghelium from the system, cooling it, and admitting it to rotary-type compressors at various points inthe system. Monitoring for fluoride leaks into the helium is achieved by passing helium samples through halogen detectors. Pumps The pumps to be used in the fuel circuit and in the moderator coolant circuit are vertical-shaft, tangential-discharge, centrifugal pumps employ- ing a gas sealand a maintained liquid level. A pump of this type, designed specifically for the ARE, will provide flow of 41 gpm with a 40 ft head (Fig- ure 15). Electromagnetic pumps cannot be used because of the poor elec- 2o TR e o PIEL Y Py : 59 To Dram - an i Dump Tank a2aneg U_-_]__-___ - - __-_L-_! ! o= ———— (e m——n - —— — —— | : o= Fuel - He : I 1 : He ~wWater L |; Fue!- He 1 | i : ; } HHe -Wuh; ! | L_..—_-._ _peg From vater Surly, e p— From Weter Sorly FIGURE 13. PRIMARY GOOLING CGIRCUIT 60 WAl el b & ,{C'e. .y S il rom Waler 5 -—-——-—N—-B—F vpply From V‘_.'atcr S He - Water {—— —a—-.- Source - - - - FIGURE 14. SECONDARY COOLING CIRCUIT UNCLASSIFIED DWG. ESK-15329aR1 oiL ouT |. SHAFT, TYPE—316 STAINLESS STEEL. 8. SHAFT LINER, ANY 300 SERIES STAINLESS STEEL. 2. SHAFT LINER SPACER, ANY 300—SERIES STAINLESS STEEL. 9. SINGLE-ROW RADIAL BALL BEARINGS 3. "0" RING, MATERIAL TO BE RECOMMENDED BY SEAL MANU- 10. SLINGER, TYPE- 3i6 STAINLESS STEEL FACTURER. 11, SPEGIAL GRAPHITE RING ,ALTERNATE WITH SEAL 4. IMPELLER, TYPE—3i6 CORROSION— RESISTANT STEEL ARRANGEMENT. GASTING. 5. PRESSURE SLEEVE, ANY 300-SERIES STAINLESS STEEL. I2. GASKET , GOPPER . 6. SEAL FACE RING, ONTARIO STEEL. 13. PUMP BODY ASSEMBLY. 7. BEARING HOUSING, TYPE— 316 CORROSION—RESISTANT 14.IMPELLER CASING ASSEMBLY. STEEL. 15. BELLOWS SHAFT SEAL, FULTON SYLPHON BB8ISR. 6. BELLOWS SHAFT SEAL, FULTON SYLPHON 8816,7R. FIGURE 15. ARE CENTRIFUGAL PUMP 19 62 trical conductivity of the fluoride fuels. It is possible that a centrifugal pump employing a frozen fluoride seal=a technique in which fluorides are frozen about the pump shaft effecting a seal—may be adapted; however, a gas seal pump has been satisfactorily operated at temperatures above 12000F for several hundred hours. The gas seal 1sformed by a floating graphite ring located between the stationary nose on a bellows seal and the hardened rotating nose on the shaft. The level of the pumped fluid inthe expansion chamber is maintsited by gas pressure so that the fluid cannot come in contactwith the seal. The primarypump seal enclosing the bearings carries no appreciable pressure differential and, consequently, requires that an oil-circulating system a- bove the seal be pressurized to balance that of the circulating system be- low the seal. The circulating oil serves to cool the shaft bearing and lower seal. In addition, internal cooling of the pump structure is obtained by the circulation of helium. Another advantage of this pump, in addition to the proved sealing de- vice, isthatthe entire rotating assembly canbe assembled outside the pump casing and in event of failure can be installed with relative ease. None of the liquid lines need be broken to accomplish this replacement nor will any of the welded joints in the pump permit fuel holdup. All mechanical pumps that seal directly against system pressure are provided with a monitoring space to permit the detection of leakage, should such occur. The pumps will be constructed of type 316 stainless steel except that those parts in contact with the fluid stream will be of Inconel. Heat Exchangers The fuel-to-helium heat exchangers are on order with the Griscom- Russell Co., who designed them to ORNL specifications. The heat ex- changers are of the cross-flow type with the fuel flowing through 1 in,OD, 0.109 in. wall Inconel tubing. The tubing is finned on the gas side with helically wound, stainless steel strip. Each heat exchanger includes five parallel tube banks; each bank consists of seven to nine horizontal tubes in series with integral return bends. The fuel inlet and outlet headers are constructed of 4 in.pipe to which the five tube banks are welded. The helium -to-water heat exchangers also are of the finned-tube cross- flow type, but steel tubes with copper fins are used. 63 Preheating Sys‘tefn . The relatively high melting point of the circulating fuel and the second- ary coolant (around 500°C for the NaF-KF-ZrF,-UF, fuel-with a composi- tion in mole % of 34.7-17.4-44.4-3.5) requires that all equipment within which these coolants would be circulated be heated to permit loading, un- loading, and also the possibility of low power operation. Reactor Assembly The insulation heat-leakage at operating temperature has been calcu- lated to be approximately 15 kw, and the heat capacity of theassemblyhas been estimated at 4000 Btu/OF. After having studied various preheating methods, including pumping hot fluids through the fuel passages or through the moderator insterstices, it was decided that the simplest method was the application of electrical resistance heaters to the outside of the pressure shell, with heat transport to the inner regions by conduction. With a con- stant heat input of 14kw in addition to the insulationlosses, it is estimated that the assembly can be heated from room temperature to 12000F in ap- proximately four days. | Piping The heat capacity of the piping is small relative to the piping-insula- tion leakage. Consequently, the power required to preheat the piping is substantially the same as the power required to maintain itat temperature. If the fluid circuit pipes are empty, as during preheating, the pipe resist- ance to axial heatflow is high, and the application of heat to specific points on the pipe surface tends to cause large temperature valleys to exist be- tween the points of heat application. By attaching electrical resistance heaters to the outer pipe of the double-walled piping, however, the flowing helium acts as a thermal conveyer and facilitates axial heat transport. Accordingly, substantially isothermal helium contacts the inner pipe, and enables it to be heated uniformly. 3 Heat Exchangers The fuel-to-helium heat exchangers require preheating prior to filling and require the addition of heat to compensate for thermal leakage during filling and after filling at any time that the fuel pumps are inoperative. These heat exchangers have a total of approximately 40 £t of exposed free- flow area (bothupstream and downstream) that, ifallowed to radiate to ad- 64 jacent cold structure. would dissipate approximately 40 kw with the tubes at 13250F. Accordingly, it appears necessary to include radiation barriers in the form of gates that can be iowered during warm-up, zero, or low power operation. When these radiation barriers are in position, the fuel- to-helium heat exchangers are enclosed and may be preheated by means of hot helium supplied by the external piping annuli. Reactor Controls A significant characteristic of the entire ARE-ANP control philosophy is that the power extracted from the reactor is determined solely by that part of the over-all power plant external to the reactor. In the power re- gime, reactor conirol rods do not influence the steady-state power produc- tion. Conversely. the power extractionfrom the reactor does not influence the steady-state reactivity of the reactor. A reactor that satisfies the cri- terion given here is almost certain to have these control characteristics regardless of whether or not circulating fuel is used. The fact should not be overlooked that if the reactor behavior for a low power level differs radically from the behavior of the aircraft reactor, there is less incentive for building an ARE. Much of the data obtained from such a low power experiment would be functionally dependent upon power level. However, the two basic characteristics of the circulating-fuel re- actor are: (1) the reactor operating temperature is independent of the ex- ternal pumps and heat exchanger; and(2) the power extracted is independent of reactor control rod motion. Essentially then, the objective of the ARE control work is to provide satisfactory reactor control and external system control for the entire ex- periment. In addition to creating satisfactory control for the experiment, the system should be such that control data can be obtained for use in de- signing the aircraft reactor. Thesedata will be essentially time-response data for perturbations in reactor temperature and external power load, as wellas transient and steady-state data on the coupling between reactor and load. There are three separate parts to the control system for the ARE: (1) shim control (2) regulating system, and (3) safety system. Since these - three divisions perform distinctive functions and since each is almost, if not entirely autonomous, each will be treated separately here. Reactivity calculations of the various control mechanisms are reported under 'Re- actor Physics.' Additional information involving the reactor controls is discussed under 'Reactor Operation. " 65 Shim Control The reactor will be made critical only after the fuel-carrying liquid has beenbrought up to a temperature of approximately 12000F. The liquid will then be maintained at as nearly constant temperature over the entire primary circuit loop as the external heating system will allow. After the system has been checked for leaks and found to be tight, a fuel enrichment system will be used to bring the reactor to critical at the temperature of 12000F. This fuel-enrichment system comprises the shim control of the reactor and is designed only to add uranium to the carrier; there is no plan to try to go subcritical by fuel depletion. At the surge tank, located between the heat exchanger and the reactor inlet, a positive-volume fuel-injection system is provided. This fuel-in- jection operation is accomplished by means of a mechanism shown in Fig- ure 16. A small fuel line first carries the fuel from the storage tank to the transfer tank and thence to the surge tank. The enriched fuel is held in the transfer tank where it is weighed, and when the gas pressure over the liquid in the tank is raised, fuel is forced through the connecting line intothe system surgetank. The transfer tank ishung from a beam balance sothatan accuratedetermination of the weight of fuel is directly obtainable. This method of fuel addition permits the addition of known amounts of fuel, and the magnitude of these guantities can be controlled over a con- siderable range as a critical loading is approached. By introducing the fuel into the circulating system just before the liquid enters the reactor, safe loading of the system is assured. Should a slight excess of fuel be added to part of the fuel stream, the reactor would give an indication of supercriticality whenthe enriched liquid eniered the lattice and before the uranium concentration of the entire fuel circuithad been increased. If this occured, the fuel addition could be stopped and subsequent mixing of the supercharged liquid with subcritical fuel would lower the reactivity. After eachfuel injection, which will be made with the three safety rods cocked and the regulating rod in inid-position, the regulating rod will be withdrawn in accordance withregular procedure for critical loading of any reactor. Should this shimmingbe slightly overdone, which would result in the reactor going on a long positive period beyond the compensating worth of the regulating rod when fully inserted, the temperature of the fuel-coolant can be raised enough to restore the reactor tothe critical condition. After | | I | L YENT Line | g He Sverty e Suert l —t Vewr Live | g | | | | I { } He Sveely ! TRAMSFER | TANK | | | | — — ¥ PR Twermocovrif WeLL STorAGE Tank SurGE TANK FIGURE 16 FUEL INJECTION SYSTEM 3 HEATER S b KEMOTE ALY & 0] FREEZE Vacve D WerGH CELL [P\ Fréssure \__/ INDICATOR TEMPERATURE INDICATOR WEeIGHT INODICATOR G3141SSVIONN 99 e o et e P s ae e 67 the loading is accomplished at the starting temperature of approximately 1200°F mean fuel temperature, additional injections can be made when it is desired to raise the mean fueltemperature. If it becomes necessary to lower the mean fuel temperature before termination of the experiment, the safety rods can be lowered slightly into the core. Although not planned in the operation of the reactor, such action is possible if conditions should demand. Regulating System The reason for providing a mechanical regulating system is that the magnitude of the reactivity change effected bythe expansion of the fuel can- not be accurately predetermined. To supplement the negative temperature coefficient, in case it is of insufficient magnitude, a single absorber rod having a total value of approximately 0. 75 % Ak/k is provided inthe center of the reactor. The absorber rod is located in a 3 in dia hole that runs through the pressure shell and core in the center of the lattice. By means of a direct drive, a 2 in.dia boron absorber rod is inserted into this hole. The rod is moved by one of two possible drive motors. One of the motors is a 60-cycle reversible motor of constant speedthat is manually actuated by a zero-po- sition, momentary-contact, reversing switch located on the operating con- sole. The motordrive is used to make manual adjustments during the crit- ical experiment and during the rod calibration operation. It is also avail- able for manual control of the reactor during the power regime. The other drive, which can be independently used to actuate the reg- ulating rod, is a d-c servo motor of the same design as that used on the MTR and also on the MTR mockup - the LITR. 9 This drive can never be operated manually, and it is actuated at all times by the control function system in accordance with the operation of the reactor, as subsequently described. The servodrive will run during the entire reactor experiment. However, exceptfor the brief period of time required for the dynamic test of the reactor, the servo will be disconnected from the rod and used only to drive a monitor selsyn to aid in following the behavior of the reactor while it is under manual control. The servosystem is arranged so that the reactivity changes introduced by the regulating rod behave like a negative temperature coefficient asso- ciated with the fuel. When the regulating rod is connected to the servo drive and temperature or flux perturbations are introduced into the reactor, control of the reactor will be accomplished. 9. T. E. Cole, E. E. St. John, and S. H. Hanauer, 'The MTR Safety System and Its Components,' ORNL-1139, April 25, 1952. 