tr I OFL. ORNL P 2491 1 * S + TT " 0 : 14 . wwwe ! TERTI .. MICROCOPY RESOLUTION TEST CHART NATIONAL BUREAU OF STANDARDS - 1963 . RELEASED FOR ANNOUNCEMENT LOR NL-P-2491. CoNL-660925-1 CISTI PRICES IN NUCLEAR SCIENCE ABSTRACTS V -SE SEP 2 6 1966 *** * 2 . THE OAK RIDGE HIGH FLUX ISOTOPE REACTOR* H.C. $ 9.0 MNO DESIGN AND INITIAL OPERATION ..MASTER T. E. Cole, Oak Ridge National Laboratory, Oak Ridge, Tennessen ABSTRACT: The High Flux Isotope Reactor (HFIR) was designed and constructed at the Oak Ridge National Laboratory (ORNL) primarily for the production of trans- uranium elements for use in the heavy-element iesearch program of the United States /17. The central target assembly will initially contain about 250 g of Pu242 which will be irradiated in the ~2 x 1015 n/cm2.8 flux of the HFIR to pro- duce significant quantities (tens of milligrams) of the californium isotopes in addition to larger quantities of other heavy elements Guring the first year of irradiation. Although the primary purpose of the HFIR is the production of transuranium elements in the high flux region, its usefulness f.s increased by several other ex- periment facilities which are located in, or terminate in, the reflector. These include four horizontal beam tubes, four slant access Cubes, and 38 vertical holes of various sizes. The HFIR is designed to operate at a maximum power level of 100 MW. The fueled region is a cylinder having a volume of about 50 liters which leads to an average power density of about 2 Mw per liter. The 9.4 kg of u235 fuel is con- tained in curved, aluminum-clad plates 0.050 in. (1.27 mm) thick separated by 0.050 in. (1.27 mm) coolant channels. Ordinary water is used for coolant and moderator, and beryllium metal is used as a reflector. The HFIR first achieved críticality on August 25, 1965. Following an ex- tensive "zero-power" program the power level was increased to 20 mw on January 29, . 1966. Operation for a full fuel cycle at 75 Mio was completed on July 31, 1966, and it is anticipated that the design power level of 100 Mw will be achieved in early fall of this year. The HFIR, including the building and auxiliaries, cost slightly under $15 x 106, exclusive of development cost. Development costs were approximately $6.7 x 105, of which about $2.5 x 106 was incurred in der elopment of the fuel element. The annual operating cost, including the cost of fabricating the fuel elements, but not the cost of the fissionable material consumed or of reprocessing to recover the urburned fuel, will be about $3.8 x 106. 3 -* 22 mu CU . " . *Research sponsored by the U. S. Atomic Energy Commission under contract with the .- 1. S o mo .********** 1. INTRODUCTION The general features of the HFIR facility have been described previously L2,37. Therefore, this paper will present only a summary description, taken pri- marily from reference 27, followed by information on the startup programa, de- velopment, construction and operating costs, plans for utilization oi the experi- mental facilities, and a few comments regarding the possibility of achieving a higher neutron flux with a reactor of this general type. 2. REACTOR CORE Investigation of various reactor types and plutonium target designs indi- cated that a flux-trap type of reactor operating at ~100 Mw could produce the necessary Lhermal and fast neutron flux to meet the requirements of the heavy element research program. The desire to minimize development, capital, and opera- ting costs led to selection of aluminum-clad fuel plates, normal water coolant and moderator, and a beryllium metal reflector, arranged in a core of cylindrical geometry. The nuclear characteristics are summarized in Table I. A cross-section of the reactor in its pressure vessel is shown in Fig. 1. Table I General Nuclear Characteristics of the HFIR 100 100 Reactor power, Mw Neutron fluxes, n/cm2.s Maximum unperturbed thermal flux in island Average thermal flux in target (300 g of Pu242) Average nonthermal flux in target Maximum nonthermal flux in fuel region 5.5 x 1015 2.0 x 1015 2.4 x 1015 4.0 x 1015 Maximum unperturbed thermal flux in Be reflector Beginning of fuel cycle End of fuel cycle 1.1 x 1015 1.6 x 1015 35 70 Prompt-reutron lifetime, us Beginning of cycle End of cycle Effective delayed neutron fraction Fuel loading, kg 1235 Length of typical fuel cycle, days 0.0071 - 2 - VESSEL SUPPORT RADIAL BEAM TUBE ( TYPICAL OF 4) FACILITY ENGINEERING PRESSURE VESSEL SUB PILE ROOMS . REGION FUEL 1000 -3. (Vertical Section) Fig. 1. High Flux Isotope Reactor T 1 - - ✓ OUTLET ::: ASSEMBLY TARGET FACILITY VERTICAL IRRADIATION INNER CONTROL DRIVE 1 1 DRIVE (TYP OF 4) 0 OUTER CONTROL ws INLET ImUZ SIA (TYP. OF 2 REFLECTOR ABERYLLIUM, BEAM TUBE TANGENTIAL ACCESS HATCH CALE . ORNL.LR-DWG 61384 UNCLASSIFIED 1 . ...:::: : min LEGAL NOTICE The report we prepared u an account of Governmeat sponsored work. Neither the United Hote, me the Commission, nor any porno roting and behalf of the Commission: A. Maka may warrauty or rapreneatation, expressed or implied, with respoot to the acou- moy, com piscGSı, or unfaissus of the latormation contained in this report, or that the we of my taformation, apparıb, method, or proosas desloued in this report may not lattwago patrately owned states or 1. Annum.. any liabilues with respect to the ww of, or for damago rinting from the ure of Informadon, apparna, bmthod, or procou dinoloned in this report, de wind in the above, pornon setting on behall of the Commisstan" inoludas wy on- plogue or contractor of the Commission, or employto of much oontrnotor, to the extent that wol omployu or contractor of the Commission, or employee of muok oontractor preparu, dissonbilan, or provides nodos lo, cay taformation purmeat to Ho employment or contrast wid the Countestoa, or ho employment with suod contructor. The cylindrical HFIR core consists of four concentric regions. The central region, or flux-trap, contains normal water and 18 5 in. (12.7 cm) in diameter, extending the length of the core. The fuel region located immediately outside the flux-trap consi8ls of two cylindrical fuel elements containing U235 in vertical, curved plates. The fuel is surrounded by a cylindrical beryllium reflector, 12 in. (30.5 cm) thick. A normal water reflector of effectively infinite thickr.288 sur- rounds the entire core. Control plates in the form of two poison-bearing concen- tric cylinders are located between the outer fuel element and the beryllium.' 3. FLUX-TRAP PARAMETERS Three major factors control the magnitude of the thermal neutron flux in the flux-trap: (1) the amount of neutron leakage from the fuel to the trap, (2) the slowing down and absorption characteristics of the trap moderator, and (3) the diameter of the trap-region. An optimum trap diameter exists because of the con- flicting requirements for complete moderation (lerge trap-region) and for high neutron density (small volume). A moderator having the shortest slowing-down length provides the highest thermal neutron flux. Even though normal water has a relatively high thermal-neutron-absorption cross section, its slowing-down distance is sufficiently small to result in a higher flux in the trap-region than attainable with other moderators. Neutron leakage from the fuel to the flux-trap is a function of core shape, core size, the metal-to-water ratio, power distribution, and the amount of para- sitic absorption. Decreasing the core size and increasing the metal-to-water'.'