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- Savannah River Project Site Report: 30,000 KW Prototype Partically Enriched Uranium Gas Cooled, Graphite Moderated Nuclear Power Plant for United States Atomic Energy Commission Idaho Operations Office
- Report describing a modified prototype of a nuclear reactor that uses partially uranium-enriched fuel and is cooled by helium. The construction site, site safety aspects, and design and construction costs are included.
- The Theory and Operation of Shock-Load Ferroelectric Transducers
- A report discussing "an explosive ferroelectric power supply and the experimental approach followed in determining its operating characteristics."
- Van Slyke Factors for Hydrogen, Oxygen, Carbon Dioxide, and Carbon Monoxide
- Tables of data calculated on an IBM 610 automatic computer are given to convert the measured pressure differences in the Van Slyke gas analysis apparatus into micromolar concentration of gas in solution for hydrogen, oxygen, carbon dioxide and carbon monoxide in water. Measured solubility coefficients for carbon dioxide in various aqueous solutions of formic acid are given together with the conversion factors for use with the Van Slyke analysis apparatus. (auth)
- Extraction of Uranium, Magnesium, Zirconium, and Cerium From Bismuth With a Fused Fluoride Salt Mixture
- The extraction of uranium, magnesium, cerium, zirconium, and niobium from bismuth with a molten mixture of sodium fluoride and zirconium fluoride was demonstrated. Comparative rates of extraction were obtained. The effects of high concentrations of magnesium and of hydrogen fluoride sparging on the extraction process were investigated. Tracer studies demonstrated that exchange occurs between zirconium dissolved in the bismuth and zirconium in the fused salt. The applicability of the fused fluoride extraction step to the processing of the Liquid Metal Fuel Reactor'' solution fuel is discussed. (auth)
- Technical basis for establishing process tube pressure limits for KER loops 2 and 3 and for the NPR Prototype Facility
- In compliance with a request from Coolant Testing Operation, the Reactor Engineering Operation has made a study to determine the maximum operating pressure limits for the pertinent Zircaloy-2 process tubes. Since these tubes shall be used for testing NPR fuel elements, it is considered desirable that KER Loops 2 and 3 permit operation at temperatures of around 300{degrees}C while the NPR prototype facility permit operation at about 316{degrees}C in a manner such that there is minimum hazard to the KE-Reactor and to personnel.
- Estimate of Hazard Produced by Accidental Release of Gaseous Fission Products from an ORR Fused Salt Capsule Experiment
- An accidental release of gaseous fission products from an ORR fused salt capsule, containing 26 mg. of U/sup 235/, was postulated and the resuiting hazard estimated by calculating the maximum external and internal dose an individual could receive from exposure to the gaseous fission products and their decay products. Assuming all the contained gaseous fission produets are released, the resulting external and internal dosc, to organs other than the thyroid, arc insignificant. The dose to the thyroid by radioiodine is considered to be significant. By retaining at least 90% of the iodine isotopes in the experiment system through use of an iodine trap, a large reduction in both the external whole body and internal thyroid doses may be achieved. Therefore, assuming an iodine trap is utilized, it appears that the consequences of am accidental gaseous fission product release from an ORR fused salt capsule experiment would not be serious. (auth)
- Diffusion of Uranium with Various Transitional Metals; DIFFUSION DE L'URANIUM AVEC QUELQUES METAUX DE TRANSITION
- The diffusion process in uranium and its alloys was studied from 550 to 1075 deg C with diffusion couples of U with Zr, Mo, Ti, and Nb and with the alloys U--Nb and U--Mo. A brief description is given of the experimental methods. Results relative to the concentration-penetration curves are presented, and the coefficients of diffusion are calculated. The equilibrium diagram was established for the U--Zr system. The results obtained by micrographic examination, microhardness measurements, and autoradiography are compared with each other. The mechanisms of diffusion are investigated by studying the Kirkendall effect and calculating the Darken intrinsic coeffi cients in the U--Zr and U--UMo diffusion couples. (J.S.R.)