68 During normal operation of the reactor, when power changes are being made at a slow rate, the reactor will be manually controlled. Improper manipulation of the manual control, which would result in getting the re- actor on too short a period, is safeguarded by the cocked safety system plus the coefficient of reactivity associated with the expanding fuel. Pre- cautions inthe form of limit switches, interlocks, alarms, etc., in accord- ance with previous reactor installations, are provided to further minimize the opportunity for improper manipulation of the manual control. The regulating rod drives are located on the concrete slab directly over the reactor and are accessible to personnel at all times. The rod linkage goes through holes in the concrete slaband provides means for po- sitioning the absorber portion of the rod in the center hole provided in the reactor. The general arrangement of the regulating rod and drive mech- anism is shown in Figures 17 and 18. Since the regulating rod will run within the core while the reactor is running at full power, the rod must be cooled. This is accomplished by the forced circulation of helium in a closed loop down around the rod. The nuclear instrumentation (fission chambers, ion chambers, etc.) and the electronic components (preamplifiers, power amplifiers, etc. ) are copies of similar instruments and gear in use on other reactors. There- fore it is notnecessary to describe these items indetail. Since the fission chambers will be located in a high-temperature region within the reactor reflector, helium cooling will be used to keep the chambers within a safe temperature range. Safety System Three safety rods for the ARE are located on 120deg points at a radius of 7.5 in. from the center of the reactor. The rods insertinto 3 in.sleeves that passthrough the reactor shell and core vertically from top to bottom. The rods are of hot-pressed boron carbide slugs canned in stainless steel sections slipped over a flexible tube. Sections of the rod are cored to per- mit passage of helium down through the rod for cooling purposes. Helium also passes down around the rod in the space between the rod can and the core sleeve so that cooling of the outside surface and the inside of the rod is accomplished by a parallel helium flow. Located above the reactor, on top of the reactor pit, are the actuating mechanisms for each rod. An actuator is provided for each rod but all actuators are energized simultaneously and the maxium speed of withdrawal of the rods is fixed by the speed of the constant speed a-c motor used to actuate the driving mechanism. e 69 Y-F7-212, PART 1 L5 Ne 1 sec. PERIOD FAST SCRAM N>LS N 1 SEC. PERICD -~ - INSBRT 3AFRTY RODS - - -~ - HISH TEMPR Tz I MANUAL SCRAM ] | MANUAL SCRAM | [mwum. ScpAM ] | MASNET POWER % TO BE CHANGED TO LowW LEVEL SCRAM FOR START P X M NEUTRON LavEl. Np = DESEN PoiNT NEUTRON Lavel. - FIGURE 19. SAFETY INTER LOCK AND SCRAM CIRCUIT L] 73 b e d-c circuit iheluding indépendent buses, switches, and a-c¢/d-c motor gen- erator sets. Theuse of direct current for these circuits permits convenient speed control where required and the use of battery sets floating on the line to safeguard against outside power failures. Alternating current instru- ments are fed from the d-c circuits via d-c/a-c motor generator sets to enable these instruments to derive their power from the batteries in an emergency. This electrical power circuit is shown in Figure 20. ~E, gy o The heat capacity of the system is large relative to the heat leakage rate sothat no large temperature loss will be incurred during a power in- terruption of 2 to 3 hours. Consequently, it is not necessary to employ battery power for system heat addition. System heaters therefore are connected directly to the 440-v a-c circuit. Similarly, the building crane and certain other items need not be operative during any short period in which operation from battery power is required, and these componentsare supplied only by the 440-v a-c system. Instrumentation Since a basic purpose of the Aircraft Reactor Experiment is the ac- quisition of experimental data, the importance of complete and reliable - instrumentation cannot be over-emphasized. Most ARE process instru- mentation is intended to observe and record rotational speeds, flow rates, temperatures, pressures, or liquid levels. The various stations and meas- urements desired around the ARE fluids circuits are shown in Figure 21. Since most commercially available equipment for performing these functions is limited totemperatures considerably lower than the minimum operating temperature of the ARE, an extensive devélopmental program has been carried out to either produce new instruments or remove the temperature limitations placed on existing instrumentation. Development has now pro- gressed to the point where levels can be controlled inhigh-temperature fluorides with great reliability for extended periods of more than 2000 hours. Although equipment is still lacking for the calibration of flow-measuring devices, such devices are now available and have operated for extended periods. Thedegree of accuracy eventually attainable will depend upon the procurement of accurate calibration equipment. Pressures are reliably measured at many points in fluoride forced-circulation systems with both commercial and locally developed instrumentation. Flow Measurement Electromagnetic flowmeters have proved to be very reliable when used with systems circulating sodium but are not applicable with fluorides, which have a very low electrical conductivity. ~J o 250 ErA Beng. Marn /8. 8rv/480v. 38 Berdeer GOOA LA0V 4.C. 3P Powre IiSTRIBUTION FANEL BOARD — 120, 208V AC. AP LienwTing & Power DiSTRIBUTION PANELACARD SuBSTATION P GOOV AcC. . /\’D -—_) C 7 SoL/b NEYTRAL ) 200 kra 480/ 7120/ 204V —— ) . . . . . . . v 4C. 30 & wiee . ) Dey rree j J J ) TRANSFORAMER - . . . T x X . - ¢ Tl e = 3 € 5 v W O \ w @ L5 = 2 N boupPs £ £ o ¥ 3 = N 7 S AL BEEd O % 3 3 3 : a2 M“w — 2 | - - L3236, L8 50 K 8 % 5 & g m £33 ! T wmrmmmwm 3 I= w % 8§ 0 %5 5 M > Z ay o o = = 3 3 y ¥ £ 0o 2 ) I~ & ‘ 553555034 v S8V 232 Ers. Loy ok X1 e B e 3 g & & 232 = = 5§ L K P24 2 38 5K s3I .23 8 3 3 L _ _ s ¥ £ f sz £ 5, ¢ 2 s, 2 5 ! vl 0 9 5 ¥ 2 @ ® < ¥ 2 £ £ a © _ i _¥ P n@ Y £ 3 N = 7L 3 - = 8 0 & o " i o e SRR AR E R EE R R R 53y g 3 w ’ w a o = g T O = 2z =z 3 > 2 “ fa flw _ fl | M td P rPEaBST4AEAEYLEZEELS 8 &>% &3 A _h_v/‘ Rt : * ! j - - I e T T SN T S & g s ) . o . | Af [ PEIELT A48 9 &4 HeE & | . - b m O . i & §— -4 - - ST i 1@ - &) .}@ ey o) 7 .:é-r_ s o ] N | | | | i | | | | | | | | | | | | | ] e Detads See ORML Dug DAL For Qeterve 4 for Hest Brch. Pt B aE uean ey strumeng Auvxtlary Systems. Z All Coatrols Are Located on in i- Far Continuation of Cirepits See Panels or Congole. NeTes INSTRUMENTATION CIRCUIT. FLUID FIGURE 2] i 77 A venturi section, with associated pressure sensing instruments, ap- pears to be the most attractive device for measuring flow in a high-tempera- ture fluoride system because of low head loss. The extrapolation of water- calibrated venturi measurements to fluorides is reasonably accurate. Pressure Measurement Todate, nopressure-sensitive instrument has been developed or found commercially available that will operate at the temperatures to be encoun- tered in the ARE fluid circuits. The highest temperature rating of any commercial instrument known is 450°F. As a consequence, all pressure- measuring devices have used trapped gas, and as a result the measured pressure isthat at the gas-liquid interface instead of liquid stream. How- ever, this error inpressure measurement can be minimized by calibrating each set of instruments and associated gas traps with accurate weigh-tank calibration equipment. Bourdon-tube pressure gages are sufficiently reliable for rough pres- sure measurements but are notdesirable ifa high degree of accuracy is to be obtained. Bourdon-tube gages, which indicate the differential pressures directly, are commercially available. Null-balance type of pressure gages indicate differential pressure on a single gage tothe nearest hundredth of a pound per square inch. Although this instrument is temperature-limited, it was connectedto the above-de- scribed gas trap tanks and operated successfully at room temperature. Temperature Measurement Temperatures to 20009F are measured by means of Chromel-Alumel thermocouples and recorded on multipoint Brown temperature recorders. It has been found that the accuracy of temperature measurement depends to a great extent on the fabrication and installation of thermocouples. Two methods of thermocouple attachment are used atpresent: (1)pulse welding each thermocouple wire to the outside surface of pipes or containers, and (2) forming a beaded junction of the two thermocouple wires. Ineachmethod the thermocouple is given additional rigidity by tying it down with nichrome wire. With either of these methods a temperature error is introduced if the thermocouple is not insulated from ambient air or other adjacent mate- rial. A borax coating has been found to form an airtight coating that pre- vents further oxidation. Tests conducted with thermocouples placed in deep wells centered in the flowing stream and other couples directly opposite onthe external sur- 78 face of the pipe have revealed that there is no more than a 15 deg temper- ature drop from the internal thermocouple to that placed on the outside. Thermocouples made by beading the wires are accurate within £ 10F, The Brown temperature recorders, with careful adjustment, will re- cord temperatures accurately within £ 1°F over the full range from 0 to 2000°9F, They do, however, show atendencyto drift after extended periods of operation and cause errors of the order of 5C0F., This necessitates fre- quent calibration of the instruments; therefore a program for routine in- spectionhas been setup that should reveal any instrumentation errors soon after their occurrence. Liquid Level Indicators and Controls Four level-indicating devices have proved satisfactory for use in high- temperature liquids. The conventional probe type of level control operates a relay or solenoid when the level of the conducting fluid rises to short out the probe. Such a level control has proved most satisfactory with molten fluorides. Another level indicator using the principle of a resonant cavity has been designed and simulated experiments indicate that this instrument may give satisfactory performance. The variable inductance level indicator consists of small coil wound on the outside of the tank in which the level is controlled, and a tapered iron core is mounted on a float inside the tank. The change in level raises and lowers the tapered core inside the coiland altersthe inductance. Tests made with the iron core at room temperature demonstrate that it gives a linear response that canbe directly correlated to a level inside the system. The fourth level indicator consists of a small tube that extends to the bottom of the tank in which the levelis to be controlled. Gas flowing through this tube bubbles through the liquid. The pressure required to maintain a constant flow of gas through the liquid is directly related to the height of the liquid level above the end of the tube and is measured by a manometer connected betweenthe gastube and the gas space above the free surface of the liquid. Suchan instrument can be used for both indicating and control- ling liquid levels. 79 REACTOR PHYSICS The total uranium investment of the ARE comes out reasonably low, in spite of the fact that thereactor hasnot beenoptimized to give the lowest uranium investment. Engineering considerations and expediences designed to shorten the construction time were in many cases regarded as more important than optimization. Little concernis felt about the possibility of self-sustained oscillations. Because of the low power theywould be slow enough so that a servosystem could damp them out. Furthermore almost all delayed neutrons are given off in the reactor, and are available for the damping of oscillation. Statics The static calculations are based on the multigroup method, as de- cribed in references 10 and 11. For reflected reactors this method is at présent available only for spherical geomeiry. However, the geometry of the ARE involves a cylindrical core with a side reflector and a partial end reflector, the side reflector and end reflectors being of somewhat different composition and thickness. In order to get kg for this geometry, two reflected reactor calculations for spherical geometry were made. The first of these calculations used a core of a radius and composition equal to the radius and composition of the actual cylindrical core, and a reflector composition and thickness equal to the composition and thick- ness of the side reflector. The second calculation used a core of the com- position of actual core, a core radius equal todhZ+ 1 RZ, (where 2h is the height of the actual cylinder core and R is its radius), and a reflector corresponding in composition and thickness to the end reflector. A few bare reactor calculations for spherical geometry yield a curve, giving ke¢r versus the buckling, or reactor radius. With this curve the k s values from each of the above reflected reactor calculations were com- pared. This gave for each reflected reactor an equivalent bare reactor radius. By subtracting from each of these equivalent radii the corres- ponding radius assumed for the core, the reflector savings is obtained. The reflector saving for the side reflector is then added to the cylinder radius, and the savings for the twoend reflectors are added to the cylinder axis. An equivalent cylinder is thus obtained, for which a bare reactor 10. M. J. Nielson, "Bare Pile Adjoint Solution, ' Y-F10-18, Oct.27, 1950 11. D. K. Holmes, "The Multigroup Method as Used by the ANP Physics Group, "' ANP-58, Feb. 15, 1951. 80 calculation can easily be performed. A number of these bare reactor cal- culations for different amounts of uranium yields a curve from which the amount of uranium required for any given k. ¢ is specified. The above calculation is carried out assuming homogeneous distribu- tion of the materials in the core and the same for the reflector. The effects of "lumping' and self-shielding of the material are computed afterwards. If they appear important, an iteration of the multigroup calculation is usually necessary. The multigroup calculations depend, of course, on the cross-section data. Fortunately, the ARE has about two-thirds of the fissions occurring at thermal energy and is, for this reason, not too sensitive to the uncer- tainties of high energy cross-section data, including the ratio of the fission cross-section to the absorption cross-section of U?