. ratio increased the core leakage. Optimizing the length-to-diameter ratio of the fuel region increased to a maximum the total leakage per unit core height into the island. The limitation on reduction of the core volume was associated with heat removal characteristics. A significant improvement in reactor performance was attained by incorporating features which reduced the ratio of maximum-to-average power density (maxlaque) to 1.45 (exclusive of hot-spot factors). These features include cylindrical core geometry, a symmetrical reflector control system, and radial variations in the fuel and burnable poison concentrations. By comparison with a uniform fuel distribution, the radial Laxlave was reduced from 3 to 1.2. $9.33 . Joonasoo pone 22. form 20 4. FUEL ELEMENTS The compact fuel assembly of the HFIR and its dimensions are shown in Fig. 2. The inner element, which contains 171 plates, is initially loaded with 2.6 kg of v235 and 2-4 g of B20 as burnable poison. The outer element contains 369 plates, initially loaded with 6.8 kg of U235 and no burnable poison. The in- dividual plates are of a sandwich-type construction with a fuel-bearing cermet . m . : 2. ORNL-LR-OWG 61472AR3 - OUTER ANNULUS, 369 PLATES - INNER ANNULUS, 171 PLATES NOTE: NOT TO SCALE ALUMINUM ADAPTOR - TA / / 2:::..... 412 in - LUI IIIIIIIIIII IIIIIII LIN III IND TIL 1111111 III IN 1 11 IIIIII -.. - MINIT .•.••• UUDII ---- - OVERALL LENGTH 31%8 in. FUEL PLATE LENGTH 24 in. TI ACTIVE FUEL LENGTH 20 ira. LLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLLL ܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠܠ ---- ••••••••••• ZITIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIII - UIT I HUU ·-- .-- : 1 THI ........ LIIT TITUUT UUTIS IIIIIII IIUMI IAITU 1 IIIIIIIII MIUININ III TITUUT IIIIIIII -- --- U11 VIII -------...---•••••• . IN UL .. uch ... - 5.067-in. DIA --- IIIII -10.590-in. DIA 11.250-in. DIA 17.124-in. DIA ----- 0.050-in.-THICK PLATES 0.050-in.-THICK COOLANT GAPS Change last figure to 17.134. - 1. Fig. 2. HFIR Fuel Element - 5 - .. . . i inni www.T win White . W our ! Y . When sealed between covers of 6061 aluminum. The cerüet is a mixture of aluminum and U30e, 35% U308 by weight in the inner plates and 40% in the outer. The finished plates are 0.050 in. (1.27 mm) thick, including a minimum cladding thickness of 0.010 in. (0.25 mm). The plates have an involute curve which provides a cooling channel between plates of constant width, 0.050 in. (1.27 mm). The plates are welded to the cylindrical aluminum side plates. The radial variation in the fuel and poison concentrations is obtained by varying the thickness of the cermet cores across the width of the fuel plate as illustrated in Fig. 3. The B20, in the form of B4C, is added to the aluminum filler pieces of the inner element fuel plates. VAR. . S-. rz t isex *runstw * 5. CONTROL PLATES The control "rod" system of the HFIR was selected primarily for 'its ability to control reactivity adequately without introducing undesirable perturbations and asymmetries in the power distribution or in the neutr'n fluxes of the flux-trap. The system constitutes a reflector control system that regulates the flow of ther- mal and near-thermaí nenit cons from the beryllium reflector to the fuel region. As shown in Fig. 4, the narrow annulus between the fuel and the beryllium reflec- tor contains two thin, concentric cylinders which are separated from each other and from the adjacent regions by narrow coolant gaps. The inner cylinder has a single drive rod and is used for both shim and regulation. The outer cylinder is divided into a. drants, each with its own drive rod and release mechanism; these four control piates are used for both shim and safety. During normal operation, the four shim-safety plates and the shim-regulating cylinder are moved in unison to avoid asymmetries in power distribution. When used in an emergency, the shim- safety plates are released separately, thus nroviding multiplicity for shutdown. To hold variations in the longitudinal power distribution within acceptable limits and to prolong the neutron absorber life of the cods, the control cylinders are divided into three discrete longitudinal regions: a highly neutron-absorbing (black) region, a moderately absorbing (gray) region, and a comparatively poorly. absorbing (white) region. By locating the black regions of the two control cylin- ders at opposite ends of the core (Fig. 4) and by moving the cylinders in opposite directions symmetrically, it is possible to maintain power distribution symmetry about the horizontal midplane of the core. Materials selected for the control plates are a 33-volume percent Eu2O3-A1 dispersion clad with aluminum for the black region, a 40-volume percent Ta-Al dispersion clad with aluminum for the gray region, and solid aluminum for the :: white region. " . : -6. y ÜRNL-UWG 64-4971 INNER FUEL ... ANNULUS OUTER FUEL -.. ANNULUS 6061 AH Nil B&C + AI il: TUTHAITH I!! i liliii!? UzOg + A1 0.050 in. v • 7. -- - - - . 0.050 in. U30g + A1 - ... . –H20 ISLAND -- -- -- - - - - --- - Fig. 3. Schematic Representation of Core Cross Section, Showing Fuel Contours ORNL-OMG 63-0100N SECTION A-A - . - OF SYMMETRY REGULATING-SHIM CYLINDER >SHIM-SAFETY QUADRANTS (4) - CONTROL REGION SHROUD . RE We -Eu2O3-A1 BLACK REGION -TO-AI GRAY ASGION OUTEA FUEL ELEMENTY ISLAND INNER FUEL ELEMENTI HORIZONTAL MIDPLANE Hyo v BERYLLIUM REFLECTOR . FUEL PEDESTAL ::.:.:.:.:.:.: Town .:::: WHITE REGION REFLECTOR PEDESTAL ::: SHIM-SAFETY DRIVÉ ROO (4) REGULATING-SHIM DRIVE ROO Hae t . . 32 Fig. 1. Schematic Representation of Control Rod Arrangement in HFIR; Rod Position Shown Approximate for Clean, Critical Core -.8 - . 6. BERYLLIUM REFLECTOR Because of the high performance required of the HFIR and because of its sınall size, the choice of reflector was limited to beryllium or D20. Beryllium was selected initially because of generally satisfactory, long-term experience with beryllium reflectors, the advar.ced state of beryllium fabrication, and the relatively complex auxiliaries necessary to cool D20. Although tue magnitude of the deterioration of beryllium in the Materials Testing Reactor (MTR) did not be- come known until design of the HFIR was well along 14,57, the reflector had been designed as three concentric regions with provision for simple replacement of the inner region. The MTR experience indicated that exposure of the HFIR beryllium should be limited to 1-2 x 1022 nvt of fast neutrons (E > 1 Mev) unless substan- tial mechanical distortion and/or cracking could be tolerated. Because the inner region of the beryllium would accumulate this exposure in a few months, revisions in the design were made to preclude operational difficulties which might arise from cracking or distortion of the beryllium before replacement. 