- An Evaluation of Mercury Cooled Breeder Reactors
- Abstract: The technical feasibility and economic potential of fast breeder power reactor systems cooled with boiling mercury have been investigated by American-Standard under the United States Atomic Energy Commission's New Reactor Concepts Evaluation Program.
- Army Gas-Cooled Reactor Systems Program Monthly Progress Report: April 1959
- Abstract: This monthly progress report covers the activities of the Army Gas-Cooled Reactor System Program for April 1959. The program includes a water-moderated heterogeneous reactor (Gas-Cooled Reactor Experiment I), a graphite-moderated homogeneous reactor (Gas-Cooled Reactor Experiment II), a mobile gas-cooled reactor (ML-1), and the coordination of the Gas Turbine Test Facility. [It reports] the progress of each project, the associated tests and data evaluation, the applicable design criteria, and the fabrication of reactor components" (p. 1).
- Army Gas-Cooled Reactor Systems Program Semiannual Progress Report: January 1 - June 30, 1959
- Report documenting the progress of the Army Gas-Cooled Reactor Systems Program to develop a mobile, low-power, nuclear power plant for Military field operation.
- Request for irradiation, request for examination and evaluation of new, improved fuel elements in ETR facilities. Extension of GEH-10
- No Description Available.
- Thermal Performance of UO$sub 2$ in Existing and Planned Reactors
- BS>The thermal characteristics of various reactors fueled with uranium dioxide fuel pellets are listed. (W.L.H.)
- THE NUCLEAR FUEL CYCLE: PROSPECTS FOR REDUCING ITS COST
- Nuclear fuel cost of 1.25 mills/kwh would make nuclear power competitive with conventional power in lowcost coal areas if capital and operating costs can be brought to within about 10 percent of those of coal-fired plants. Substantial decreases in fuel fabrication cost are anticipated by 1970: other costs in the fuel cycle are expccted to remain about the same as at present. Unit costs and irradiation levels that would be needed to give a fuel cost of 1.25 mills/kwh are believed to be attainable by 1970. (auth)
- THE DIFFUSION OF HYDROGEN IN BETA ZIRCONIUM
- Diffusion coefficients for hydrogen in beta zirconium were determined from permeation rates in the range 650 to 850 deg C. Both the steady-state method, which is dependent upon the hydrogen concentration, and the time-lag method, which is independent of hydrogen concentration, were employed to obtain diffusion data. Zirconium disks, 0.03 to 0.1 cm thick and varying in hydrogen concentration from 9 to 33 at.%, were used to measure permeation rates. The diffusion coefficients determined by the steady-state and time-lag methods on samples of differing thickness were in agreement. It was concluded that the permeation process was diffusion controlled. The diffusion coefficients were found to be independent of concentration and can be expressed by D = 6.14 x 10/ sup 4/ exp (--45,900/RT). (auth)
- REACTION OF NITROGEN WITH NIOBIUM
- Reaction rates of niobium with nitrogen were determined gravimetrically from 675. to 875 deg C with a recording microbalance and volumetrically from 1100 to 1600 deg C with a modified Sieverts apparatus. Diffusion coefficients and terminal solubilities were determined from 800 to 1600 deg C by the concentration- gradient technique. Tne reaction of nitrogen with niobium follows a parabolic rate law at 675 to 1600 deg C. The expression for the diffusion coefficient for nitrogen in niobium at 800 to 1600 deg C is given as well as the expression for the terminal solubility for nitrogen in niobium. (auth)
- REACTIONS IN THE NIOBIUM-HYDROGEN SYSTEM
- Equilibria in the niobium- hydrogen system were determined in the range 100 to 900- deg , 0.1 to 1000 mm of mercury hydrogen pressure, and hydrogen/ niobium atomic ratios of 0.01 to 0.85. X-ray measurements were obtainpd at 25 to 400 deg C at hydrogen/niobium ratios up to 0.54. The studies showed thnt a solid solution of hydrogen in niobium is produced throughout most of the system. A miscibility gap was found at low temperatures and pressures, with a critical point at about a temperature of 140 deg C, a hydrogen pressure of 0.01 mm of mercury, and a hydrogen/niobium ratio of 0.3. Sorption rates at 300 to 550 deg C wore initially linear. At higher temperatures, sorption rates were controlled by diffusion in the metal matrix. Diffusion coefficients at 600 to 700 deg C can be expressed by D = 0.0215 exp STA(-9370 plus or minus 600)/RT!. Desorption rates were lower than those predicted by diffusion. (auth)
- Design Analysis of a Prepackaged Nuclear Power Plant for an Ice Cap Location
- Report describing a the design requirements of a proposed nuclear power plant for use on an ice cap.