%%, The dataused were essentially preliminary data of the AEC Neutron Cross-Section Advisory Group and are probably inagreement with those published in BNL-710 and AECU-2040. Whether or not any minor discrepancies exist has not been determined as the publications were just received. The fissioning power distribution for the equivalent spherical reactor is obtained from the multigroup calculation. For the cylinder, the radial distribution in a plane through the center and perpendicular to the axis is assumed to be the same as that for the equivalent sphere. The distribu- tion along the cylinder axis is assumed to be a cosine distribution, reaching zero at a point beyond the core boundary, by a distance given by reflector savings plus extrapolation distance. The fissioning power is, of course, proportional to the neutron flux so that both can berepresented by the same curve. The radial distribution curve is multiplied by a normalizing factor, so that the total power (or neutron flux) per cc averaged over the whole core comes out to be one. That is, the points on the curve are maultiplied by 2w and the radius r. The result is integrated over r, and this integral is divided by the cross- sectional area of the core to get the average over the cross-section; this average is then multiplied by 2/m to get the average in the axial direction under the assumed axial cosine distribution. The final result is one, due to the choice of the normalizing factor. The axial power distribution can be normalized in the same way. The multigroup calculation alsc gives the neutron spectrum, from which the power generation by moderation of neutrons can be computed. The spatial distribution of fissions and absorptions - and hence the sources of gammas - is derived from the spatial distribution of the neutron flux. The Y heating may then be determined. The energy contained in the motion of the fission fragments is given off entirely in the fuel tubes. Neutron mod- eration creates heat essentially in the moderator. Gamma-ray heat is generated in all parts of the reactor, in accordance with their energy ab- sorption coefficients (see for instance ORNL-421). 81 From the neutron spectra, one also obtains the neutron leakage from the reactor surface, which is of some importance to the calculation of the activation of external structure and so on. The leakage from the ends is, of course, obtained from the spherical calculation using the end reflector composition and thickness, and the leakage from the sides is obtained using the side reflector data. A coefficient of reactivity is the change of k with the change in some parameter of the reactor. For instance, the temperature coefficient of reactivity is the change of reactivity with reactor temperature. Reactivity coefficients are obtained by performing the multigroup calculations for at least two values of the parameters in question, and taking the ratio of the reactivity change over the change in the parameter. It should be noted that this procedure will not give the change of re- activity with fuel temperature correctly if a great fraction of the fissions or absorptions occur in the high energy ''Doppler' region. However, the ARE is largely thermal, and no concern was felt regarding the Doppler effect. In the past, the multigroup method has given good agreement with ex- perimental data regarding critical mass. It is not absolutely certain that this implies that the method will give power and spectral distribution cor- rectly, in particular since the method is obviously based on a number of approximations. For this reason, critical experiments, now in prepara- tion, will be carried out prior to the operation of the ARE. Reactor Volume Fractions The design data of the reactor are given in Table 7, pp.- 45-46. For the purpose of the reactor calculation, the total volume of the reactor was broken down into five regions which are listed in Table 8, together with their compositions. The "solvent" for the UFy was the fluoride mixture NaF-KF-ZrF,(mole % 36-18-46 respectively). The fuel now pcstulated for the reactor has a somewhat different composition, but the computation can be performed in an analogous manner for the new fuel. Critical Mass Total Uranium Investment The kq; required by the ARE is 1.0415 as computed in Table 9. The calculations described above then show that 27.6 lbs of U2 are required in the reacting volume. Table 10 summarizes the computation of the ""hest guess'' for the total uranium requirement, the result being 120 1bs. The reflected reactor has about two-thirds thermal fissions and a leak- age-to-absorption ratio of about 1:3. TABLES SUMMARY OF DESIGN DATA OF THE EXPERIMENTAL REACTOR No. fuel tubes 66 Core length - 35.250 in, Core diameter 33.000 in, _ Fuel tube size 1.235 in, O.D. x 0,060 in, wall No. control rods 4 Item Volume (ft3) Volume (%) Space Between Pressure Shell and Top of Core Fuel . 0,134 6.57 Structure . 0,058 2.84 Inert salts 1.815 88.97 Control rod 0,033 1,62 Totals 2.040 100,00 Core BeO 14.235 81.63 Fuel 1.315 7.54 Inert salts 1.225 7.03 Structure : 0,378 2.16 Control rod 0,286 1.64 Totals 17.439 100,00 Control Rod Structure In Core Structure . b.030 0.17 BeO 0.24 ' 1.38 He 0016 _0.09 - Totals 0.286 1.64 Space Between Pressure Shell and Bottom of Core Fuel 0,134 6.57 Inert salts 1.450 71.07 Structure 0.423 20.74 Control Rod 0.033 1.62 Totals 2.040 100,00 Reflector BeO 17.947 92.2 Inconel 0,136 0.7 Inert salt 1.246 6.4 Void 0.136 0,7 Totals 19.465 100,0 83 TABLE 9 REACTIVITY EFFECTS PROVIDED FOR BY THE CRITICAL MASS Effect Reactivity (k) Critical 1.0000 Fission product override 0.0050 Excess for experiment 0.0240 Excess for delayed neutron loss 0.0050 Excess for control 0.0075 Total 1.0415 TABLE 10 URANIUM REQUIREMENTS OF THE EXPERIMENTAL REACTOR Location Pounds Uranium in reacting volume 28 Additional uraniuvm in tube bends at ends of reactor 3 Uranium required in the reactor 31 Total uranium requirement in the system 140 A best guess of the ultimate uranium re- quirement allowing -2 1/2% in k¢ for critical experiment correlation 120 Reactivity Coefficients The values of the reactivity coefficients, summarized in Table 11 were obtained by bare reactor calculation methods, for a fuel slightly different from that now preferred. The values are, however, essentially correct for the new fuel. Power Distribution The power distribution in the ARE fuel-coolant has been evaluated by separating the total fission energy into three parts: (1) fission-fragment energy absorbed in the fuel-coolant (2) fission-neutron energy absorbed in the moderator and (3) gamma-ray energy from direct fission, fission prod- ucts, and (n, v ) absorptions. The large energy density is that of fission fragments in the fuel-coolant. The radial power distribution curve is shown in Figure 22. The experimental reactor core has an axial peak-to-average power- density ratio of 1. 6. Three megawatts in 17. 4 ft3core vclume represents an average power density of 6 kw/liter, 88% of which is in the fuel-coolant. The normalizing factor for the radial power distribution curve is 1. 4. 84 TABLE 11 REACTIVITY COEFFICIENTS OF THE EXPERIMENTAL REACTOR Change of Reactivity Range of With Change Of Variable Szmbol Value Thermal base (reactor /1283 to 1672°F %}i 5x 106 femperature) 68 to 1283°F Ak/k -0.011 Uranium mass at k=1.0 M 0.4 Am/ U Coolant density 90 to 100% of Ak/K -0.05 quoted density \AQ/R coolant Moderator density 95 to 100% of Ak/K 0.5 quoted density Ap/,o mod Density of structure 100 to 140% of /A g /K -0.17 (Inconel) quoted density AD/p Inc Hence the multiplying constant to determine radial power-density distri- bution (Figure 22) in the fuel coolant is thus 70 x 1.4 = 100 kw/liter. The power density in the moderator has an average value of 12% of the reactor power in 83% of the reactor volume. Neutron Flux and Leakage Spectra Neutron leakage from the surface of the reactor has been calculated. Leakage from the side reflector surface in neutrons/sec. cm? is shown in Figure 23 and leakage from the ends is presented in Figure 24. The rel- ative importance of the almost-open ends on neutron flux out of the reactor core may be noted. At full power, 3 megawatts, the total flux in the center of the reactor is 17 x 10*3 neutrons/sec. cm® The thermal flux is 3.0 x 10'® neutrons/ sec. cm?, and the fast fluxis 14 x 10*3 neutrons/sec. cm® The non-thermal flux averaged over the whole core is 7 x 10'® neutrons/sec.cm? and the likewise averaged thermal flux 1.5 x 10% neutrons/sec.ch. Neutron flux spectra at three points in the reactor are given in Figure 25. The important difference is that of amplitude. Note that high-energy neutron flux is relatively somewhat higher toward the reactor center, but the dif- ference is small. Figure 26 shows the corresponding plots of flux vs radius for four encrgies. The importance of moderation by the reflector is quite apparent. RADIUS/Ar .6 o O —_ NOTE. TO OBTAIN POWER DENSITY uél 5 \\ IN THE FUEL- COOLANT IN o WATTS 7cc AT FULL POWER (v o w 813 \ o w Ol.2 > o & T ) X o . 50 < ] € \ 1 Ar=2.206 cm 0.9 N A1 w N — z | wo.8 | o NORMALIZED TO ONE ' . |NEUTRON PER cc OF CORE. I m , a g CORE --{ REFLECTOR < 0 2 4 6 8 10 |2 14 16 |18 20 22 24 26 28 30 32 FIGURE22. RADIAL POWER DISTRIBUTION IN THE EXPERIMENTAL REACTOR . L& S » ol 58 LEAKAGE PER UNIT LETHARGY PER LEAKAGE PER UNIT LETHARGY PER FISSION .007 0 006 | .005 I""LL - A907 L THERMAL L 004 ESCAPE "-n_l__ AT | = = S.003 uz 18.6 o n & 002 001 4 0 8 16 14 12 10 8 6 4 2 LETHARGY (u) FIGURE 23. LEAKAGE SPECTRUM FROM REFLECTOR OF EXPERIMENTAL REACTOR 016 e 0l4 ] 012 010 | — 008 .00022 r_r—’ THERMAL .008 ESCAPE 7 AT n 004 U= 18.6 | J__ f’ 002 \\J /J_,J L 0 FIGURE 24. 18 LEAKAGE SPECTRUM FROM ENDS OF THE 16 14 12 10 8 6 LETHARGY (u) EXPERIMENTAL REACTOR 4 87 ANP—-PHY=-259 THERMAL VALUES: AT CORE CENTER AT —%— RADIUS AT CORE RADIUS AT CORE CENTER FISSION per cc per sec IN GCORE) AT —'2— CORE RADIUS AT CORE BOUNDARY NEUTRON FLUX, nper cm? per sec (FOR AVERAGE OF ONE O 2 4 6 8 10 12 14 16 18 20 LETHARGY (u) FIGURE 25. NEUTRON FLUX SPECTRUM AT VARIOUS | LOCATIONS IN THE EXPERIMENTAL REACTOR NEUTRON FLUX PER UNIT LETHARGY nper cm? per sec per u (FOR AVERAGE OF ONE FISSION per cc per sec IN CORE) 3| @) H O, H o Ol O W o N an n O ! Mev 194 500 ev NEUTRONS o ) O o N H FIGURE 26. THERMAL NEUTRONS REACTOR 8l NEUTRONS ‘CORE INTERFACE 0.4 ev NEUTRONS 6 8 0 12 4 16 I8 20 22 24 26 28 30 32 RADIUS/Ar RADIAL NEUTRON FLUX DISTRIBUTION IN THE EXPERIMENTAL REACTOR 34 0S2-AHd-dNV 89 Kinetics The above temperature coefficients of reactivity refer to slow changes in temperature. For sudden power surges, only the fuel temperature fol- lows the power, and onl rAi/f‘- and are effective. The P Y Am m): A p/pooiant Y give, together, a negative temperature coefficient of about 10-4/0C. There appears to be no conceivable mechanism by which a large change of keff could be introduced very much faster than it can be compensated for by the control rods. However, if the control rods lagged behind the reactivity change to the extent that Ak, remained uncompensated, a temperature in- crease of Ak.s /(107%/°C) would result. For Ak.; = 1%, this amounts to 100°C. Fast power oscillations would have a period given by 2w}/ \aSP 12 where g is the reciprocal of the mean neutron lifetime, about 10*4 sec™}, a the temperature coefficient of reactivity about 10 *oc S the reciprocal of the heat capacity of the total fuel, 6.4 x 10 ¢/ watt sec (assuming a heat capacity per unit volume of about 1 cal/ccOC) P the reactor power, 3 megawatt o This yields a period of 1-1/2 sec, long enough for a servo system to keep up with the oscillation. Furthermore, at 4 ft/sec it takes the fuel about 8 sec to pass through the eleven tubes, each 3 ft. long. Hence, about 3/4 of the delayed neutrons are given off inside the reactor and the loss of delayed neutrons due to fuel circulation does not appreciably decrease the damping which the delayed neutrons afford. Furthermore the circulation of the fuel itself creates a certain amount of damping, as yet not evaluated quantitatively. The damping effect due to the circulation of the fuel is illustrated by Figure 27 takenfrom Y-F10-109, '"Note on the Non-Linear Kinetics of Cir- culating-Fuel Reactor,' by S. Tamor. The curve illustrated refers to a somewhat simplified model, having the following properties: (1) All particles of the fuel spend the same time @ in the reactor. (2) The flux and power distributions are constant over the reactor; so that the temperature rise of a fuel particle during a time in- terval dt is proportional to the total reactor power P(t), the pro- portionality constant being independent of spatial coordinates; and so that a temperature change of an element of fuel influences the reactivity to an extent independent of the position of the element. 12. ORNL-63, p. 25, eq. 9. POWER — Po 24 2.2 2.0 4.0 3.8 3.6 P ———— TEMPERATURE (— 31%9F, dt 34 ———— POWER (Po) 3.2 3.0 2.8 2.6 @ _r =< & AVERAGE TEMPERATURE o J 2 3 4 5 6 T 8 9 10 LI 12 13 14 15 16 I7 18 19 20 21 22 23 TRANSIT TIME —& FIGURE 27 POWER AND TEMPERATURE OSCILLATIONS OF A CIRCULATING FUEL REACTOR e t—— . —taem—— 91 (3) A deviation T of the average fuel temperature from the equilibrium value T = Ocauses instantaneously a reactivity change proportional and opposite in sign to T. (4) The reactor inlet temperature is constant. The numberical values used were: temperature coefficient of reactivity a = 10™/°C, fast generation time U= 107*sec, reciprocal total heat ca- pacity of fuel 2.67 x 10°%C/watt sec, average power P, = 3 x 10® watts, fuel transit time through the reactor 1/8 sec. These values fall in the region applicable to ANP size reactors, but the general kinetic behavior is the same for an ARE. The initial conditions were: power equal to P, at all timesup to 1/10 of a transit time before t = O. Fora period of 1/10 transit time just before t = O, the power is forcibly held at 10 i,. At t>0O, the system is left to itself. The unit of the abscissa is the transit time 0. The ordinate on the right side is P in units of Py, and the ordinate on the left is the average temperature T in units of T/a8. It can be seen that al- ready the second overswing is much smaller than the first disturbance. This problem of damping fuel circulation is discussed in somewhat more general terms in Y-F10-98. Control Static calculations of the various mechanisms for the regulation, safety, and shim control of the circulating-fuel ARE are givenin terms of changes in reactivity. The one boron carbide regulator rod, fully inserted, effects a net change in reactivity of 0.0075. The three safety rods each effect a net change in reactivity of 0. 053. Details of the design of these control mechanisms are given under "Engineering Design' and their operation is further discussed under "Reactor Operation. " Shim Control Requirements Estimated reactivity changes, Ak/k.y, from room temperature to 1000°F and to an assumed controlled reactor temperature of 1283°F are given in Table 12. (Though this table was computed for a somewhat dif- ferent fuel, it is essentially correct for the presently preferred fuel.) The effect of xenon on reactivity at full power is to reduce the kqfof the clean reactor by Ak = 0. 