7. HEAT TRANSFER AND CORROSION Careful calculation and experimental verification of heat-transfer charac- teristics of the HZIR core were required because of the desire to extract as much power as practical from the smallest practical core, As 'indicated in Table II, the configuracion that was selected leads to an average power density in the core of almost 2 MW.2, and an average fuel element heat flux of 7.7 x 105 Btu hr-1 ft-2 (58 cal s-1 cm-2). over that corresponding to the onset of nucleate boiling at a hot spot. Calcula- tions of this margin take into account the fuel plate tolerances, creep of the fuel plate during operation, fuel loading variations and non-bonding in the plates which may be below the sensitivity of detection. Measurements of heat transfer rates, burnout heat flux, plate deflection resulting from thermal and hydraulic forces and flow through the fuel assembly have been made to verify the calcula- . tions used in the heat transfer design. The ability to use aluminum for fuel plate cladding and core structural material depends upon satisfactorily low fuel plate and interface temperatures, as well as adequate corrosion resistance to the water coolant. With respect to the latter, not only corrosion per se, but also the formation of aluminum oxide films on the fuel plate surfaces which would adversely affect heat-transfer rates, were of concern. An experimental corrosion program established that aluminum, particularly type 6061, should be satisfactory under the anticipated conditions provided the acidity of the water coolant is adjusted to about pH = 5. Under these conditions, Table II Heat Transfer and Removal 100 97.5 50.6 Total reactor power, MW Power in fuel region, Mw Volume of active core, liters Heat transfer area Coolant velocity through core Inlet water temperature Exit water temperatuce, average Coolant flow through fuel channels Coolant flow, total Coolant inlet pressure 40 15.5 49 73.3 ~812 428.8 51 120 m2 m s-1 °C °C 28-1 & 8-1 ft2 ft s-1 °F 120 ~1000 ~13,000 ~16,000 ~ 600 gpm gpm psig 41 atm Power density, Mwll Average Hot spot 1.9 4.3 Heat flux Cal sol cm-2 53 Average Hot spot Btu hr-1 ft2 7.7 x 105 2.0 x 106 151 Incipient boiling power level, MW >130 Temperatures, °F (with 121°F inlet water temperature): Seart of Cycle End of 15-day Cycle Fuel Fuel Average Hot Streak Outlet Hot Spot Hot Streak ot Average Outlet Hot Spot Fuel plate temp, °F 205 Metal-oxide interface temp, °F 197 Oxide-water interface temp, °F 197 Water outlet temp, °F 178 312 303 303 246 390 371 371 246 212 2 04 197 178 360 351 303 246 480 461 . 377 246 *** . corrosion is expected to be satisfactorily low and film growth during the expected life of the fuel plates should be sufficiently low that acceptable fuel plate temi- peratures will be maintained. The oxide buildup results in an appreciable increase in plate temperatures, as evidenced by the hot-spot metal temperature (Table II) which increases by about 100°F (55°C) during the fuel cycle. Since plate strength is more closely related to plate average rather than hot-spot temperatures, it appears from these data that plate strengths will be adequatė throughout the fuel cycle. While the . KE ... . - 10. available data on oxide buildup and irradiation damage 167 indicate that operation at 100 MW should be possible, it is in this area that the major uncertainties re- garding performance are present. The combined effects could possibly lead to the necessity to limit the power level and/or core lifetime until modifications to the fuel element design are made to counter the problems encountered. As will be described in a later section, the approach-to-power phase of the program includes the operation of the reactor for a full fuel cycle (until shut- down is caused by fuel depletion) at each of several successively higher power levels to demonstrate fuel element behavior under conditions of approximately the same fuel burnup but at successively higher fuel plate temperatures. 8. TARGET ASSEMBLY The target assembly, shown in Fig. 5, is located in the flux-trap region and contains the Pu 242 target material. The thirty-one 3/8-in. (0.95-cm) OD aluminum tubes are filled with pellets made from a mixture of Pub, and aluminum powder. The tubes are capped and collapsed onto the pellets to provide good heat transfer. These are centered by fins integral with the tube inside shrouds which define the coolant channels. The shrouds are clamped securely in a framework which accurately positions the tubes and facilitates femoval and maintenance operations. For an initial charge of 300 g of Pu 242, which is calculated to be about optimum for maximum heavy isotope production, the heat generated in the target will be about 900 kw. This is removed by a flow of about 675 gal min-1 (42.6 18-2) around the target rods, and about 113 gal min-? (7.1 18-2) between the target assembly and the inner fuel element. 9. EXPERIMENT FACILITIES Although the primary purpose of the HFIR is the production of transuranium elements in the flux-trap region, its usefulness is increased by several other experiment facilities which are located in, or terminate in, the reflector. These , include four horizontal beam tubes, four slant access tubes called "engineering facilities," and 38 vertical holes. Figure 6 shows the general arrangement of these facilities in the core region. A hydraulic tube for irradiation of small samples in the target region is planned for future installation. This tube would replace the central target rod and would permit short irradiation in the very high flus region without interfering with operation of the reactor. The four nominally 4-in. (10-cm) ID horizontal beam tube experiment facili- ties extend outward from the reactor core at the midplane. One beam tube extends radially from the reactor center line with its inner end penetrating the permanent reflector. Another tube extends tangentially from the core, offset approximately . • 11 - UNCLASSIFICO ORML•LR.O .49750 HOLD DOWN RING CLAMPING BOLT . TARGET CLAMP TE TARGET SUPPORT RING . S L WHAVIA FLOW CHANNELING CAN % ACROSS FLATS " - FUEL ELEMENT INNER SIDE PLATE Se TARGET ROD %' DIA. 35' OVERALL LG. 20° ACTIVE LG. 31 RODS TY . 7 edias brownish personisso ACTINIDE PELLETS V DIA. i' LG. 40 PELLETS PER ROD . MIELUSTITIVO -- GRID PLATE HANGER Philibrio . 1 ROD : MAI iwand TARGET GRID PLATE- 1 .44 Fig. 5. HFIR Target Assembly • 12 - ORNL - DWG 63-4084R REACTOR PRESSURE VESSEL- Ć REACTOR -COOLANT DISCHARGE (TYPICAL) HB-1 - HB-4 -THRU TUBE REMOVABLE REFLECTOR -PERMANENT REFLECTOR TARGET INNER FUEL ELEMENT - SEMI-PERMANENT REFLECTOR- -- É REACTOR - REFLECTOR LINER CONTROL PLATES CONTROL PLATE ACCESS PLUG TANGENTIAL TUBE EXPERIMENT ACCESS OUTER FUEL ELEMENT ISOTOPE IRRADIATION ACCESS HB-3 HB-2- "RADIAL TUBE Fig. 6. Horizontal Section Through Pressure Vessel - 13 .. - 13 - : 10.5 in. (26.7 cm) from the reactor center line. It also penetrates the permanent reflector. The remaining two tubes are aligned on a tangential line approximately 15 in. (38 cm) from the reactor center line with ends entending outward from the reactor; they are arranged to allow the installation of a single through-tube, should this be desirable in future experimental programs. Design of the beam tube installation was subject to the criterion that they were not to significantly affect the power distribution in the fuel region. In order to meet this criterion it was decided that about three inches of beryllium should separate the fuel re- gion from the beam tubes and the design proceeded on this basis without a signifi- cant effort to investigate the optimization of beam tube locations. Early con- . cepts of the design were based on radial