- Intermediate Heat Exchanger Preliminary Design. Vol. 1, IHX Preliminary Design
- Preface: The intermediate heat exchanger is designed for operation in a nuclear power plant using liquid sodium as the primary and secondary coolant. Since the primary fluid coming from the reactor is radioactive, the purpose of the IHX is to transfer heat to a nonradioactive fluid which then goes to a steam generator. Because of this activity the until will be enclosed in a concrete pit and will not be accessible during periods of operation. Immediately after shut down it will be necessary to allow time for radioactive decay before the unit will be accessible to personnel. Because of inaccessibility and possible long periods allowed for decay time, it is imperative that the unit give trouble free operation. During periods of shut down, the internals should have easy access for inspection and repair if necessary so that down time is held to a minimum. The general arrangement of the heat exchanger described in this report presents a conventional design utilizing known materials and existing methods of fabrication. In further consideration of all concepts, designs and analyses developed during this period of the program, it is felt that this preliminary design will provide an intermediate sodium heat exchanger of lower cost and more reliable operation.
- Brief Review of Heat Transfer Problems Encountered in the Production of Magnetic Fields
- The design of internally cooled electrical coils for the production of high intensity magnetic fields presents many new aspects and combinations of the familiar modes of heat transfer. However, the customary methodology appears to be sufficient for preliminary analysis and understanding of those problems. This methodology comprises the derivation of a qualitative, approximate equation expressing the relative performance of the various parts of a system, followed by an examination of this equation in order to locate the limiting features of the system. These features are then investigated by more powerful methods, which in turn provide guidance for development research in the laboratory. (auth)
- Brief Review of Heat Transfer Problems Encountered in the Production of Magnetic Fields
- The design of internally cooled electrical coils for the production of high frequency intensity magnetic fields presents many new aspects and combinations of the familiar modes of heat transfer. However, the customary methodology appears to be sufficient for preliminary analysis and understanding of those problems. This methodology comprises the derivation of a qualitative, approximate equation expressing the relative performance of the various parts of a system, followed by an examination of this equation in order to locate the limiting features of the system. These features are then investigated by more powerful methods, which in turn provide guidance for development research in the laboratory.
- TABLES FOR A SEMI-INFINITE CIRCULAR CURRENT SHEET
- The concept of an ideal current sheet is frequently useful in the design of electromagnets. Since a current sheet of any length can be represented by the superposition of two semi-infinite sheets, it is desirable to have tables for the magnetic field produced by a semiinfinite current sheet. The tables presented include both the magnetic field intensity and the magnetic vector potential. To use the tables it is not necessary to apply the symmetric properties of the fields. This doubles the size of the tables but greatly simplifies their use. (auth)
- Aerodynamic characteristics of the X-15/B-52 combination
- Report presenting an investigation to determine the carry loads and mutual aerodynamic interference effects from high-speed wind-tunnel tests and the drop characteristics of the X-15 through the B-52 flow field from low-speed dynamic-model drop tests and six-degree-of-freedom calculations. The X-15 installation was found to increase drag at cruise conditions by approximately 15 percent.