0031. Maximum transient xenon reduces K further by Ak = 0.0082. The temperature coefficient as a result of the xenon at full power is(gi‘-é:k)( =+6.4x 10 'per OF inthe vicinity of 1400°0F. € TABLE 12 SHIM CONTROL REQUIREMENTS Z6 A k/k g for Change Effect Assumption Expansion of liquid fuel Volume expansion coefficients per OF: fuel, 1.67 x 10~% (melt- ing ignored). Inconel, 0.20 x 1074, Difference, 1.47 x 1074, Radial expansion determined by inconel, axial by BeO. Linear expansion coefficients per °F: Inconel, 6.7 x 1075, BeO, 4.9 x 1075, Weighted | average expansion coefficients linear, 6.1x 1078 volume. 18.3 x 1075 per ©°F. ) Dimensional expansion with constant material density Change of BeO density — Change of density of inconel in core Inconel volume expansion co- efficient per °F, 20 x 1076, Same as metal cooled ARE (See Y-F10-77), Change of density of reflector BeO Change of cross sections with reactor temperature Except Xe cross section. Change of Xe cross section with reactor temperature 0.26 of decay products in the core, remainder in external circuits. TOTAL Equation for (Ak/K.¢¢)/AT -1.47 x 10-4| Ak/Kk (flm/m)U -6 Ak/k 6.1x 10 OR/B) corec - -6 dk/k 18.3 x 107 ¢ 80,4 -6 Ak[k -20 x 10 (Ap/p)lnc Ak/k A k/k A k/k from 100 from 1000 of i°F at to 1000°F 10 1283° oper. temp. - -0.046 -0.015 -5.2x 107° ( p/p)coolant +0.002 +0.001 +0.3 x 1079 -0.008 -0.003 ~0.9x 1070 +0.003 +0.001 +0.3x 10°° -0.005 -0.002 0.6 x 10°° -0.027 -0.005 ~0,2 x 1072 +0.1x 10~° 0.081 -0.023 6.2 x 107° 93 Regulating Rod In the poison control system there is one axialboron carbide regulator rod lying along the longitudinal axis of the cylindrical reactor. The per- manent reactivity effect of the regulator structural material is included in the core volume fractions. Detailed calculations have been made for the sodium-cooled ARE reactor of this reactor control system design. The only important reactor characteristic that affects the control rod effectiveness is the neutron spectrum, and this spectrum is quite similar for the sodium- cooled reactor and the present ARE. Hence, the control rod effect on re- activity is assumed to be the same. The net change in reactivity when the regulator rod is fully inserted is 0. 0075. Safetx Rods Three safety rods are equally spaced on a 15-in. circle around the reactor axis. The spacing is sufficiently large to reduce the shadowing effect of each rod on the others to a relatively small value, so this effect is not considered. The rodsare 2-in. -dia cylinders of boron carbide with wall thickness of 0.335 in., and a beryllium oxide rod is attached to the end of each rod. The normal position of the safety rods is out so that beryllium oxide will be added to core material with a resulting decrease in the uranium requirement. The net change in k. g for each rod is thus the sum of two reactivity contributions: (1) from the removal of beryllium oxide moderating material, and (2) from the insertion of a 2-in. -dia boron carbide neutron-absorbing rod. Each boron carbiderod is worth -5. 3% in Ak/k, and each beryllium oxide rod in the same position is worth +0. 16%. The Inconel around the boron carbide rods is not an effective poison when the rods are inserted and results in a decrease in poison rod effect of 0.21%. The net effect of insertion of the three safety rods is thus 15. 75% In kegs . This value is larger than that quoted for the metal-cooled reactor becausethere isno poison NaK todisplace. The 15. 75% inkeff corresponds to about 6 1b of uranium in the core. 94 REACTOR MATERIALS High-Temperature Liquid Fuels High-temperature reactors in which the fuel is liquid promise certain advantages in power output per unit weight, control characteristics, and ease of fuel recovery over comparable solid fuel models. However, such designs impose stringent specifications on the liquid chosen as the fuel. Among the general and specific characteristics of the liquid which must be considered are: 1. Theliquid must contain sufficient uranium for criticality. . No elements of high cross section can be present. 3. The melting point must be safely below lowest design temperature. 4. The vapor pressure must be low at the highest design temperature. 5. Theviscosity must be low throughout the design temper- ature range. 6. The coefficient of thermal expansion must be large. 7. Theheat transfer properties (specific heat, thermal con- ductivity, and heat transfer coefficient) mustbe favorable. 8. The material must be radiation stable atdesign temper- ature and otherwise suitable structural metals must be able to serve as container materials. 9. The fuel must be amenable to simple chemical reproc- essing. 10. Additional advantages may be obtained if the liquid con- tains sufficient light elements to make it self-moderating. The recent adoptionof a design in whichthe liquid fuel is circulated to the heat exchanger effects some considerable engineering simplifications, particularly in removing the heat transfer system from the core. Such designs, however, impose additional demands on the fuel system in terms of efficient start-up of the reactor. It will obviously be advantageous to melt the fuel to a mobile fluid at the lowest possible temperature to mini- mize the required preheating of the reactor. In addition, the liquid system must be chosen so that a dilute (subcritical) fuel solution may be intro- duced into the core, and small increments of a concentrated solution of uranium inthe same solvent may be added to bring the system to criticality. This requires that solutions of widely differing uranium content melting at temperatures well below the operating range be available and that no high- melting-point compounds be formed at intermediate concentrations. 95 Liquid fuels possessing the capacity for self-moderation would permit the simplest reactor design. Although a mixture of sodium hydroxide and lithium hydroxide will dissolve sufficient uranium, hydroxide mixtures cannot be contained at reactor temperatures, and adequate quantities of the heavy lithium isotope are not available. For liquid fuels in which no moderator is present, solutions of uranium in liquid metals and inorganic salts are the two outstanding possibilities. Although it has not been pos- sible to prepare a solution of suitable uranium concentration inliquid met- als, a number of fluoride salt combinations, all including UFs, have been developed as potential reactor fuels. Of these, the system NaF-KF-ZrF;- UF4 appears to best fulfill the required characteristics of the fuel for the ARE, as itemized above. Homogeneous Fuels The logical, and indeed almost the only, choice of materials for self- moderating fuels are solutions of uranium compounds inthe alkali hydrox- ides. Preparation of such solutions in the proper concentration ranges is not 2 simple matter. Uranium trioxide seems to be soluble in a mixture of sodium and lithium hydroxides to an extent that would permit construc- tion of a homogeneous high-temgerature reactor if the low cross-section lithium isotope were available.!? Lack of this isotope and a suitable con- tainer for the resulting liquid does not permit construction of a feasible reactor at present. Nonmoderating Fuels From the standpoint of heat transfer properties and probably from that of minimum radiation damage, solutions of uranium metal inliquid metals would be a superior fuel. It does not, however, seem feasible to prepare a solutionwith suitable uranium concentration and especially solutions with relatively wide ranges of concentrations at the proper operating tempera- tures in such systems. Liquid fuels in which uranium compounds are dissolved in fused salts appear more promising. (Although such fuels seem practical for the ARE, considerably more experimental effort will be required to demonstrate their utility in a full-scale aircraft reactor) The restrictions on vapor 13. L. G. Overholser, D. E. Nicholson, J. D. Redman, 'Suspensions or Solutions of Uranium in Molten Hydroxides,'" Aircraft Nuclear Propul- sion Project Quarterly Progress Report for Period Ending June 10, 1951, ANP-65, p. 96. 96 pressure, thermal stability, and low neutron cross section, as well as preliminary experiments on radiation stability, suggest fluorides as the best chemical compound for this purpose. Of the uranium compounds, UFy seems to offer the most promise although the little-known UF3 might offer certain advantages. The high melting points of the pure compounds in- volved seem to be the greatest disadvantage. A number of inorganic fluorides, notably NaF, KF, BeFy, ZrFy4, LiF, PbFg, and BiFj, are useful, especially in combination, in lowering the melting point of UF,. As the data in Table 13 indicates melting points sufficiently low for reactor operation may be obtained by dissolving UFy in any of several different systems. Most of these systems, however, show low melting points only at high uranium concentrations. TABLE 13 MELTING-POINT DATA FOR POTENTIAL FUEL SYSTEMS U Concentration Range Melting below 550°C Lowest Melting Mixture System {mole % UFy) M. P, (£10°C) Composition (mole %) NaF-BeFg-UF4 0-15 330 50 - 47 -3 LiF -NaF-KF-UF4 0-6 450 97 (LiF-NaF-KF)* - 3 UFy NaF-RbF-UF4 24-41 500 45 - 23 - 32 NaF -KF-RbF ~-UF 4 18-41 500 56.3 - 3.7 - 15.0 - 25,0 NaF-KF-UF4 26-30 530 46.5 - 26.0 - 27.5 NaF-LiF-UF4 15-39 450 16.5 - 52.5 - 31.0 LiF-UFy4 ' 24-32 480 73.5 - 26.5 LiF-NaF-BeFg-UF4 0-15 235 49 - 30 - 20 -1 NaF-PbFg -UF4 0-22 430 57 - 28 - 15 NaF-KF-ZrF4-UFy 0-5 ' 425 5-51-42-2 * LiF - 42 mole % NaF - 11.5 mole % KF - 46.5 mole % 97 ~For the circulating-fuel reactor the desired uranium concentration is about 4 mole % (15 to 20 1b of U2¥ per ft3). Thus, many of the potential systems are not suitable with high U239 enrichment.l4 Use of more nearly normal uranium would result in increased critical masses and would prob- ably be feasible at reasonable penalty only for the NaF-RbF-UF4, Li'F- NaF-UF,, or Li'F-RbF-UF, systems, * Phase equilibria in the system NaF-BeFo-UF, (Figure 28) show this system to have a wide range of uranium concentration even at melting points as low as 350°C. However, because of its relatively high viscosity the system may not be used in the ARE although it could be advantageously employed in an aircraft reactor system. The most promising fuel for the ARE seems to be obtained by addition of UF4 to suitable compositions in the ternary system NaF-KF-ZrF4 shown in Figure 29. Mixtures obtained by addition of 5 mole % of UF4 to a melt containing 46 mole % ZrF;, 36 mole % NaF, and 18 mole % KF melt inthe range 465 to 4950C. It appears practical, therefore, to start the reactor by adding small, uranium-rich portions to a dilute, subcritical fuel. The viscosity of this fuel has been shown to be low enough to permit turbulent flow in the ARE core system. In addition, the ternary system NaF-RbF-ZrF, seems to be of con- siderable value as a fuel solvent. The cross section of rubidium is mark- edly lower than that for potassium, and the melting points available with RbF appear to be considerably lower than those for the KF prototypes. Although RbF is notnow commercially available in quantity ata reasonable price, it would definitely seem to be worthy of future consideration. If necessary, UF3 could probably be substituted for UFys with little penalty so far as melting point is concerned. It should be noted, however, that if the heavy isotope of lithium were available in pure form, several additional low-melting-point systems of possible value would warrant im- mediate consideration. Chemaical Properties Uranium tetrafluoride is the most reactive of the materials contem- plated for use in the reactor fuel. This well-known material reacts with oxygen at elevated temperatures to yield UO9F9 and UFg as the major re- action products. It reacts readily with water vapor at temperatures above 14. For all fuel solutions an isotope concentration of 93. 3% U“° has been assumed. 98 / ® 680°C @90 /)\ / o @ 710°C 1035°C UF, 9%0 / 900 e INDETERMINATE BeF, = FIGURE 28. 380°C NaBeF; 590°C Na,BeF, THE SYSTEM NaF-BeF,~UF, Gaaaw DWG. 14621 995°C NaF ZrF4 FIGURE 29, THE SYSTEM NaF-KF-ZrF, (TENTATIVE) 99 100 5000C to give HF and UO,. The alkali fluorides are capable of slight hy- drolysis at elevated temperatures into HF, and the alkali hydroxides and ZrF, are intermediate between the alkali fluorides and UFy in ease of hy- drolysis. None of the latter compounds react appreciably with oxygen. UF; would almost certainly hydrolyze readily and probably would be oxi- dized simultaneously; the major products probably would be UOy and UFy if the reaction took place in insufficient oxygen. It isessential that such reactions be avoided since the oxides formed would be insoluble in the melt. In addition fo the possibility that the ox- ides would contribute to the corrosion difficulties, they would certainly segregate and lead to control troubles during reactor operation. Preven- tion of oxidation of these materials should present little difficulty since they would be prepared in a closed system and thoroughly outgassed at low temperatures before the melting operation. The alkali fluorides and beryllium fluoride, however, are extremely hydroscopic materials and cannot be dried in reasonable periods of time by evacuation at low temper- atures. Elevation of the temperature after addition of UFy would result in hydrolysis of the material. It should be possible to dry the materials separately and thenperform all the operations incompletely dry atmospheres. It seems more reason- able, however, to mix the materials in a reasonably dry condition and accomplish the final drying and reconversion to fluoride of any oxide pro- duced by hydrofluorination at 500 to 600°0C. Under these temperature con- ditions UQg would be completely converted to UF4by treatment with dry HF. The acid fluorides of the alkali metals are completely decomposed at 4000C sothat complete removal of the excess HF would presentno prob- lem. The purified and oxygen-free fuel would be maintained under a com- pletely inert atmosphere in its original container and charged as a liguid directly into the reactor. Physical Properties Since a final choice of the fuel composition cannot be made until after the critical experiment, it is not possible to cite specific values for the physical properties. Several compositions in the system NaF-KF-ZrF,- UF, are under concerted study at present. It seems apparent that the vapor pressure of the NaF-KF-Zr¥F,-UF, system will be low (probably less than 5 mm) at 1500°F. The vapor pres- sure of ZrF, is estimated in the open literature at 10 atmospheres at 9009K and the vapor pressure of UF4 is known tobe about 6 mm at 12739K. Sodium and potassium fluorides are even less volatile. However, neither ZrF, nor UF, exists as such in these fuels. The zirconium is apparently 101 combined in complexes of the type NagZrFs and NaZrF;, whereas the ura- niui” fluoride seems to exist in similar compounds. These materlals, along with the alkali fluorides. are considerably less volatile. | The coefficient of thermal expansion has been studied for a number of fused salt mixtures containing uranium tetrafluoride. The value of3x10"%/ OC should be sufficient for control purposes. The density for a fuel con- taining 18 to 20 1b of uranium tetrafluoride per cubic foot is about 3.0 g/cc at1500°F. The density of this fuel NaF-KF-ZrF,-UF, (mole % 34.7-17. 