orientation of the beam tubes; however, most experiments being considered were of a type using thermal or only slightly epithermal neutrons. Experiments conducted by the Naval Research Laboratory 407 and Brookhaven National Laboratory [87, and calculations at ORNL 97 showed an ad- vantage in terms of reducing the very high energy neutron and gamma rays from the fuel region and therefore three of the tubes were changed to tangential. One radial beam was retained primarily for those experiments in which the higher energy radiation would be of interest. Each tube 18 sealed to, and supported by, the reactor pressure vessel by means of a system of clamped. and bolted flanged joints. From the flanged connec- tion at the pressure vessel, each tube continues through the reactor pool and pool wall and terminates in a recess located in a large beam port cavity in the shield wall. This recess allows the installation of a shutter and shielding and also provides space for experimental equipment. Provision is made for the future installation of four engineering facili- ties to accommodate experiments that require a relatively low neutron flux. These facilities consist of 3-1/4 in. (8.25 cm) ID tubes which enter the pressure vessel and extend downward at an angle of approximately 41° to provide access to the flux at the outer periphery of the beryllium. The upper ends of the tubes terminate at.. the outer face of the pool wall in the experiment room which is located on the level above the beam tuve level. The permanent reflector is penetrated by 22 vertical holes which extend completely through the beryllium. In addition, 16 similar vertical holes are pro- vided in the removable beryllium. These vertical holes provide space for various types of static irradia- tions. In addition, it is anticipated that several of them may be used as loca- tions for future hydraulic facilities which will permit the insertion and removal of samples during reactor operation. When not in use, the vertical facilities will be filled by beryllium plugs sized to provide an adequate cooling annulus between the hole liner and the plug. - 14 - REACTOR BUILDING AND AUXILIARIES The main reactor building, associated electrical building, cooling towers, filter pit, exhaust stacks, and maintenance-office building are shown in Fig. 7. Detailed descriptions of these facilities are given in reference 37. 11. CORE PHYSICS Because of the desire to achieve unusually high performance in the HFIR, an extensive critical-experiment program was conducted in parallel with the ana- lytical studies. However, since the fuel region consisted of only two fuel ele- ments, and since the radial fuel distribution could not be readily modified, the core was not amenable to parametric-type experimental analysis. It was therefore necessary to rely upon calculations for determining the proper island diameter and fuel region metal-to-water ratio and length-to-diameter ratio, and to some extent the fuel loading and fuel distribution. It was also necessary to rely upon cal- culations for determining fuel-cycle characteristics such as the power distribu- tion and core lifetime. In these instances, the role of the critical experiment was relegated to that of verifying the adequacy of the calculational methods. The following parameters were investigated in the critical experiments: power distribution, flux distribution, control rod reactivity characteristics, fuel worth, temperature, void and fuel coefficients of reactivity, plutonium target characteristics, and neutron 11fetime. Most of the analytical studies were made with one-dimensional, multigroup, and two-dimensional, two-group, diffusion cheory reactor codes 10,117 because they provided the only practical means for making fuel cycle calculations; good agree- ment was obtained between analytical and experimental power distributions. Neutron cross sections used in the above calculations were obtained using a multiregion, thermalization, transport-theory code (127 for the thermal neu- trons, and a single-region, multigroup, consistent By theory code (137 but the non- thermal neutrons. The use of the thermalization code made it possible to obtain spectrum-averaged thermal neutron cross sections as a function of radial position. in the fuel region. After the first two critical experiments established the adequacy of the calculational methods, fuel cycle calculations were made to help determine the proper fuel and burnable poison loadings and distributions. The final results indicated that an average fuel cycle time of 14 days could be achieved with an initial v235 loading of 9.4 kg, and that the maximum power density and the thermal neutron flux in the target would be essentially constant throughout the cycle. A fuel loading was then fabricated to these specifications and a third set of cri- tical experiments was run to verify the results of the calculations. The element was constructed in such a way that the fuel plate coolant channels could be filled . - 15 - Hle - 2 1 -- . ... 16 . * * mi - at we HIGH FLUX ISOTOPE REACTOR U.S. ATOMIC ENCOGY COMMISSION LAK 21055. INNESSEE SINGHASTER E BREYER, ENGINEERS NEW YORK, NEW YORK Fig. 7. High Flux Isotope Reactor . (Use photo with buildings labeled 63-6034? from Geneva paper). G + ---- - -. - - - - - Y a A s osivt z ua -tua - UN -*- . . - -" : . ... i V with various solutions in order to uniformly poison the fuel region without changing the condition of the flux trap or reflector regions. This element was also febricated to correspond, in essential dimensions, to the dimensions speci. fied for ihe reactor fuel elements in order that it could be used in the reactor for low-power experiments during the startup program. Typical radial neutron flux distributions across the core are shown in Fig. 8. 12. STARTUP AND APPROACH-TO-POWER PROGRAM The program 1.8 divided into two phases and covers operation of the reactor from initial criticality through 100-Mw operation. The first phase covers opera- tion through 20-MW power level and 18 termed the startup program. The second phase 18 termed the approach-to-power program and covers operation from 50 MW through 100 MW. Startup Program The 20-Mw end point for this phase was selected on the basis that the power level would be sufficiently high to bring all normal instrumentation into operation while still being sufficiently low that the heat transfer requirements were essentially negligible in erms of the design. On August 25, 1965, the HFIR achieved criticality for the first time, and on January 29, 1966, the power level was increased to 20 MW. Operation at or near 20 MW was continued until May 3.., 1966, at which time the power level was increased to 50 MW. During the period from August 25, 1965, to January 29, 1966, an extensive nuclear "zero power" experimental program was conducted for the purpose of inves- tigating the nuclear characteristics of the actual HFIR facility, for compiling generally useful data associated with the routine operation of the reactor, and for operator training. The types of experiments conducted for the purpose of investigating the nuclear characteristics are as follows : (1) Reactivity shutdown margins. (2) Control rod reactivity differential worth. Fuel, fuel plate, and void reactivity coefficients for the fuel regions. (4) Isothermal