- Steady-State Recirculated Reactor Stability and Operational Characteristics - Water and Metal Temperature Coefficients
- It is desirable that a reactor exhibit a self-regulating effect. If this were not true any disturbance to the reactor would result in a continual increase in the magnitude of the disturbance and the reactor would be unstable. In this investigation the reactor is considered to have two reactivity feed-backs: metal temperature and water temperature reactivity effects. These two variables through a metal temperature coefficient and water temperature coefficient determine not only the reactor stability but also determine many operational characteristics.
- The Thermal Expansion of Synthetic Graphites at Temperature Intervals Between 80 and 2000f
- The mean linear and cubical coefficients of thermal expansion of eight commercial samples of graphite were determined for temperature intervals between 80 and 2000 deg F. The linear thermal expansion was measured with an automatic recording dilatometer using a rod-shaped specimen 2 in. long and 1/4 in. across. The specimen was heated in an atmosphere of helium. The results were in good agreement with those of Currie, Hamister, and MacPherson. The mean linear coefficient was found to increase with temperature. For the samples studied, the mean linear coefficients from 80 to 2000 deg F were 1.50 to 2.34 x 10/sup -6// deg F parallel and 2.26 to 3.45 x 10/sup -6// deg F perpendicular to the grain and were found to vary linearly with the electrical resistivity measured at 32 deg F. (auth)
- Hot-Pressure Bonding of OMR Fuel Plates
- Abstract: An alluminum-clad low-enrichment, uranium-alloy fuel element of flat plate configuration has been proposed for the Organic Moderated Reactor (OMR).
- Possible Test Sites in Granitic Rocks in the United States
- Introduction: This report describes areas of granitic rocks suitable for underground nuclear tests within Federally-controlled land in the continental limits of the United States. This information was requested of the U. S. Geological Survey by the Albuquerque Operations Office of the U. S. Atomic Energy Commission, and was compiled during March 1959 by D. C. Alvord, W. J. Carr, P. M. Hanshaw, S. P. Kanizay, C. S. Robinson, R. W. Schnabel, J. A. Sharps, and C. T. Wrucke.
- A Collection of Excerpts and a Bibliography relative to the Proposition "That Congress should be given the Power to Reverse Decisions of the Supreme Court"
- This report is a Collection of Excerpts and a Bibliography relative to the Proposition "That Congress should be given the Power to Reverse Decisions of the Supreme Court".
- Physical Properties of Neutralized Zirflex Waste
- Zirflex cladding waste is to be neutralized to pH 10 before transfer to waste storage tanks. This treatment causes the precipitation of zirconium oxide or hydroxide, which may lead to flow difficulties during transfer. The purpose of this investigation was to determine the physical properties and flow characteristics of the neutralized slurry to assist in the selectin of satisfactory transfer equipment and storage conditions.
- NON-PRODUCTION FUELS REPROCESSING, CENTRIFUGATION STUDIES ON VARIOUS DISSOLVER EFFLUENT SOLUTIONS
- >The proposed flowsheets for reprocessing of nonproduction fuels include centrifugal separation of particulate matter from various dissolver effluent solutions. The settling characteristics of process solids were determined in water and in cold process solutions. Uranium dioxide particles will be recovered from Zirflex and Sulfex cladding waste solutions, and core-dissolver solutions will be centrifuged for removal of ZrO/sub 2/, metallic slimes, siliceous matter, and uranium-bearing solids. (W.L.H.)
- PHYSICAL PROPERTIES OF NEUTRALIZED ZIRFLEX WASTE
- An investigation was made to determine the physical properties and flow characteristics of the neutralized slurry to assist in the selection of satisfactory transfer equipment and storage conditions. The neutralized Zirflex waste slurry contalns 20 vol.% rapidly settling solids. It can be transferred easily if the flow is in the turbulent condition, but agitation is needed during temporary storage. Pipe lines should be flushed with water after transfer of the waste slurry. (W.L.H.)