4- 44.4-3.5) is shown in Figure 31. One of the more important physical properties, sofar as the ARE de- sign is concerned, is the viscosity of the fuel throughout the temperature range of operation. Examination of the literature has indicated that the viscosities of pure fused salts, including fluorides, should be 1 to 2 centi- poises at temperatures 300 to 4000C above their melting points. Prelimi- nary experiments with the system NaF-KF-UF, at high uranium concen- trations indicated that the viscosities were relatively high(2 1/2 to 4 cen- tipoises at 800°C). There was definite indicationthat evenhigher viscosity values would be obtained if hydrolysis occurredand if apprec1ab1e amounts of UO, were suspended in the measured liquid. | Experiments with various mixtures in the NaF-KF-ZrF,-UF, system containing-2. 0 mole 9% UF4 have shownthe viscosity to beless than 10 cen- tipoises at 6250C; these data are shown in Figure 30. Although further study will be necessary to establish accurate values for the actual fuel mixture (containing 3.5 mole % UF,) it is evident that the viscosity values will be within acceptable limits. | Chemical Processing An attractive feature of the liquid fuel reactors is the ease with which the fuel may be removed from the core, manipulated by remote-control devices, and chemically processed to decontaminate and recover the ura- nium for subsequent reuse. For the liquid fluoride fuels, at least two simple methods of chemical processing should be feasible. : The fluoride fuels may be dissolved in aqueous solutions containing oxidants and agents capable of strong complexing of fluoride ion. The lig- uid fuel could be drained to a remote, shielded container, cooled, dissolved inan aqueous solution of aluminum nitrate andnitric acid, and treated with additional aluminum nitrate to form a solutionfrom which the uranium may be recovered by extraction with organic solvents. The organic solvent layer could subsequently be selectively washed to remove various fission products and finally scrubbed to recover the uranium in aqueous solution. 102 12 — FUEL GCOMPOSITION 210 NaF — Smole% 0 \ NaF ~ Slmole % g 8 N ZrF4— 42mole % ';:‘ \ UFs4 - 2mole% S 6 L = s 4 o ? S 2 0 400 500 600 700 800 900 1000 TEMPERATURE (°C) FIGURE 30 VISCOSITY OF NaF-KF-ZrFs—-UF4 AS A FUNCTION OF TEMPERATURE wl_—_— 3.4 3.2 A o 3.0 \GOOQ FUEL GCOMPOSITION NaF-34.7mole % KFq-17.4 mole % ZrFy- 44.4 mole % 2 2| UF4-3.5mole % N o DENSITY (g/cc) n ® n S P=3.72-0.824x10"3 ¢t 500 600 700 800 900 1000 TEMPERATURE (°C) FIGURE 3| DENSITY OF NaF-KF-ZrFq—UFs AS A FUNCTION OF TEMPERATURE 2.0 103 This uranium would, presumably, be precipitated with ammonia or hydro- gen peroxide, calcined to UsGsor UQOs, reduced with hydrogen to UOg, and hydrofluorinated to UF,. If all these operations were to be carried out by means of remote-control devices, as should be possible, a particularly high degree of decontamination would not be required. Another possible recovery cycle of apparent attraction involves suc- cessive hydrofluorination and fluorination of the solidified fuel. In this method the fuel would be drained from the reactor into storage tanks be- hind proper shielding, allowed to solidify, and treated with gaseous HF at elevated temperatures to remove all fission products whose lower valence fluorides are volatile. At the conclusion of the hydrofluorination reaction, gaseous fluorine would be admitted at considerably lower temperatures to volatilize the contained uranium as UFg The sodium, potassium, and zir- conium fluorides contaminated with the fission products that failed toform volatile fluorides would be left. The UFg could be reduced directly to UF, by use of anorganic reducing agentat a moderate temperature inthe pres- ence of sufficient sodium, potassium, and zirconium fluorides to form a liquid suitable for reuse as fuel. This set of reactions could probably be conducted by remote control; it remains to be seen whether adequate de- contamination of the uranium would be provided. Structural Metals The structural metals for the final aircraft reactor must be carefully selected to satisfy a large number of very stringent requirements. Since heat transfer will be a major problem, the heat exchanger design must make use of large numbers of extremely thin-walled tubes. The metal chosen must therefore be easy to fabricate in reasonably intricate leak- tight assemblies and must be virtually completely resistant to the fluorides sofar as corrosion and mass transfer are concerned. In addition, high- temperature strength, creep strength, etc., of the metal must be satis- factory in the thin wall sections, and these properties must be unaffected by the fluorides and their corrosive action. Not all these problems are of great importance in operation of the ARE. The heat exchanger system to withdraw power at the modest design rate need not be elaborate; therefore thin-walled tubing in the heat ex- changer region is not necessary. Within the reactor the critical mass is not particularly sensitive tothickness of the metal tube walls; the ARE may be designed withlarge factors of safety from all points of view and, if nec- essary, considerable corrosion by the fuel may be tolerated. 104 Molybdenum, which would be superior from a high-temperature strength viewpoint, cannot yet be fabricated easily. Among the common high-temperature structural metals none appears clearly superior in strength, and several of them (Inconel, 400 and 300 series stainless steels) possess sufficient high-temperature strength and creep resistance to be suitable, especially if thick-walled tubing is used. Choice of the metal for the ARE can therefore be made largely on the basis of resistance to corro- sion by molten fluorides. Many structural metals, including nickel, Inconel, and the 300 and 400 series stainless steels, have been shown to be adequate containers for the fluoride mixtures under static, isothermal conditions. Nickeland Inconel, however, appear to be most resistant to these materials under dynamic conditions. In 500 hr of exposure, with the heated portion of a convection loop at 1500 and 250°F temperature gradient, some 3 to 7 mils of pitting has been observed on an Inconel test loop containing carefully prepared NaF-KF-ZrF,-UF, mixture. The ARE Inconel fuel tubes with 60 milwalls should certainly be adequate for the scheduled 1000 hr of reactor operation; meanwhile, it is believed that this corrosion can be materially reduced. Static Corrosion Corrosion of a variety of structural metals under static isothermal conditions was shown to be very slight when the melt consisted of UF4 in alkali and alkaline—earth fluorides with and without zirconium fluoride. Mixtures containing lead fluoride were unsatisfactory since structural metals reduced this compound to metallic lead. : The metals which proved to have adequate resistance to static corro- sion by mixtures of alkali, alkaline earth, zirconium, and uranium fluo- rides are Inconel, nickel, molybdenum, and the stainless steels of the 300 series. The corrosionobserved by metallographic examinationof the cap- sule wall generally consisted of intergranular penetration of the metal and attack at the grain boundaries to the extent of 2 mils or less. In these static tests Inconel seemed to be slightly more resistant to attack than did the 300 series stainless steels; the 400 series stainless steels appeared somewhat less satisfactory. On the basis of a few tests, molybdenum ap- peared somewhat better than any of the metals listed. It appears possible that molybdenum, when its fabrication is better understood, will be of value in later and higher temperature models. The data in Table 14 indicate that at 1500CF, and for intervals of 100 hr, it is possible to contain the fluorides in Inconel or type 316 stainless steel with little corrosive action. TABLE 14 105 STATIC CORROSION OF STRUCTURAL METALS BY POTENTIAL FUELS Fluoride Fuel NaF-KF-LiF -UF4* NaF-KF-LiF -UFy4 NaF-KF-LiF-UFy, NaF-KF-LiF -UF4 NaF-KF-LiF-UF4 NaF-KF-LiF-UF4 NaF-KF-LiF-UFy4 NaF-KF-LiF-UF4 NaF-KF-LiF-UF4 NaF-KF-LiF-UF4 NaF-KF-LiF-UF, NaF-KF-LiF-UF4 NaF-KF-LiF-UFy4 NaF-KF-LiF -UF4 NaF-KF-LiF-UF4 NaF-KF-LiF-UF, NaF-KF-ZrF4-UF4™ NaF-KF-ZrF4-UF4 NaF -KF-ZrF4-UFg NaF-KF-ZrF4-UF4 NaF-KF-ZrF4-UFy4 NaF-KF-ZrF4-UF4 NaF-KF-Z1F4-UFy4 NaF-KF-ZrF4-UFy4 (100 hr Tests at 1500°F) Containe: Type-316 stainless steel Type-316 stainless steel Type-~316 stainless steel Type-316 stainless steel Type-316 stainless steel Type-316 stainless steel Type-316 stainless steel Inconel Inconel Inconel Inconel Inconel Inconel Inconel Inconel Inconel Inconel Inconel Type-316 stainless steel Type-316 stainless steel Inconel Type-316 stainless steel Inconel Type-316 stainless steel * Composition in mole %: 10,9-43.5-44.5-1.1. # Composition in mole %: 5.0-51.0-42.0-2.0. Weight Change (mg/dm?/day) -21 -33 +10 +14 +14 + 1 - 6 -2 -16 -9 Penetration No attack No attack Slight pitting 1 mil 0.5 mil 0.5 mil 1 mil No attack No attack No attack 1 1o 2 mils 1 to 2 mils 1 mil 1 mil No attack 1 mil Light pitting Very light pitting No apparent attack No apparent attack Very light pitting Very light pitting Very light pitting Very light pitting 106 In similar static tests other specialalloys such as Timken alloys Nos. 3 and 6 and Stellite, as well as Hastelloy B, Nimonic, and some molyb- denum-nickel alloys, have shown good corrosion resistance. The former alloys will probably be of value in bearings and seals, and the latter may be shown to be superior structural materials. Dynamic Corrosion In tests in which the fluorides are circulated through closed loops by forced thermal convection, however, a definite and more severe corrosion pattern has been revealed. The loops for these tests have been prepared by heliarc welding of 1/2 in. —IPS schedule 40 —pipe of the desired ma- terial. The hot leg of the loop is maintained at 15000F by electrical re- sistance heaters. Temperatures inthe cold leg of the loops are maintained at 150 to 1800F below those of the hot leg. The loops are sealed with an atmosphere of dry, oxygen-free helium above the liquid. Under these con- ditions of operation the circulation rate is 6 to 8 ft/min. In initial tests using mixtures of UF; with alkali fluorides only loops of Inconel and nickel have operated for 500 hr at temperature without dif- ficulty. Loops of series 400 stainless steel were plugged by the deposition of high~-melting-point material in the cold leg after less than 30 hours. Loops of 300 series stainless steel showed similar but less rapid plugging, and circulation was stopped by this phenomenon after 100 to 150 hours. Examination of the Inconel loops after scheduled termination of the tests indicated some mass transfer of metal from the hot to the cold portion of loops, but the quantity was insufficient to cause failure of the loop. Solutions of UF, in NaF-BeF, have been shown to be somewhat less corrosive than similar solutions of this material in NaF-KF-LiF mixtures. In general, however, the NaF-KF-LiF mixtures without UF4 have been less corrosive than either. Addition of small quantities of NaK to the NaF-KF-LiF mixtures (with- out UF,) considerably improved the corrosionresistance of structural met- als. Tests of 500 hr duration in loops of stainless steel have indicated virtually negligible corrosion; photomicrographs of hot-leg sections of loops exposed to NaF-KF-LiF mixture with and without added NaK are shown in Figure 32. Recent tests in which solutions of UF, in NaF-KF-ZrF, mixtureshave been exposed for 500 hr in Inconel yielded considerably more favorable results. The hot legs of the loops showed scattered pitting to a depth of 3 to 4 mils with infrequent intergranular attack to a depth of 7 mils (see Figure 33). A thin (perhaps 1 mil) nonmetallic layer was observed in the Foar i\, & UNCLASSIFIED PHOTO NO T-934 : x.’; M‘ “: _7 ,)/’; ; o 5 . . i i Lo A R CORROSION OF HOT LEG OF INCONEL LOOP AFTER {000 HOURS B NaF-KF-LiF(11.5-42.0-465 MOLE %) 250X UNCLASSIFIED PHOTO NO.T-966 'cofinosmN OF ”HOT LEG OF INCONEL LOOP AFTER 500 HOURS BY NaoF—KF—LIiF (11.5-42.0-46.5 MOLE %) WITH 35 VOL.% NaK AEE(I)T)I(VE FIGURE 32. EFFECT ON NoK ADDITIVE ON THE CORROSION OF INCONEL THERMAL CONNECTION LOOPS BY NaoF — LiF-KF AT 1I500° F. 107 108 PHOTO NO, T—I1335 ..... CoLD~1275 °F— LEG OF LOOP 250 X PHOTO NO. T-1334 HOT - IS00°F -LEG OF LOOP 250 X FIGURE 33, CORROSION OF INCONEL THERMAL CONVECTION LOOP BY NoF-KF-ZrF, —UFs (48—-501-41.3-38 MOLE%) AFTER 500 HOURS. 