temperature reactivity coefficients. (5) Reactivity worths of various targets and voide in the flux trap. (6) Reactivity worth of water in the beam tubes. (7) Reactivities associated with replacement of beryllium reflector ver- tical experimental facility beryllium plugs with water and voids. (8) Effect of flow and pressure on reactivity and reactivity stability. (9) Reactivity worth of B20 stainless steel strips in the fuel elements. (secondary shutdown). . - 17 - XY ... . - . - . . . - . - :-- a CA ORNL-LR-DWG. 75092 Н20 CONTROL RODS H20 201 [A1+H20 -30 NON-THERMAL NEUTRON FLUX (neuts/cm2-sec) THERMAL -18-' VERTICAL DISTANCE (cm) . . ISLAND U FUEL FUEL H20 | Al+H2O||||AI + H2O REFLECTOR Be + ~5% H20 10 20 30 RADIAL DISTANCE (cm) Fig. 8. Core Schematic and Flux Distribution (Clean Core at 100 Mw) (10) Power distributions (11) Response of safety system to neutron-generated signals. In the zero power experiments several combinations of fuel elements, con- trol rods, and targets were used. The fuel elements included the critical ex- periment element (9.4 kg U235, 2.12 8 B10), which 18 referred to as the HFIRCE-3 element, the ORNL production element (9.4 kg U235, 3.60 8 B10), and the first element fabricated by Metals and Controls, Inc., 19.4 kg U235, 2.80 8 319). The only significant difference between these elements 18 the loading of the B20 hurnable poison. Subsequent elements are presently specified to contain 2.80 8 B20. Two sets of control rods were used: the HFIRCE-3 rods and the first pro- duction rods. With the exception of a few hydraulic relief holes these two sets of rods are identical. The different targets used in the flux trap are listed in Table III. TO avoid the possibility of voids entering the flux trap during the early experiments, two different plastic containers were used to displace the water in the flux trap. One contained the HFIRCE-3 simulated 300 g Pu 242 target* plus the corresponding optimum void space; the other was provided with the optimum void space that cor- responds to no target in the flux trap. Both of these plastic containers could be voided or filled with water. Table III Targets Used in Flux Trap Nomenclature PI + W PI + V PT + W PT + V + Description Plastic container filled with water Plastic container with optimum void Plastic container with HFIRCE-3 target and filled with water Plastic container with HFIRCE-3 target and optimum void Dummy flow test target (all aluminum) Standard target using U235, 1238, Ta to simulate maximum heat generation rate Standard target containing 248 g Pu242, 2.5 & Pu239, 2.5 8 Pu 242 DFTT U235 Rods Pu 242 Rods In view of the fact that much time and effort was spent on HFIR critical experiments at the critical facility, there is perhaps some question as to why so many experiments were proposed for the HFIRCE-3 fuel element and control rods in the HFIR facility. There are two main reasons. One is that near the completion *The HFIRCE-3 target contained u235, u238, Ag to simulate a maximum reace, tivity 300 & Pu242 target. . - 19. V of the HFIRCE-3 experiments a change was made in the design of the coolant chan- nels in the control region of the HFIR. The effect of this change was to decrease the worth of the control rods; thus it seemed prudent to rerun many of the rod calibration experiments in the HFIR facility. The other reason stems from the fact that the HFIRCE-3 experiments were terminated before all of the desired power distribution data were obtained. The data of particular interest 18 that which can be used to predict the power distri- bution variations in the vicinity of the control rods (window peaking effects) for various control rod positions. Although this additional information is not neces- sary for evaluating the present fuel element design, it will be needed for extrap- olation to more heavily loaded elements should such changes be desirable in the future. There are, of course, other reasons for conducting the HFIR experiments such as operator training and compilation of generally useful data associated with the routine operation of the reactor. Furthermore, there are "reactivity" experi- ments associated with coolant flow and pressurization that could not be performed in the HFIRCE-3 facility. Differential worth of the control rods was determined for different rod positions by maintaining criticality with the rods (asymmetrical rods) and also by supplementing the control with boron in the confined moderator (~symmetrical rods). The former type of rod calibration was performed at the HFIR facility before any of the boron experiments were conducted. Therefore, these experiments provided data with which to compare the reactive state of the IFIRCE-3 and HFIR facilities. A summary of these data is presented in Fig. 9. As indicated, the agreement be- tween the two facilities is very good. A comparison of the dashed curves, repre- senting the HFIRCE-3 data, with the solid curves presumably shows the effect of the different coolant channels in the control regions of the two facilities. As was expected, this difference tends to reduce the worth of the safety rods in the HFIR facility. The differences observed in the differential worth curves are probably within the degree of accuracy associated with the experiments; however, the rod position data should be accurate enough to indicate a real difference. Power distributions were determined in essentially the same way as was done during the HFIRCE-3 experiments, using the same removable fuel plates which con- tained "punchings" used as fission foils. These punchings were made at selected locations and had a fuel content typical of the fuel loading at that location. The foil (punching) weights (weight of u235) were determined using a scintillation counter. Relative power distributions were obtained by comparing each foil's total gamma activity to the time-interpolated activity of a normalizing foil that had been irradiated at the same time and was counted periodically during the counting of the other foils. .. 20. CLUB 140 20 X 20 TO THE INCH 7 X 10 INCHES xaurroL CSSER CO. 46 1242 NASE 14 O., de 1 o do i ó . ó RE 300 THE INCH 481242. § i t iso no ROD DIFFERENTIAL WORTH (4/11) 0 o. 2 . H. LI . . - OTO00 ON. 1 1 D AT . . . : - 21 - MEILLIG BURGHAIRCARE RODS : 6 8 ROO 10 12 Fig. 9. HFIR Differential Rod Worth Data pasitieN 14 16 (IN) 18 20 22 24 26 28 THE Tron PARLEROINH 1 Mis IUTOIli - in !!11:11CDI . . . . . - 0 - - o d 0. ŏ di Î ã ã Ô SAFETY ROD POSITION (IN) ST & OK The results of the power distribution measurements confirmed the work done at the critical facility and provided the additional information needed for studies of possible future loadings. A few general comments regarding azimuthal and longitudinal symmetry may be of interest. There is essentially no effect of beam holes on power distribution, but the window between control rod poison quad- rants and the greater water concentration in that area increases the local powe: density by as much as 25% and by no less than 5%. The latter value exists with the rods fully withdrawn, in which case the peaking presumably results from the greater water concentration associated with the stationary divider strip between the safety rods. The largest peaking occurs with the rods in their innermost critical position (clean core plus maximum