- Fuels Preparation Department Analytical Laboratory Manual
- The purpose of the Analytical Laboratory Manual is to assemble the basic procedures to be used for the analyses of materials employed within the Fuels Preparation Department. The methods appear in detailed steps suitable for laboratory use. This document replaces the "Essential Material Analytical Manual, " HW-25375 and "Metal Preparation Analytical Manual," HW-30862.
- FLIP--AN IBM-704 CODE TO SOLVE THE PL AND DOUBLE-PL EQUATIONS IN SLAB GEOMETRY
- A method of obtaining the few-group form of the P/sub L/ and double-P/ sub L/ equations is given for slab geometry. Anisotropic scattering is allowed within specified limitations. The difference equations and associated recursion relations are discussed; the features and restrictions of FLIP are explained; and a detailed discussion of the input and output is presented. Operating intructions are given, and a sample problem is included. (auth)
- CALCULATION OF THERMAL NEUTRON FLUXES IN PRIMARY SHIELDS
- A method is presented for calculating thermal neutron fluxes in the primary shields of reactor systems which eliminates reliance on mock-up experimental data. A multigroup P/sub 1/ approach is ernployed with the spatial dependence of the neutron sttenuation adjusted through use of a point source attenuation kernel for a homogeneous hydrogenous medium. Comparison of calculation with experiment is presentad. (auth)
- LABORATORY DEVELOPMENT OF A PROCESS FOR SEPARATING BARIUM-140 FROM MTR FUEL
- S>The results of all laboratory research and development on the process for separation of barium-140 from MTR fuel elements are presented. The steps include caustic dissolution separation of barium and strontium with fuming nitric acid and removal of strontium by the chromate-acetate method. The results of laboratory and pilot plant corrosion investigations and high radiation level flowsheet tests in the Multicurie Cell are also included. ( auth)
- GAMMA-RAY AND FAST NEUTRON ANNULAR STREAMING EVALUATION THROUGH SODIUM REACTOR EXPERIMENT (SRE)-MARK II CONTROL AND SAFETY ROD ASSEMBLIES
- An experimental program was initiated io determine the extent of fast neutron and gamma ray streaming through the SRL Mark II control and safety rods and to evaluate the adequacy of the shielding provided in these control and safety rods. The methods and procedures used to evaluate these problems are routine and proven for determining gamma-ray and fast neutron dosages using radiation sensitive films and gold foils. The final experimental results indicated that no excessive streaming of either gamma rays or fast neutrons is present above or around the SHE Mark II control and safety rods. The analytical attenuation methods used to calculate the fast neutron and gamma-ray streaming dose rates gave results that compared favorably with the experimental data. Even ihough the agreement was favorable, it cannot be concluded that these analyical methods would be equally valid for other annular geometries. Additional experimental work will be necessary in order to establish the validity for performing similar analysis, but the favorable agreement encourages the use of such methods until other methods are determined. (auth)
- Sodium Reactor Experiment (Sre) Shielding Evaluation for Thermal Neutron Streaming at Reactor Vessel Coolant Pipe Penetrations
- The experimental program performed in the SRE auxiliary and main primary galleries was part of a program to determine the adequacy of the shielding configuration for the SRE. The work discussed in this report is concerned with analysis of neutron streaming at coolant pipe penetrations of the reactor vessel, analysis of the shielding required, testing and evaluation of recommended shielding, and measurement and correlation of neutron streaming in labyrinths with theory. The activation analysis method using zinc sheets which was developed for the program of determining thermal neutron streaming in the SRE primary galleries was proven to be versatile, accurate, and reliable. A modified form of the theoretical method of Price, Horton, and Spinney, used to determine neutron scattring through labyrinths, was found to agree favorably with the experimental results obtained from the SRE primary galleries. The theoretical attenuation method used no determine the neutron shield configuration installed in the auxiliary primary gallery was found to give an overestimate of the actual attenuation properties of this shield. The neutron shield configuration installed in the auxiliary primary gallery proved to be adequate in reducing the thermal neutron streaming flux to an acceptable level. It is concluded that both SRE primary galleries are now adequately shielded to prevent excessive neutron- induced activation of the components and equipment located therein. (auth)
- Proposed Helium Purification System for the Experimental Gas-Cooled Reactor (EGCR)
- Liquid and dry processes suituble for the purification of gases by the removal of CO/sub 2/, H/sub 2/O, CO, H/sub 2/, and hydrocarbons are discussed. Recommendations are given for specific processes io be included in a "dry" (no liquid absorbents or chemicals used) purification system for the hellum coolant of the EGCR The recommended processes include (1) a catalytic converter for the oxidation of CO, H/sub 2/, and hydrocarbons to CO/sub 2/ and H/sub 2/O, (2) cooler-condensors for the removal of the bulk of the R/sub 2/O, (3) silica gel adsorbers to complete the removal of H/sub 2/O, and (4) Linde Molecular Sieve adsorbers for the removal of CO/sub 2/. No provisions are included for the planned removal of radioactive gases or particulates. (auth)
- USE OF THE "ACTION INTEGRAL" IN EW STUDIES
- No Description Available.