109 cold legs of the loops; virtually no metallic deposits, such as are produced by mass transfer, were visible. It is believed that the improved results can be attributed to the recently initiated techniques for purification of the liquid fuels before test. It seems certain that the optimum conditions for such purificationhave not been established and that continued improvement in corrosionresistance of Inconel to the melts istobe expected. Chemical analysis of the fuel before and after operation in the loop is given in Table 15. TABLE 15 CHEMICAL ANALYSIS OF THE NaF-KF=-Z1F4-UF4 FUEL FOR CORROSION PRODUCTS (circulated in an Inconel loop for 500 hr) Temperature Fe Ni Cr Origin of Fuel Material (°F) U(%) F(%) (ppm) (ppm) (ppm) Original batch analysis 75 5.90 39.4 180 4170 50 Hot vertical leg of loop 1550 3.88 40.1 445 30 480 Hot horizontal leg of loop® 1250 to 1500 5.69 39.6 240 20 480 Cold vertical leg of loop* 1250 5.68 39.5 250 40 460 Cold horizontal leg of loop 1370 5.61 39.8 190 30 510 * The results presented are the average of two samples from these sections of the loop. The very considerable improvement in corrosion by uranium-free alkali fluorides that resulted frora addition of NaK would seem to offer another possible technique for controlling the corrosion problem. To the extent that this improvement was the result of the cleansing action of NaK on oxidized metal surfaces, and, therefore, to the extent that prior treat- ment with NaK would effect the benefit, this improvement can be readily realized. If the mechanism requires the continued presence of elemental alkali metal there is the possibility of reduction of UFy; and ZrF,4 to com- pounds of lower valence states and to the metals. Whether the free alkali metal can be tolerated (and to what extent) in the fuel and what benefit can be obtained by cleaning with NaK and its removal prior to fuel addition is under study at present. The increased corrosion resulting from inclusion of uranium tetra- fluoride in the fused salt mixtures, if it is real, may be caused by the oxidizing power of UF4 or more probably some material {(perhaps UOgF:) added as an impurity with the uranium. If the oxidizing action of UFy is the cause, then UF3 should be a superior material to use. In the other event, increasing the purity of the fuel mixture should improve the situa- tion. 110 In any case, it appears that for the ARE the corrosion that must at present be considered normal in Inconel for periods up to 1000 hr would not be prohibitive. Present designs for the reactor call for 60-mil tube walls in the hot zone. Corrosion at the rate observedat present would not seriously weaken this structure; it is apparent, therefore that the ARE is feasible even with the corrosion rates presently observed. Control Rod Fabrication The control rod system for the Experimental Reactor consists of three safety rods and one control (regulating) rod. The material for these rods will be produced by powder metallurgy techniques employing hot pressing. The safety rods are constructedof a canned Fe-B,C mixture. Theannular cans are fabricated of type-316 stainless steel and have an inner diameter of 1 1/4 in. and an outer diameter of 2 inches. The annulus between the two walls is filled with the Fe-B,C mixture. The cans are 1.95 in. long, and about 18 cans will be required per control rod assembly. The B4C-Fe slugs contain approximately 20% by volume iron, which acts as a binder. The B4C slugs are 1.86 O.D. by 1.26 I.D. by 1.895 in. long. At this time 20 of the B,C slugs have been successfully hot pressed. The slugs were fabricated by hot pressing at 15200C under a pressure of 2500 psi using graphite dies. The first few slugs made with this process cracked longitudinally owing to the difference of thermal coefficient of ex- pansion of the B,C-Fe mixture and the graphite mandrel. This difficulty has been overcome by pressing the mandrel out of the slug while at the hot-pressing temperaiure. The regulating rod will be made by using slugs of Al,O, impregnated with B4C contained in one of the 300 series stainless steels. Rods of two strengths, in which the only difference will be the amount of B4C added to AloO3, will be made for evaluation in the hot critical experiment. In one case 0.022 to 0.024 g/cc of B,C will be added to AlyO5 and the other rod will contain 0. 005 to 0.006 g/cc of B4C with Al,0;. The Al,O; slugs will have the same dimensions as the B4C slugs except that they will be longer. The dilute slugs will be 2.14 in. long and the strong slugs will be 2 in. long. The slugs for the regulating rod being made are hotpressed at 16750C in graphite dies under pressure of 2500 psi. Metallographic examination of a slug of Al,0; and B,C has shown that the two materials are quite compatible at the operating temperature of the control rods and that there is no undue reaction between this mixture and the 300 series stainless steel container materials. Some of the B4C-Fe slugs have been canned in 300 series stainless steeland brazed through the Nicrobrazing cycle, which consists of heating at 1130°C for 10 min; they 111 were then held at 815°C for 100 hr with slight reaction. Therefore it can be concluded that there will be no difficulty in the fabrication and operation of the regulating and safety rods for the ARE. Stress-Corrosion Tests Any time a structural metal is used in a stressed condition and also subjected toa corrosive media there is the possibility of a stress-corrosion phenomenon. Physical tests are being conducted to obtain data that will indicate whether the structural metal of this reactor, e.g., Inconel, is subject to this phenomenon when in contact with the fluorides. Two types of physical tests have been conducted: (1) tube burst tests to furnish hoop stress data for metals in contact with the fluoride medium and (2) stress-rupture tests in which the specimen is in a fluoride environment. In the stress-rupture tests, dead weight loading is transmitted to a small sample by means of abellows. Elongation is measured by the move- ment of the end of the load lever arm. The sample is held in a bath of fluoride maintained at 8160C. In a tube burst test one section of the tube wall is reduced to a known wall thickness and serves as the gage length. The hoop stress is produced by placing an inert gas atmosphere inside the tube while the outside of the tube is bathed in the fluorides at 8169C. This test only yields time-to-rupture and total elongation data. Table 16 is a summary of the stress-rupture data being obtained on Inconel specimens in the fluoride salt, NaF-KF-LiF-UF; (mole % 10.9- 43.5-44.5-1.1), at 816°C. The Inconel is used in the as-received condition. These data are compared with that obtained from tests using fine-grain as-received Inconel in an argonatmosphere. As canbe seen, the fluorides did not impair the load carrying abilities of the materials appreciably. TABLE 16 INCONEL STRESS -RUPTURE DATA Stress Time to Rupture (hr) {psi) NaF -KF-LiF -UF4 Argon 12,500 6.5 6 10,000 20 - 19 7,500 25.7* 80 5,000 575" 270 4,000 e61** 430 * Average of three tests. Test apparatus failed before rupture. ** Specimen ruptured in weld area of grip rather than in test section. 8 112 In addition to the testslisted in the table, one Inconel sample has been under a stressof 4000 psi for 350 hr and another has withstood 5000 psi for 525 hr without undue elongation. Furthermore, Inconel tube burst tests have successfully operated from 500 to 1000 hr at stresses of 2000 to 3000 psi without severe corrosion being induced by the stress. Radiation Damage Inconel capsules containing several different fluoride mixtures have been irradiated in both the LITR and the X-10 graphite pile. Radiation damage studies with the NaF-KF-ZrF,-UF; fuel are not yet available be- cause of the time required for the analysis of hot specimens. Neverthe- less, the irradiation of other fluoride mixtures atpower levels inthe range that will exist in the ARE (approximately 100 watts/cc) have shown no ap- parent increase in decompositionor corrosion owing to the radiationfield. However, in the case of one fuel, irradiation at power levels approaching that in the aircraft reactor have yielded results that may be interpreted as considerably increased corrosion under strong irradiation, but no increase was evident in similar tests with a second type of fuel. Significant data are now available on irradiations of both the NaF-KF - UF, and NaF-BeF,-UF; mixtures at various power levels. These data are summarized in Table 17 for several irradiations. Metallographic exami- nation of the metal specimen and analysis of the fluoride melts for metal corrosion products are the methods used for detecting radiation damage. The data from the irradiated specimen are compared with similar data from a contirol specimen not irradiated but otherwise subjecied to asnearly as possible identical treatment. Theirradiation intensity is givenas watts dissipated per cubic centimeter of specimen, since this is more meaning- ful than neutron flux when irradiating fissionable material. As indicated in Table 17, exposures of samples of the NaF-KF-UE, fuel in Inconel at the maximum flux of the X-10 graphite pile caused a power dissipation of approximately 65 watts/cc of material and have been quite satisfactory. The irradiated samples showedno detectable gas evolu- tion because of the irradiation, nor-—after exposures up to 400 hr—was there any positive evidence of damage. The corrosion of the capsule, as measured by metallographic techniques, was the same order of magnitude of that in similar tests conducted outside the radiationfield (Figure 34), as was the chemical analysis for structural elements in the solidified liquid. Exposure of similar samples in the LITR at 800°C resulted in power dissipations of 800 watts/cc of material but yielded results that may be interpreted as considerably increased corrosion under the strong irra- diation. Although it appears probable that at sufficiently high radiation Fuel Composition (mole %) PILE IRRADIATION TESTS ON FUSED FLUORIDE FUELS IN INCONEL Temperature {oF) NaF-KF-UF,, 46.5-26~27.5 NaF-KF-UFy, 4b.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UFy, 46.5-26-27, 5% NaF-KF-UF,, 46.5-26-27.5% NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-KF-UF,, 46.5-26-27.5 NaF-BeFz-UFy, NaF-BeF,-UF,, NaF-BeFz-UFy, NaF-BeFy-UF,, NaF-BeF,-UF,, NaF-BeF2-UF,, NaF-BeF;-UF,, NEF—BEFQ‘UFp NaF-BeF;-UF,, NaF-BeF,-UF,, NaF-BeF;-UF,. 25-60-15 25-60-15% 25-60-15% 25-60-15 25-60-15 47-51-2% 47-51-2% 47-51-2% 47-51-2% 47-51-2 47-51-2 1500 1500 1500 1500 1500 1500 1500 1500 75 1500 Time {or) 458 458 458 472 299 299 299 344 1500 1500 824 (solid) 1500 1500 1500 1500 1500 1500 1500 1500 1500 1500 75 * Pressure tested, no evidence of buildup. 164 161 136 126 131 139 137 137 26 136 143 115 145 TABLE 17 Inconel Components in Fuel Irradiation After Test (ppm} 113 ‘watts{cc) Ni Cr Fe Capsule Condition {Photomicrograph) 65 194 194 1,030 65 262 20 1,630 65 20 188 1,130 0 (bench) 160 125 2,630 No apparent attack. 65 93 23z 1,530 65 148 532 2,340 Intergranuiar attack, maximum pene- tration 1 mil. 65 183 146 622 Q (bench) 93 186 548 .- 123 207 31 Original salt analyses. 800 26,800 1,870 16, 300 Intergranular, plus "diffusion” type of attack to a depth of 2 to 3 mil. Discarded because of faulty weld. g (bench) ) D30 754 1,270 800 45,100 950 6,410 Intergranular attack, not dense, to a depth of 0.5 to I mil. 0 (bench)y 1,100 645 2,650 No apparent attack. 800 1,380 160 1,220 Very small amount of intergranular at- tack, approximately 0.5 mil with gen- erally clean interface. 554 0 (bench) 1,540 7,380 2,300 Subsurface voids or pitting effects 1 to 1 1/2 mil deep. 554 1,490 1,060 1,200 Generally free from attack. 554 Metallurgy in process. 554 450 1,810 T80 Metallurgy in process. 84 5,610 3,900 1,430 0 (bench) 1,120 3,300 811 84 58,500 2,450 3,100 Some coarse intergranular attack ap- proximately L mildeep. Thecorrosion interface is generally free from attack. 84 1,880 1,880 2,690 0 (bench) &0 4, 450 445 Subsurface voids or pitting effecis to 1/2 mil deep (effect less dense than in sample seven lines above). —_ 50 330 540 This was the original salt analysis be- fore loading in capsule. 114 Y-35460 CORROSION OF UNIRRADIATED CAPSULE AFTER 115 HOURS. 300 X Y-5939 CORROSION OF IRRADIATED CAPSULE AFTER 299 HOURS IN A NEUTRON FLUX WITH AN EQUIVALENT POWER DENSITY OF 65 WATTS PER GG OF FUEL. 300X FIGURE34. EFFECT OF NEUTRON [RRADIATION ON THE STATIC CORROSION OF INCONEL BY NaF-KF- UF, (46.5—26.0-27.5 MOLE%) AT 1500°F 115 intensity the pure fused salt fuels under static conditions may suffer con- siderable radiation damage, the above experiment is not conclusive evi- dence of this since temperature gradients in the fuel or irradiation of im- purities in the fuel could be the mechanism for the observed increase in damage. Refined preparation and handling techniques will reduce the im- purities. It might be possible to reduce the potentially available fluorine by substitution of UF3 for the UF4 or by dissolving a small amount of an active metal in the fluoride melt. In another series of fuel-capsule irradiations the NaF-BeF,-UF, sys- tem in an Inconel capsule has been exposed inthe LITR to power dissipations of both 84 and 554 watts/cc of fuel. Inonly one of five exposures(Table 17) was there any indication of damage beyond that experienced in the control specimen. The nickel content of the salt washigh, but upon metallographic examinationfew signs of radiation inducedcorrosion were present. These particular results are most encouraging since they represent power levels considerably in excess of that to be realized in the ARE. Unless studies in progress on the NaF-KF-ZrF,;-UF; system show it tobe more suspectible than either the NaF-KF-UF4 or NaF-BeF,-UF; sys- tems, it is not apparent that radiation damage will introduce any consider- able uncertainty at the fluxes tobe maintained in the ARE. On the basis of the above static tests, the ARE at this point appears feasible from the standpoint of radiation damage. 116 CRITICAL EXPERIMENTS A preliminary experimental study of the nuclear design of the experi- mental reactor is to be made during the summer of 1952 at ORNL by build- ing some of the major components into a system that will be made criti¢al with enriched uranium. The experiments will be done at a power level of the order of 1 watt, maximum, at roomtemperature and will yield informa- tion on uranium requirements, control problems, and power and neutron- flux distributions. It is expected that assembly of the mockup will start about June 1 and that the system will be made critical about July 15. Experimental Materials The beryllium oxide hexagonal blocks prepared for the ARE will be used as the reflector and moderator inthese experiments andwill be trans- ferred to the ARE later. The fuel is to be a dry-packed powder having nu- clear properties closely approximating those of the ARE fuel. At present the latter is designed to be a molten mixture of the fluorides ZrF4, NaF, KF, and UF,. In the preliminary assembly the fuel is expected to be a mixture of ZrQOy;, NaF, KF, enriched UF, and graphite. The ZrO; is being substituted for the ZrF4 because the latter is unavailable at this time; the graphite is being added to give total moderation equivalent to that of the fluorine. The composition will be that required to give the design uranium density of the ARE fuel. It is probable that the relative quantities of the constituents willapproach ARE design, with the over-all density somewhat lower. The simulated reflector coolant has the same composition as the simulated fuel except for the omission of the UFy. The materials will be packedas a dry powder in stainless steel tubes, 1 1/4 in. in dia and 40 in. long, that can be sealed to minimize moisture pick-up. The stack of beryllium oxide blocks will be surrounded by an Inconel shell. Experimental Assembly The beryllium oxide blocks will be stacked, with axes vertical, in a right cylinder, 36 in. high and 47 in. in dia;, alsc with its axis vertical. The fueland coolant mixtures in the stainless steel tubes, will be inserted vertically through the axial holes in the beryllium oxide blocks. Some of 117 the tubes will be movable by remote operation and will serve as control rods. A section of tubing containing a nuclear poisonwill be attached above other fuel elements, and the fuel elements will be magnetically supported. Upon release by the magnet, the fuel tubes will fall out of the core and will be replaced by the poison. This safety mechanism will be operated both manually and by a high-flux-level detection instrument. Provisions will be made for inserting neutron-detecting foils, such as indium and gold, through the reflector and moderator. The usual neutron- sensitive and gamma-sensitive detectors will be placed adjacent to the assembly for operational control and safety and will feed recording and indicating instruments in the control room. Experimental Program The first result, obtained from the initial loading, will be the mass of uranium required for the assembly to become critical, thereby fixing the uranium density. After calibration of the control rods, reactivity changes incurred by the introduction into the reactor of other materials canbe eval- uated. For example, at least one fuel element designed for the ARE itself will be used. In addition, the nuclear properties of different types of con- trol devices can be initially tested. It is intended that extensive studies will be made of the distribution of neutrons throughout the assembly by the usual activation method. 