fission target). These effects will have to be considered when evaluating the overall window peaking problem for fu- cure cores. However, the present core design has sufficient margin in the area of interest to accommodate these local peaks since they are no different than observed in previous experiments. In the longitudinal direction there is a tendency for the power density to be greater at the coolant inlet end than at the outlet because oi the slight lon- gitudinal asymmetry associated with the radial displacement of the shim-safety rods relative to the shim-regulating rod. The effect 18 more pronounced for the clean core condition than for the fully withdrawn-rod condition because as the rods are withdrawn the "reflector" asymmetry decreases. A preliminary analysis of the data indicates that for the clean core, symmetrical rod condition the power density at the outlet end of the core is ~17% less than at the inlet end for the inner element and ~28% less for the outer element. When the rods are fully with- drawn, the corresponding values are ~3% and -11%. This agrees reasonably well with previous data and even after making a correction for inaccuracies indicates a pos- sibly significant improvement in heat removal in the early part of a fuel cycle. Fuel and void reactivity coefficients were obtained for the inner and outer fuel elements by replacing regular fuel plates with water and by replacing non- fuel-bearing aluminum fuel plates with water. These plates occupied the regular removable fuel plate slots; there were six nearly equally spaced inner and outer element plates involved in both tests. The reactivity worth of replacing these plates with water was obtained by withdrawing the appropriate number of plates with the reactor initially critical and determining the positive stable period. Table IV summarizes the reactivity worth of several changes in target arrangements in the reactor and the fuel, fuel plate, and void reactivity coefficients of the inner: and outer fuel elements. The reactivity changes resulting from flooding of the beam tubes and re- placement of the reflector beryllium plugs with water and voids, as shown in Table V; are negligible. :: . - 22 . .. -- - - - - - -- - - - - - -- - - - - Table IV Summary of Reactivity Worths and Coefficients (PT + W) - (PI + W) + 0.0070 Sk/k (PT + v) - (PT + W) + 0.015 Sk/k (PI + V) - (PI + W) a + 0.032 Sk/k Fuel Coefficient Ak/k Iriner Element + 3.72 x 10 + 0.0966 11235 ) Am/m + 3.72 z 10*5 metres+ 0.0966 + 1.09 :: 10's anything + 0.074 et +0.171 Stelle Akk Outer Element + 1.09 + 0.0744 ak/k Am/m 8 0235 y + 0.0 Total Average Fuel Plate Coefficient Inner Element .72 x 10 io - 0.0458 - .x 10-6 slip - 0.13 May - 11.6 x 10-s AI Outer Element 0.113 .. 11.6 x 10° ) AV/v in 3 Fuel Region Al Coefficient Inner Element - 0.104 Style, - 16.7 x 10-5 Aksel in. 3 Outer Element - 0.188 Ak/k vlü, - 15.9 x 10-5 Ak/k in, 3 Fuel Region Void Coefficiente Inner Element - 0.080 Sk / Outer Element - 0.170 en la (a) Refer to Table III for nomenclature. (b) m refers to weight of fuel in specified region. (c) v refers to active volume of fuel plate. (d) v refers to the entire volume of fuel plate. (e) v refers to the volume of water. - 23 - Table V Reactivity Effects Due to Changing Experimental Facilities Symmetrical Critical Position Core Condition Reactivity Change (0) 16.311 Standard core - all Be plugs in Standard core - all Be plugs out Standard core , six air filled plugs 16.332 16.328 ~ 4.5 ... 4.0 + + 1.05 Standard core - all bean holes empty Standard core - HB-1 filled Standard core - HB-1, 2, filled Standard core - HB-1, 2, 3, filled Standard core - HB-1, 2, 3, 4, filled 16.310 16.306 16.305 16.302 16.302 + 1.68 + 2.40 + 2.10 Rhoette reading The only result from all these experiments that deviates significantly from what was expected is that associated with the isothermal temperature coefficient. In the HFIR critical experiments the isothermal temperature coefficient was found to be slightly negative over the expected temperature range (68-160°F), whereas that measured in the HFIR facility was slightly positive up to about 120°F, with a maximum increase in reactivity from room temperature of 0.07% Ak/k. This consti- tutes no particular operating problem since the fuel region temperature coeffi- cient is strongly negative. The difference in the two measured isothermal coef- ficients is attributed to a difference in the control rod drive mechanisms, a difference in the water content of the control region, which has a slightly posi- tive coefficient, and a difference in the fuel elements (an 8-kg core with uniform burnable poison distribution was used in the original measurement). In addition to the measurements described above, several sets of flux measure-, ments were made in the flux trap and beryllium irradiation facilities. These measurements were made for two purposes: first, to measure the thermal, resonance and fast flux in the various facilities; and second, to obtain information on the neutron energy and intensity distribution to allow comparison with calculations set up for use by operations personnel and experimenters. The reduction of data from these irradiations has not yet been completed, however, preliminary results extrapolate to a perturbed thermal neutron flux of 2.6 x 1015 at 100 MW at the center of the flux trap with the trap loaded with Al dummy rods. This flux is based on the irradiation of cobalt and gold monitors. A joint Argonne National Laboratory and Oak Ridge National Laboratúry experi- ment on the quality of neutron beams from the HFIR beam tubes was conducted. The * A . - 24 - leakage radiations from the HFIR are not expected to be greatly different than those from the Argonne Advanced Research Reactor (AARR) which is being designed, and therefore information regarding the beam quality is of interest in the AARR program and for use in planning experiments. While the data from these experiments have not been completely analyzed, the following preliminary results are presented due to the interest in the design of beam tubes which are better "tailored" to the needs of the experimenters. Experiments were pe:formed using two of the horizontal beam tubes, HB-2 (radial) and HB-3 (tangential). Neutron-spectrum measurements for both radial and tangential tubes should reveal the extent of the improvement (greater slow-neutron to fast-neutron ratios and fewer gamma rays) in beam quality as the line from beam origin to experiment ceases to intercept the fission source. Ionization chamber measurements indicate a gross gamma ray intensity for HB-2 (radial) about nine times that for HB-3. For the full beam tube (no added coilimation), the extrapolated intensity for HB-2 at 100 MW is about 2 x 105 c/he. Comparable relative intensities were obtained with a scintillation spectrometer; however, the relatively strong Be-capture contribution observed in the HB-3 beam was the most interesting feature. Both foil-activation and 0235 fission counting indicate thermal neutron currents only about 5% higher for HB-2 than for HB-3. For an essentially uncolli- mated beam, the thermal neutron current at a distance of six feet from the shutter exit, on the beam tube axis, extrapolates to about 7 x 109 for 100-Mw operation. Comparison of this with the current observed at a