- Comments on Equipment for a PRTR Water Quality Control Laboratory
- This document describes required laboratory space and lists major equipment items necessary for a routine water quality laboratory in the P. R. T. R. Building. During discussions with R. D. Widrig and V. L. Rooney about the analytical sample program for the Plutonium Recycle Test Reactor, the author was asked to summarize equipment and space needs for a water control laboratory to provide routine analytical coverage on some of the water systems. Based upon 1706-KE-KER experience, some operating personnel may be used to provide analytical coverage on those routine analyses that are needed on around-the-clock basis with a savings of both time and money.
- The Injection Casting of Plutonium
- Plutonium metal can be injected into cold metal molds to form castings with thin walls. The operation is performed in a vacuum chamber, using an inert gas as the injecting medium. Sound pieces free of gas cavities can be made using either pure or delta-stabilized plutonium. This report describes the equipment and techniques used to cast a typical thin section piece in the form of a 6 in. diameter, 45' cone.
- Research in Photosynthesis
- The determination of the specific radioactivities of the chlorophylls and carotenoids of algae after photosynthesis with C/sup 14/O/sub 2/ is a formidable task, due to the extreme lability of these compounds. The whole success of the method depends on adequate chromatographic separations of the pigments from the colorless contaminants which are closely associated with them, and yet the time involved for such separations is sufficient for marked decomposition of the pigments to occur. It is suggested that the techniques of column chromatography, followed by centrifugallyaccelerated paper chromatography of the spectroscopically pure pigments, may resolve this problem. In an investigation of the action of cyanide on photosynthesis, green algae have been treated with radioactive cyanide. A multitude of products have been found to be formed in very short exposure times. One of these was identified with a material formed when algae are given radioactive CO/sub 2/ and nonradioactive KCN. This material has been identified as the cyanide addition product of ribulose-1,5- diphosphate. Upon hydrolysis it gives a branched-chain sugar acid (or mixture of isomers) closely related to hamamelonic acid. Perhaps the most important aspect of this work is the demonstration of the chemlcal role of cyanide. (auth)
- Physical Chemistry of the Fischer-Tropsch Synthesis
- From Summary: "This paper summarizes the results of physicochemical studies of the Fischer-Tropsch synthesis (the catalytic hydrogenation of carbon monoxide) undertaken by the Federal Bureau of Mines as part of its program on improving processes for producing liquid fuels from coal."
- Aging of Al-Li Alloys - Part I
- Technical report outlining experiments on aluminum-lithium alloys. From Abstract: "Aluminum-lithium alloys are subject to precipitation from solid solution, and may be age hardened by the same techniques used for more common aluminum alloys. Spherical particles of precipitate were observed with the electron microscope in 1.5% and 2.8% Al-Li alloys after aging for times comparable to those required to produce maximum hardness. Rod-shaped particles that were oriented parallel to either the (110) or the (111) planes of the aluminum matrix were observed in overaged specimens."