118 REACTOR OPERATION The operation of the reactor may be broken down into three separate phases, each of varying importance. The start-up phase involves pretest- ing of the fluid circuit, loading to criticality, and regulating rod calibra- tion. The subsequent power operation of the reactor following the addition of enriched fuel material is the most important phase asfar asthe aircraft reactor is concerned. It is there that the kinetic response of the aircraft reactor may be determined and all other nuclear experiments are per- formed. Following the power cperation is the reactor shutdown phase. The probable elapsed time for each of the steps and sub-steps in the operational procedure is difficult to predict. Of the 500-odd total hours available for operation of the system after the original loading with en- riched fuel, probably not more than 100 hours will be available for power operation. A careful budgeting of time and rationing of operating hours is being made and will be revised as zero hour for operation approaches. The actual number of hours of power operation and the magnitude of the power cutput from the system cannot be known until the experiment has become history. However, all pertinent experimentation must be per- formed during this time. During the power regime it is planned to run a shielding experiment completely independent of the primary ARE objective. This will involve the determination of energies of the delayed fission neutrons. Arrange- ments are being made to make these measurements in the fluid circuit external to the reactor pit. The circulating fuel system provides a unique facility for making such a determination. Determination of one other basic phenomenon, the evaluation of xenon or its precursor, will be attempted. This will markedly affect design and operational considerations for an aircraft propulsive unit particularly as it relates to behavior after shutdown when xenon poison becomes of major importance. The schedule outlined below represents an ideal and almost certainly it will have discontinuities inthe form of start-ups following scrams, shut- downs for plumbing and mechanical repairs, etc. Every effort will be made to minimize these backings and fillings but it is unrealistic to assume that none will be necessary. 119 Pretesting After the reactor plumbing hasbeen installed and the system is ready for checking, both the primary (fuel) and secondary (reflector-pressure shell) circuits will be given a pneumatic test. These tests will be made be- fore the annular chaser pipeshave been joined over the welds in the liquid lines and beforethe heaters and insulationhave been applied to the annular chaser lines. Air or other gas will be held at a pressure of approximately 75 psi at room temperature and all welds and joints of any kind will be gone over with a bubble solution for the purpose of locating leaks. This technique should reveal all gross leak faults in both liquid circuits. After both circuits are tested pneumatically, helium at 75 psi will be substituted for the gas in the primary circuit and a vacuum will be pulled on the secondary circuit. The secondary circuit will be monitored for helium to determine whether or not there is a leak in the primary circuit within the core. After both circuits have been made tight as far as the above tests are able to reveal, they will then be filled with NaK, which will be circulated at some point near design flow rate. In addition to giving a further leak test, this will allow checking of pumps, tanks, and level controls under ambient temperature conditions. After the system has been checked out with room temperature NakK, the heaters will be installed, annular joints closed and insulation installed around all the system. Since this job is quite extensive, it will require many mandays of effort. When the heating system is installed, both liquid circuits will be operated at increasingly higher temperature levels until the system operates satisfactorily at an isothermal temperature of some 1200°F with NaK. In addition to checking the system hydraulically at high temperature, this circulation of hot NaK cleans the system of oxide. The Na,O formed in this cleaning is trapped out of the system as itisformed. Following the hot NaK test and after any leaks have been repaired the NaK will be drained from bothliquid circuits. While maintaining tempera- tures at 1200°F both circuits will be evacuated by pumping with vapor traps ahead of the pumps. When all residual NaK has been boiled out of the sys- tem, it will reach a steady state pressure, from which point rates of rise of pressure can be used to detect the fluid tightness of the system. It is hoped that this final test will leave the circuits tight and that no further leaks will occur. After the fluorides are introduced, repair of the circuit will be more difficult, even in some exireme instance impossible. ' 120 Loading and Start-Up In initial loading the fuel carrier fluoride melt is released from the fill tanks into the evacuated fuel system. Checking for voids in the core and piping will be done by observing any liquid level change in the surge tank when the gas pressure over the liguid is raised. After the system has circulateduntil it is proved tobe tight and all temperatures are stable, the system is ready for critical loading. (The secondary circuit filling is an identical procedure. ) Critical loading is effected by a fuel injection system which is filled with highly concentrated UF; in the NaF, KF, ZrF; carrier. The system is so arranged that samples of known weight can be introduced into the surge tank at a controlled rate. The first load will be injected at such a rate that the 'quantum'' of fuel, which is about six pounds, will require three circuit transit times for complete injection. Since the fuel transit time is about forty seconds, the 'quantum' injection will require about two minutes. This will insure a uniform distribution of fuel in the circulating system. As the calculated critical loading is approached, the amount of fuel in each "quantum”can be reduced from the maximum of six pounds to any reasonable low value so as to minimize the chance of loading to a super critical degree. The standard checksfor criticality inloading any reactor will be employed in the ARE and thereactor will be brought tocritical with the use of all possible safety precautions. The objective will be to go crit- ical at a temperature of 1200°F with the shim rods as nearly completely withdrawn as the controlled loading will permit. As a safety precaution, attempts will be made to measure the temper- ature coefficient of reactivity and to obtain a rough calibration of the regu- larity rod as soon as possible. The first attempt to make these measure- ments will be done whenthe reactor has reached a value of k = ~.98 or at a source multiplication of about 50. The first experiment will be an attempt to determine the sign of the temperature coefficient of reactivity. To do this the fuel temperature will be lowered from the equilibrium value of 12009F to approximately 11750F when the source multiplication is approx- imately 50. If the multiplication increases with this lowering of the tem- perature, it will show that the temperature coefficient is negative, and fur- ther loading can proceed. If, however, the multiplication should decrease, it will show that the temperature coefficient is positive, and the experiment should terminate at this point since a reactor which has a positive coeffi- cient is considered to be too dangerous to operate. However, this latter eventuality is not anticipated. 121 Following this experiment, an attempt will be made to obtain a rough calibration of the regulating rod. Since the moderator is beryllium oxide there will be anindeterminate number of gamma-n neutrons, and when the fuel is circulating, some of the delayed neutrons are lost in the reacting system. These two effects make the calibration of the regulating rod by pile period measurement practically meaningless. However, a roughcali- bration canbe obtained by noting the change in source multiplicationas the regulating rod positionis changed. It is believed that afirst measurement canbe made whenthe reactivity is approximately 0.98 with higher accuracy results obtainable as k is increased toward 1.00. This will give a fair calibration of the rod interms of Ak/k before actual criticality is reached. After criticality is reached the rod will be calibrated by noting the change in rod position for criticality with small additions of fuel. After the rod calibration has been made to the best degree warranted the reactor will be raised in temperature by the external heating facilities to atemperature of some 13000F since this is the mean temperature level for all future opetation. Atthisnew temperature fuel additiomwill be made to again bring the reactor critical—but with the shim rods now inserted to a distance of some four inches in the core. This will provide of the order of 2% in k available' in shim rods to take care of poison as the reactor operates sothat nofurther fuel addition should be necessary after the criti- cal loading at 1300°F has been accomplished. The next step inreactor operationcomprises raising the reactor power from near zero ito some nominal value. This will be done by withdrawing the regulating rod sufficiently to start the reactor on a positive period of some twenty to thirty seconds. When this period has increased to near an infinite period, as it will by virtue of the temperature coefficient, the reactor will be at a power level of some ten kilowatts. This pointcan also be checked roughly by the reading of the Log N meter. The reactor is now ready for the power regime. Power Operation Operation at power will be attained as quickly as possible after the reactor has been brought up to the nominal value of some ten kilowatts. Up to this point there has been no power extraction from the system, but when the reactor has stabilized at a nominal (ten) kilowatt level, the he- lium blowers will be started. Heat extraction will be at a controlled rate, so that the reactor inlet temperature will not be brought too low. Every precaution will be used to insure no danger of freezing the fuel-coolant. 122 Since the object of the experiment is to obtain as much operation time as possible at full power, an effort will be made to reach rated power as quickly as possible. The mean fuel temperature of 1300°F will be main- tained quite closely, but the fuel inlet temperature will drop and the reactor outlet temperature will rise until at full power there will be a temperature rise across the reactor of some 3500F., Should there be a false scram when operating at power, the helium will automatically be cut off to insure against freezing the fuel. During the power regime the system will be left very nearly alone. A minimum of changes in power demand will be made and as nearly as possible the re- actor will be operated in a steady state condition. Undoubtedly, interrup- tions of one kind or another will occur, but there will be no intentional "juggling' of any part of the system beyond the minimum required to estab- lish a pattern of the transient response. The servo-regulating system will not be employed unless it is found to be necessary because of an insuffi- ciently large negative temperature coefficient. Operation of the Servo-Control During allof the runoutlined so far the reactor has been under manual control. The servo system has been operating but it has been disengaged from the regulating rodand hence has not been controlling the reactor. At some time during the power operation regime, the servo will be connected to the regulating rod so that the reactor will be on automatic control. Under these conditions, since the servo is made to operate exactly as the fuel temperature coefficient, the system response to power demands will simulate the kinetic behavior of a reactor operating at high power output. A suitable servo error signal is the following: E =g, +a© +236, +a, where a,, a5, a3 a,are proportionality constants, flux reactor inlet temperaiure reactor outlet temperature = S, # In the steady state this error signal is independent of the external load. For load changes both ©; and 6, change, and the regulating rod will main- tain its original position. Furthermore, the power demand is made through ©; from the external load. To raise the mean fuel temperature at constant load, a, is the driving function and can be changed manually and slowly. Increasing a, causes 123 withdrawal of the regulating rod, an increase in ® followed by an increase in @i, and later a decrease in @ until the error signal is again zero, after the entire system, reactor and heat exchanger, have reached equilibrium. Once equilibrium has been reached the regulating rod will no longer be at its mid-positionbecause the reactor has a temperature coefficient independ- ent of the servoerror function. Changing the shim rod position after equi- librium has been reached merely moves the regulating rod to its mid-posi- tion. This error signal characteristic is to be used for another purpose. It will be noted that changing the shim rods will give a steady state zero error signal. There isno way toraise orlower the meanfuel temperature by moving the shim rod. Furthermore, any slow depletion or poisoning would cause the regulating rod to move slowly out from its mid-position at constatn power. An effect like this should be looked for since it may suggest fuel precipitation in some part of the circuit. It is proposed to use the recently developed pile simulator for deter- mining the Ak/k for the regulating rod. It is likely that this simulator is only reasonably accurate for a circulatingfuel ARE because it has not been considered advisable to simulate transport lags in the system. However, if the simulator is operated with no provision for the temperature coeffi- cient and adequate control is maintained solely with the external servo loop, then the actual control of the reactor with a negative temperature coefficient should also be adequate. It will be possible to determine the maximum Ak/k step for which the system is stable and permits no more