typical ORR (30-Mw) beam tube experiment and making appropriate allowance for the degree of collimation indi- cates a factor of 4-10 advantage for the HFIR (100 MW) in terms of benefit for ex- periments of the types presently planned. This possible advantage does not take into account the improvement in signal to background ratio anticipated for the tangential beam tubes. Measurements taken with a boron-absorption spectrometer indicate roughly equal currents for the two beam tubes in the energy range 1.0 to 100 ev. Proton- recoil spectrometer data shows roughly comparable intensities in the key region with quite strong structures due to Al resonances, presumably due to the several Al diaphragms in the Al beam tube. Measurements made using fission counting techniques show comparable inten- sities for U233, 4235, and Pu239 and factors of 1.51 for Np 237 (0.4 Mev), 1.59 for U236 (0.7 Mev), and 1.56 for U238 (1.3 Mev) for the intensity of HP-2 (radial) over that for HB-3 (tangential). Threshold detectors for the range of 2.9 to 8.1 Mev yielded factors of 5-10. • 25 - Measurements were also made by stopping cach beam in a large water tank and making thermal neutron flux measurements in the forward direction. The re- sults indicate about a factor of 10 difference in gross Mev-neutron intensity. In general the measurements confirm the predicted advantage of the tangen- tial beam tubes over radial beam tubes. It appears that the tangential beam gamma intensity would be soniewha: lower in a D20 reactor where the H- and Be- capture gamma rays should not be present. Also it would be anticipated that fewer neutrons in the Mev range would be present in the beam from a D20 system due to the more nearly elastic scattering of the Be in contrast to that of deuterium. Operation at power levels up to 20 MW was continued during the period from January 29, 1966, through May 30, 1966. Many tests were conducted to obtain operational information and to verify proper behavior during simulated electrical power outages, etc.; however, much of the time was devoted to straightforward operation while awaiting final safeguards approval, from the USAEC, for operation at power levels above 20 Mw. The accumulated power on the first fuel loading (and reactor) was approximately 1300 megawatt-days at the end of this period. - - - - - - - Approach-to-Power Program The approach-to-power phase of the program consists of operation of the reactor for a full fuel cycle at each of several successively higher power levels. As noted previously, the main areas of concern are those influencing fuel plate temperature and influenced by plate temperature, i.e., aluminum-oxide film buildup and radiation damage effects. The first step in power level was selected on the basis that no significant effects should be observed. This step was originally specified to be 50 MW for a fuel cycle; however, due to the long period of operation at 20 Mw it was decided to simply complete the fuel depletion at 50 Mw and then proceed to 75 Mw opera- tion with the second loading. On May 31, 1966, the power level was increased to 50 MW and operation was continued, still using the first fuel loading and still with no experiments in- stalled. Operation continued at 50 Mw until June 19, 1966, at which time the con- trol rods were fully withdrawn and shutdown occurred due to fuel depletion. The total integrated power at shutdown was about 2230 Mwd; this corresponds to an . average U235 burnup of about 30%. This is somewhat in excess of the 1800-1900 Mwd estimated for the clean conditions which were present. Operation during the 50-Mw run was uneventful and no shutdown occurred other than one scheduled test of the safety rods shortly after reaching 50 Mw. No effects were observed which indicated any significant change in fuel plate dimensions due to oxide buildup or radiation damage. . . . . ::.26 . . . . : The reactor was reloaded and following several experiments to check on various operational conditions, the ceactor was taken to 75 MW on June 30, 1966. With the exception of a power reduction to about 30 Mw for a few minutes when one of the main pumps was tripped off the line during a severe thunderstorm, the reac- tor operated steadily at 75 Mw until July 31, 1966, at which time fuel depletion caused shutdown. The accumulated power at shutdown was 2310 Mwd. Evidence of some change in the fuel plates occurred during this run. The fuel element pre- sure drop increased about 5% and the flow decreased about 5% during the cycle. The next operating step in power level is scheduled for 90 MW and if all continues in a satisfactory manner, 100 MW should be reached in September of 1966. 13. SAM TUBE EXPERIMENTS Three of the horizontal beam tubes have been assigned to the Solid State Division and the fourth is assigned to the Chemistry Division. Two of the Solid State Division experiments and the Chemistry Division experiment will consist of triple-axis neutron crystal spectrometers which will be entirely automatic and which will be controlled by computers (PDP-8). One of the Solid State units will be used primarily to obtain added sensitivity in the investigation of certain magnetic properties of materials. The other one will be used primarily for ex- periments involving the inelastic scattering of neutrons 'to obtain information on atomic and molecular energy levels and on the dynamical properties of crystal lat- tices. The third neutron diffraction setup (Chemistry Division) will be used for crystal structure analysis with emphasis on biological molecules. The fourth beam tube experiment (Solid State Division) is currently in the proposal stage and present plans are to install a neutron spectrometer consisting of a neutron chop- . per with associated time-of-flight measuring equipment for use in those investiga- tions of solid state interest that can not be readily investigated with a triple- axis crystal spectrometer. It is anticipated that the three triple-axis spectrometers will be in ser- vice by the end of this year. The time-of-flight spectrometer, if approved, should be in service by the end of 1967. f 14. RADIOISOTOPE PROGRAM In addition to the irradiation of Pu 242 in the flux trap to produce trans- uranium isotopes, it is planned to use the HFIR for the preparation of other high specific activity isotopes. The preparation of these isotopes will provide addi- tional tools for solving medical and physical research problems. Although the vertical irradiation holes in the beryllium may be used, most of the irradiations are planned to be done in a hydraulic rabbit tube facility located in the flux trap on the vertical centerline of the core. The hydraulic system is designed . 27 and is presently being tested out of the reactor. Installation will be made following completion of the approach-to-power program for the reactor. 