- Grain Refinement of Uranium by a Beta-Quench, Alpha-Anneal Process
- No Description Available.
- NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT
- Neutron-flux measurements in a concentric-cylinder fuel element were made in a gas-cooled in-pile loop operated adjacent to the core of the BRR. The fuel element comprised four concentric fuel cylinders. Each fuel annulus (outside diameters- 1.248 1.018, 0.810 and 0.590 in.) consisted of a 0.031-in.- thick core of UO/sub 2/ dispersed in type 347 stainless steel and clad on each side with 0.007 in. of typee 318 stainless steel. The element was 24 in. long and the total uranium-235 content was approximately 192 g. Radial, vertical, and peripheral flux distributions were studied. The vertical flux profile was cosine- shaped with a peak-to-average ratio of 1.26. The peripheral variation around the loop wall could also be fitted to a cosine curve (with a peak-to-average ratio of 1.10). The average radial flux depression from the outer fuel cylinder to the center of the element was a factor of 2.14. Power generation in the element calculated from flux measurements agreed to within 10% with the power generated by measuring gas now rate and temperarure rise across the fuel element. The ratio of peak-to-average power density was found to be 1.75. (auth)
- Neutron-Flux Measurements in a Concentric-Cylinder Fuel Element
- The following report presents neutron-flux measurements made with a concentric-cylinder element (Mark II) and includes axial, radial, and peripheral flux distributions.
- CUREBO: A GENERALIZED TWO-SPACE-DIMENSIONAL CODING WITH CROSS-SECTION AND DEPLETION CALCULATIONS FOR THE IBM 704
- The CUREBO code for the IBM 704 is described. The code is divided into three parts including the calculation of nuclear cross section of the various physical components of a reactor (WOX7), the solution of the multigroup diffusion equations in two-space dimensions in order to find neutron fluxes and sources for an operating reactor containing these components ( CURE), and the calculation of fuel and poison depletion as a result of operating this reactor under steady- state conditions (BO2). (auth)
- Operating Manual for the Argonaut Reactor
- The design of the Argonaut (Argonne Nuclear Assembly for University Training) was initiated by the Reactor Engineering Division of Argonne National Laboratory to satisfy needs for a low-power reactor facility within the Laboratory, and for training uses within the international School of Nuclear Science and Engineering (ISNSE). It was intended primarily for instruction and research in reactor physics. It was also considered as a possibility that it would fulfill the requirements of universities engaged in a program of nuclear science. The cost of the facility was to be kept to a minimum consistent with the high degree of inherent safety and a great amount of flexibility in the system. The basic design stemmed from the Knolls Atomic Power Laboratory Thermal Test Reactor* (TTR), now called Nuclear Test Reactor (NTR). Modification during the course of the work justified the new name "Argonaut".
- Particle Accelerator Division Summary Report: April 15, 1958 through October, 1958
- Report issued by the Argonne National Laboratory discussing a summary report of work completed between April and October, 1958. Summaries of the studies conducted and work completed are presented. This report includes tables, and illustrations.
- First Sodium Reactor Experiment (SRE) Test of Hallam Nuclear Power Facility (HNPF) Control Materials
- An experiment was conducted in the SRE to measure temperatures and neutron flux levels in and near a boron-containing simulated control rod. The data are being used to check analytical methods developed for prediction of control rod heat generation rates and maximum temperatures in this type of control rod in the Hallam Nuclear Power Facility. The maximum observed temperatures with a reactor power level of 20 Mw were 1363 deg F for a boron-- nickel alloy ring having a 0.105-in. radial clearance with the thimble and 1100 deg F for a boron -nickel alloy ring having a 0.020-in. radial clearance. The maximum temperature difference between the coolant and the control rod was 473 deg F. It is concluded that the expected greater heat generation rates in the Hallam reactor would prohibit the use of boron-containing absorber materials in a combined a him-safety rod. (auth)