than 10% overshoot of meantemperature. Essentially this is done by sim- ulating the MTR servo system so that the servo simulator controls the pile simulator. By this means the Ak/k permissable in the regulating rod can be determined and if it is desired a suitable factor may be employed. With a safety factor in the permissible regulating rod Ak/k and inthe mini- mum period for actuating the scram, the control system would be quanti- tatively described, but not in terms of some arbitrary step function re- sponse. The only arbitrary quantities would be the selection of the MTR servo system and the length and weight of the regulating rod. It might be said that the reactor will be controlled by the servo system to the limits of the system's capabilities and that for more severe control demands the scram system will take over. The safety factors insure that these two systems overlap. Shut Down After the power run has been made, the reactor will be scrammed. When temperatures have reached equilibrium and after the iodine has had time to decay into xenon, an attempt will be made to go critical again. If xenon or iodine is evolvedas a gas from the liquid fuel, the reactor should 124 be made to go critical at essentially the same conditions (as far as rods are concerned) as obtained when it was shut down since the xenon poison will not be present. If, however the reactor will not go critical with shim rod adjustment, no fuelwill be added but additional time will be allowed to elapse and another attempt to go critical will be made. If this latter at- tempt is successful it may be assumed that the xenonhashad time to decay and that the poison on the first attempt was xenon which had not evolved from the fuel. The fuel circuit will then be drained as much as possible but it is ex- pected that some of the fuel mixture will remain in the bends of the fuel tubes. Non-uranium bearing fluorides will then be introduced into the sys- tem to flush out the fuel into the hot fuel dump tanks. This flushing will be continued until the fuel is sufficiently extracted from the system. There- maining fluoride mixture will then be drained (probably leaving some in the core) and the system will be allowed to cool so that physical inspection can be started as soon as possible. 125 ARE BUILDING The ARE building is a mill type of structure designed tohouse the ARE and the necessary facilities for its operation (Figure 35). The building has a full basement 80 x 105 ft, a crane bay 42 x 105 ft, and a one-story serv- ice wing 38 x 105 feet. The reactor and the necessary heat disposal sys- tems are located in shielded pits in the part of the basement serviced by the crane bay. Half of the main floor area is open to the reactor and heat exchanger pits in the basement below; the other half houses the control room, office space, shops, and change rooms. Building Schedule The Nicholson Company, contractor for the Test Facility Building, was notified to proceed on July 6, 1951, and this contract work is now ap- proximately 95% complete. The architect-engineer for this phase of the work was the Austin Company. The AEC is now negotiating an addendum to the initial contract for the remaining portion of the capitalized addition and for some of the noncapital addition. The additional work includes the concrete test pits, concrete heat exchanger pit, basement floor, crane bay floor, additional partitions, process water supply lines, process water disposal lines, settling pond, parking area, and paving. The engineering for this phase of the work was performed by the ORNL Engineering Depart- ment, and it is estimated that the Nicholson Company will complete the work by June 1, 1952. At completion, drawings will be available so that the Laboratory may begin installation of equipment, shielding, and addi- tional process services. Arrangement The building consists of a full basement 80 x 105 feet. In half of the basement are located the shielded reactor and heat exchanger pits; the other half of the basement is servicearea. The control room, office space, and some shops are located on the first floor over the service area. The first floor does not extend over that half of the basement containing the pits, and this area is serviced by the crane bay. The crane is a floor-operated 10-ton bridge crane having a maximum lift of 25 ft above the main floor level. Plan and elevation drawings of the building are shown in Figures 36 and 37. €8001 0LOHd 01-X d314ISSVIONN 931 z —~ Y-F7-140a | | L b g SONTOO, PIT [ B ‘s L/ | W\ < > O lre | wo d | ! W e b NOAONNN NN '-——> 0 X ] 7 P r s P E e ’/.:/_1’/ Lxf Lor b F Sl kS /‘I LIEAT BXCHANGRR PIY_ 5 BAYE SLR B0 Ti;s SNY N iy |—- > 28:c* RUOVEN _LEGEND [ - "] STANDARD Mix SONCQETE o PSS soUn LoNL MARONRY « STD. X 3 — 7 1 JC/ SIS ST s ’? _BOWRD SOPMENT RO 1 [LZ72] oup cone siocks Lao oavsmwg AL LSS 4 2777777 nEMovABLE SPRCIAL Mi% BLOCKS e 1 e I . E% g% o B A w LN & 4 2C@” SLAB STORAGR |_.c |_..c FIRST FLOOR PLAN DASEMENT PLAN FIGURE 36. PLAN OF ARE BUILDING . W bfi" ,@% CRANE BAY Y-F7-i4la NOIE : FOR LBGEND SEE FLODRPAAN DWENTTUAG . O ———— SECTION Ach FIGURE 37. ELEVATION OF ARE BUILDING g T Y 129 The basement includes the test and heat exchanger pits and space for emergency power equipment and direct-current operational power equip- ment. Crane access is provided to the heat exchanger and test pits by re- movable roof slabs. The slabs cover the entire roof area of the test and the heat exchanger pits except for supporting beams. Lifting hooks in each slab enables it to be removed and stored in the slab storage pit, or if con- taminated beyond a usable condition, it may be loaded onto a truck and dis- posed of. The main floor plan shows the control room, general shop, in- strument shop, conference and office space, and toilet and locker room. The western portion or crane bay provides a truck loading area, a slab storage pit, and the removable roof slabs of the heat exchanger and test pits. The building foundation, basement walls, and all floors are of rein- forced concrete. The frame is structural steel with insulated flexboard walls, and there is built-up roofing on precast gypsum slabs. The parti- tions in the service wing are movable and made of insulated metal, and the ones in the basement are of masonryblock. The test pits are provided with process water supply and disposal lines to permit flooding for disas- sembly work. The walls and floor are designed to be watertight. It is planned to further waterproof this structure by application of Amercoat on the interior. Flanged sleeves that may be covered after removal of piping and before flooding are provided through the walls. The shielding of the testand heat exchanger pits is provided by mono- lithic concrete walls and solid masonry concrete blocks stacked dry on the sides and removable concrete slabs on the top. The masonryblocks stacked dry were used to permit greater flexibility for future experimentis. The monolithic concrete walls were therefore held to a minimum thickness consistent with accepted design. Removable blocks within the test pit are shown in Figure 37. These blocks are designed to prevent contamination of the monolithic concrete walls and will be disposed of as required. Utilities The building is to be heated with plant steam, and the pipe lines are being constructed. Fin-type radiators and steam unit headers are used throughout the building. Forced ventilation is furnished to the interior rooms of the service wing. The crane bay is provided with gravity-type ventilators mounted on the roof. The electrical supply to the building consists of an extension of the plant area 13.2-kv line. Both 440 and 110 to 208-v distribution are made within the building. Direct current for all prime operating power will be 130 provided by ac.—dc. rotating equipment. A portionof direct-currentpower will be converted to alternating-current power for instrumentation. Bat- ~ teries will be provided in a floating arrangement to insure continuous op- eration. A 100, 000-gal earth settling pond is provided approximately 1500 ft south of the building and connected to the test pits by a buried steel pipe line. The pond will permit disposal and decay of contaminated water used as shielding in disassembly work. The pond may be emptied by means of a syphon into a branch of White Oak Creek. C ey 131 CONSTRUCTION SCHEDULE Scheduling of the construction of the ARE is obviously afunction of the manpower available. So far as can be determined, the activities of the Laboratory that will be competing for available manpower have been taken into consideration in the preparation of the timetable for the ARE. Every effort has been made to be completely realistic in setting up the schedule. At present all major layout plans are frozen and no basic changes are contemplated. Core detail shop drawings are 99% complete. Heat ex- changers, dump tanks, helium blowers, and valves are being fabricated either by outside vendors or by local shop facilities. Control components (actuators, housings, and absorber rods) are in process of manufacture. Heaters for the fluid circuits are on order. Instruments are about 90% ordered, the remaining 10% are awaiting final detailing as the operational procedure modifications dictate. A schedule of the Experiment is tabulated below: Comgletion Date: 1. Building accepted from contractor June 15, 1952 2. Detailed equipment fabrication drawings July 1, 1952 3. Approximately 60% of components available July 1, 1952 4. Pump for use of fuel circuit tested at design flow and temperature August 15, 1952 5. Cold critical experiment completéd September 15, 1952 6. Installation of electrical system, pipe fitting, and all reactor circuit components completed November 15, 1952 7. Reactor plumbing, ducting, and interconnecting completed January 15, 1953 8. Control room wired and checked January 15, 1953 9. Reactor system hydraulically tested and tight March 1, 1953 10.Reactor dry-run test with inert fuel at temperature, completed and reactor ready for final loading April 15, 1953 11.Critical experiment at temperature followed by run at power May 15, 1953 132 In order to carryout the manydetails of the ARE, the effortis divided into four main categories, each of which is the definite responsibility of one engineer. These divisions are: Building construction Equipment installation Equipment design, procurement, and assembly Operation All operations are under way and although eventualities cannot be foreseen, the ARE should meet the schedule of operation outlines above. 133 CONCLUSIONS It is felt that sufficient basic work hasbeendone to establish the possi- bility of building a circulating-fuel reactor for operation ata maximum fuel temperature of approximately 1500°F for a significant number of hours. Much more work on corrosion and fuel characteristics will be done and significant improvement, such as in the case of corrosionresistance, is anticipated during the construction period of the ARE. The present reactor does not represent an optimum design. Rather, it represents a compromise dictated by several concepts evolving from procurement difficulties. No important principle has been sacrificed in effecting the final design, but in the interest of getting much-needed data, it was decided to get the ARE into operation as quickly as possible. The first effect to be checked on the experimental reactor is the re- activity change as a function of temperature. General experience in the operation of a high-temperature circulating-fuel reactor will be gained. Checks on nuclear calculations that must be used on aircraft reactors will result from the operation of this experiment. Some additional corrosion data will be obtained but should not be significantly different from that which can be taken from less involved experiments. Informationinvolving interaction of the various separate problems, which could not be achieved except by the integrated operation of the entire system, will be obtained. The ARE, in effect, established the first outpost in a new frontier of re- actor design. When design considerations are as critical as those de- manded for aircraft adaptation, they provide a sufficient, if rather nebu- lous, criterion by which to evaluate the worth of the experiment. A second ARE almost certainly should be built, and possibly a third. Each reactor should be planned to supplement the previous one and to bridge the gap as completely as possible between its predecessor and the ultimate aircraft reactor. Materials research should continue at the greatest possible level of effort. Specifically, the fabrication of molybdenum should receive con- siderable attention. Liquid hydrogeneous moderators, such as hydroxides, 134 have been studied experimentally. The results have shown promise but are far from conclusive. Further work on such moderators should be carried on at a high level of endeavor. A second ARE should incorporate improved structure, moderator, and fuel materials so far as these de- velopments warrant. PART III APPENDIXES e ———— - ReEort No. ANP-65 ORNL-1154 ORNL-1170 ORNL-1227 ORNL-1294 Lex P-1 ANP-~52 NEPA-1830 ANP-53 ORNL-1287 Y-F15-10 BIBLIOGRAPHY Title of Report Aircraft Nuclear Propulsion Project Quarterly Progress Report for the Period Ending June 10, 1951 Aircraft Nuclear Propulsion Project Quarterly Progress Report for the Period Ending Sept, 10, 1951 Aircraft Nuclear Propuision Project Quarterly Progress Report for the Period Ending Dec. 10, 1951 Aircraft Nuclear Propulsion Project Quarterly Progress Report for the Period Ending March 10, 1952 Aircraft Nuclear Propulsion Project Quarterly Progress Report for the Period Ending June 10, 1952 Nuclear Powered Flight - A Report to the Atomic Energy Cominission by the Lexington Project Report of the Technical Advisory Board to the Technical Committee of the Aircraft Nuclear Propulsion Project Final Status Report of the Fairchild NEPA Project Report of the Shielding Board for the Air- craft Nuclear Propulsion Program Preliminary Investigation of a Circulating- Fuel Reactor System for Supersonic Aircraft Propulsion Several Circulating-Fuel Reactor - Heat Exchanger - Shield Designs for Supersonic Aircraft Author(s) R. C, Briant C. B. Ellis W, 3. Cottrell R, C. Briant C. B. Ellis W. B. Cottrell R. C, Briant W.B. Cottrell R. C. Briant A.J. Miller W. B, Cottrell R. C. Briant A.J]. Miller W. B, Cotrell R. W. Schroeder A. P. Fraas G. F. Wislicenus 137 Date Issued Sept, 13, 1951 Dec. 14, 1951 Feb. 18, 1952 May 3, 1952 to be issued Sept, 30, 1948 Aug, 4, 1950 no date, circa late 1951 October 16, 1950 to be issued to be issued 138 Y-F10-103 ANP-064, Part 1 ANP-64, Part II _"_“-' B Statics of the ARE Reactor, Summary Report C. B. Mills The Technical Problems of Aircraft Reactors, C. B, Ellis Part I The Technical Problems of Aircraft Reactors, C, B. Ellis Part 11 to be issued June 1, 1951 Jan, 2, 1952