15. COSTS Table VI lists the approximate capital expenditures incurred in the design and construction of the HFIR. Costs for equipment associated with beam tube ex- periments, etc., are not included in this breakdown. Table VI Design and Construction Costs of HFIR Engineering, design, and inspection Reactor (core, reactor vessel, cooling system, etc.) Instrumentation Building (site preparation, structures, electrical) $ 3,175,000 4,150,000 1,300,000 6,100,000 $14,725,000 The development costs incurred in the HFIR program total about $6.7 x 106, of which about $2.5 x 106 was expended on fuel element and control plate develop- ment. Operational costs totaling about $6.8 x 106 were expended up to July 1, 1966. Of this amount about $3.6 x 106 was expended in preoperational and low power testing, training of the operating staff and miscellaneous technical sup- port. About $2.3 x 106 was spent on fuel element procurement and the remainder for procurement of spare parts and miscellaneous materials. Table VII lists the estimated annual operating cost including the ORNL overhead and the cost of fabricating 24 fuel loadings, but not including the cost of the uranium burned or the cost of reprocessing the depleted fuel elements to recover the unburned uranium. Costs, other than incidental services by operations personnel, associated with experimental programs are not included in this estimate of operating costs, It is estimated that each of the beam tube experimental setups will cost about $215,000 for equipment and installation, with an estimated annual operating cost for each program of about $100,000. The estimated cost of fabricating and installing the central hydraulic rabbit tube and loading station is $50,000. .- . - - * - * > 16. POTENTIAL OF HFIR-TYPE REACTORS FOR HIGHER FLUXES Several preliminary studies have been undertaken to assess the possibility of achieving fluxes significantly higher than 1016 neutrons/cm2 sec in the flux trap of a reactor similar to the HFIR. One study (unpublished) by R. D. Cheverton of the HFIR Project indicates that a five-fold increase in the power density to · S . 2 + - 29 .. 3 Table VII Estimated Annual Operating Expense of HFIR“ Reactor Operating Personnel 425,000 475,000 General Operating Costs (Utilities, Liquid Waste Handling, Health Physics, etc.) 380,000 Instrumentation and Controls Division Craft and Engineering Support Mechanical Craft and Engineering Support 450,000 Analyses and Technical Support 140,000 Miscellaneous Material Costs 120,000 Spare Components 100,000 Control Plate Fabrication 210,000 Fuel Element Procurement 1,500,000 Total $3,800,000 - - Includes ORNL overhead, etc. "The unit cost used here is $62,000 per loading. This is an optimistic figure as present experience indicates a cost of about $95,000 per loading for the first 33 loadings and about $75,000 for the next 24 loadings. ~10 MW per liter is probably ultimately possible. With a modest increase in core size and the use of D20 as a coolant, it appears that an unperturbed thermal neu- tron flux in the island of about 3 x 1016 would be achieved at a power level of about 750 Mw in this "ultimate" HFIR-type reactor. In order to achieve this sort of performance significant improvements would have to be made in fuel element technology; however, the necessary improvements do not appear to require "inven- tions" but extensions of techniques currently under inves:igation. It is highly questionable whether such a system could be considered feasible at this time from the standpoint of fuel cycle and pumping power considerations. However, the study certainly serves to place a reasonable upper limit on the probabie maximum per- formance of this type of reactor at the present time and probably for some time to come. In addition to consideration of this "ultimate" HFIR-type reactor, Mr. Cheverton also considered the question of whether or not one could build an HFIR- type 'reactor to produce an unperturbed thermal neutron flux of 1 x 1016, using only a modest extension of current technology. It was concluded that such a system could be built and that only a modest development effort would be required. . - 29. 17. ACKNOWLEDGMENT It is manifestly impractical to give recognition in this paper to all of the ORNL personnel who have participated in the development, design, construction, and initial operation of the HFIR facility; however, the following are noteworthy among those who have made major contributions: project directors, C. E. Winters (1960-1961), A. L. Boch (1961 to present); physics and general reactor engineering, R. D. Cheverton; heat transfer and hydraulics, N. Hilvety, H. A. McLain, and T. G. Chapman; shielding and general reactor engineering, H. C. Claiborne; reactor design engineering, J. R. McWherter, J. H. Westsik, R. E. Hoskins, W. G. Cobb, and R. E. Schappel; fuel element and control plate development and fabrication, G. M. Adamson and C. F. Leitten; reactor control and instrumentation, L. C. Oakes; construction coordination, J. W. Hill, Jr. and H. Grimac; reactor component fabrication coordi- nator, R. M. Hill, Jr.; reactor operations supervisor, R. V. McCord. F • 30 - IHIM References G. T. Seaborg, "Progress Beyond Plutonium," Chemical and Engineering News 44(25): 76-88 (June 20, 1966). J. A. Swartout et al., "The Oak Ridge High Flux Isotope Reactor," Proceedings of the Third International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1964, Vol. 7, 360-371, United Nations, New York, 1965. F. T. Binford and E. N. Cramer (eds.), "Te High Flux Isotope Reactor: A Functional Description," ORNL-3572 (May 1964). J. M. Beeston and R. L. Tromp, Phillips Petroleum Company, "Irradiation Ef- fects on Beryllium," 16th High Temperature Fuels Committee Meeting, Brook- haven National Laboratory, May 15-17, 1963. J. W. Dykes and J. D. Ford, "A Preliminary Report of Beryllium Damage Observed in the MTR Reflector," IDO-16899 (June 20, 1963). V. A. Walker, M. J. Graber, and G. W. Gibson, "ATR Fuel Materials Development Irradiation Results--Part II," IDO-17157 (June 1966). F. E. Jablonski and A. F. DiMeglio, "Naval Research Laboratory Reactor, Part IX--Measurement of Fast-Neutron Beam Currents," NRL Report 5213 (1958). A. J. Court and K. Downes, "Beam Tube Flux Evaluation on the High Flux Ream Reactor Critical Experiments," Trans. Am. Nucl. Soc., 3(1): 103-104 (June 1960 Annual Meeting). . H. C. Claiborne and G. Rakavy, "A Transport Calculation of the HFIR Beam Hole Currents," ORNL-CF-60-12-18 (Dec. 5. 1960). (107 J. Replogle, Modric: A One-Dimensional Neutron Diffusion Code for the IBM-7090," AEC Research and Development Report K-1520 (September 6, 1962). [117 M. L. Tobias and T. B. Fowler, "The Twenty Grand Program for the Numerical Solution of Few-Group Neutron Diffusion Equations in Two Dimensions," ORNL-3200 (Feb. 7, 1962). : (127 H. C. Honeck, "Thermos: A Thermalization Transport Theory Code for Reactor Lattice Calculations," BNL-5826 (September 1961). 137 G. D. Joanou and J. S. Dudek, "GAM-1: Â Consistent P, Multigroup Code for the Calculation of Fast Neutron Spectra and Multigroup Constants," GA-1850 · (June 28, 1961). • 31 - List of Figures Fig. 1. High Flux Isotope Reactor (Vertical Section) Fig. 2. HFIR Fuel Element Fig. 3. Schematic Representation of Core Cross Section, Showing Fuel Contours Fig. 4. Schematic Representation of Control Rod Arrangement in HFIR; Rod Posi- tion Shown Approximate for Clean, Critical Core Fig. 5. HFIR Target Assembly Fig. 6. Horizontal Section Through Pressure Vessel Fig. 8. Core Schematic and Flux Distribution Fig. 9. HFIR Differential Rod Worth Data . - 32 - MOB AN UWV ". 20 , ON WWW. -" vel" I . A W . .. . .. n . .. i Withing . . . . . way YA NA W... ... . W ". " Nu-WWWW END DATE FILMED 10/ 26 /66 "W . 1,Q, . ! . 2 hy < - I. - - - NIM N ' NK