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Status Report on the Argonne Advanced Research Reactor
The interim design and development status is reported. The scope of the work was limited to conceptual design studies supported by critical experiments plus heat transfer and hydraulic tests. Design criteria, facility and site, reactor, core geometry and composition, fuel elements, reflector, core and reflector support structure, reactor vessel, control and instruments, primary coolant systems, secondary coolant system, auxiliary systems, experimental facilities, building layout and construction, plant ventilation, heating and air conditioning, critical experiments, reactor physics, heat transfer studies, and shieldings are discussed. (M.C.G.)
Date:
November 1, 1961
Creator:
Lennox, D. H.; Barts, E. W.; Batch, R. V.; Beyer, F. C.; Jorgensen, G. L.; Kelber, C. N. et al.
Object Type:
Report
System:
The UNT Digital Library
Cost Function Studies for Power Reactors
A function to evaluate the cost of electricity produced by a nuclear power reactor was developed. The basic equation, revenue = capital charges + profit + operating expenses, was expanded in terms of various cost parameters to enable analysis of multiregion nuclear reactors with uranium and/or plutonium for fuel. A corresponding IBM 704 computer program, which will compute either the price of electricity or the value of plutonium, is presented in detail. (auth)
Date:
November 1, 1961
Creator:
Heestand, J. & Wos, L.T.
Object Type:
Report
System:
The UNT Digital Library
New Solutions of the Boltzmann Equation for Monoenergetic Neutron Transport in Spherical Geometry
Solutions of the Boltzmann equation for monoenergetic neutron transport in spherical geometry are derived which are respectively singular and regular at the center of the sphere. A few specific partial singular solutions are presented. The regular solutions in spherical geometry are constructed by superposition of solutions in plane geometry which belong to the same k. Finally, the solutions are compared with their representations by a series of spherical harmonics. (D. L.C.)
Date:
November 17, 1961
Creator:
Kofink, W.
Object Type:
Report
System:
The UNT Digital Library
PROCESSES FOR RECOVERY OF URANIUM AND THORIUM FROM GRAPHITE-BASE FUEL ELEMENTS. PART II
Laboratory-scale tests on methods for recovering uranium and thorium from graphite-base reactor fuel elements are reported. The 90% HNO/sub 3/ process, which involves simultaneous disintegration and leaching in 21 M HNO/sub 3/, is applicable to all fuel elenments which do not contain coated fuel particles. Leaching of irradiated (0.001% burnup) fuels containing 3 and 12% uranlum recovered approximates 99.3 and 99.9%, respectively, of the uranium in two 4-hr leaches with boiling acid. The graphite residue retained > 50% of the long-lived fission products. Three successive leaches of fuel containing uranium and thorium recovered approximates 99% of both elements. Uranium recoveries by combustion in oxygen followed by dissolution of the ash hn nitric acid or fluorlde-catalyzed nitric acid are quantitative only when the fuel is not coated, does not contain Al/sub 2/O/sub 3/-coated fuel particles, and is free from impurities such as iron. During combustion up to 95% of the Ru-106 was volatilized from irradiated specimens. Recoveries, by leaching with 70% HNO/sub 3/, from fuel specimens containing Al/sub 2/O/sub 3/-coated fuel particles were greater than 99% when the specimens were ground finer than 200 mesh to ensure crushing of the fuel particles. (auth)
Date:
November 30, 1961
Creator:
Ferris, L.M.; Kibbey, A.H. & Bradley, M.J.
Object Type:
Report
System:
The UNT Digital Library
Evaluation of Wire Scanner for SM-1
Preliminary design concepts are presented for a wire scanner for experimentally evaluating spatial variations of neutron flux in the SM-l reactor core. Results of a literature search and determination of optimum criteria for flux mapping the core in minimum time dictated requirements for design concepts and specifications. The utility of both manually instrumented and automatically instrumented wire scanners was analyzed with respect to rapidity of measurement, selectivity of detector location, cost, value of data, plant downtime, and additional factors. (auth)
Date:
November 22, 1961
Creator:
Kemp, S. N.
Object Type:
Report
System:
The UNT Digital Library
CONCEPTUAL DESIGN AND ECONOMIC EVALUATION OF A STEAM-COOLED FAST BREEDER REACTOR
A conceptual design and economic evaluation of 300 and 40 MW/.sub e/ steam-cooled fast breeder reactor power plants were performed. A reactor core composed of U-Pu oxide rod-type fuel elements clad with Inconel-X and surrounded by a blanket of depleted UO/sub 2/ fuel was studied in some detail. Reactor breeding ratios of from 1.27 to 1.5 and overall system doubling times of from 20 to 30 years are achievable. For the near term (1967) 300 MW/sub e/ plant, an energy cost of 7.6 mills/kwh is estimated, based on AEC ground rules for privately financed plants and utilities. This cost may go down to 5.7 mills/kwh by 1975. For the 40 MW/sub e/ plant corresponding energy costs are 19.5 and 13.7 mills/kwh, r -spectively. The R&D program required for this reactor concept is estimated at million with an additional million for improvements leading to the 1975 reactor. Investigation of the operational and safety aspects of the reactor indicated that satisfactory procedures can be used for startup, shutdown, and emergency cooling of the reactor. An increase in reactivity upon flooding can be prevented by incorprating small amounts of high resonance absorption material in the core. Preliminary calculations indicate a substantial increase in …
Date:
November 15, 1961
Creator:
Sofer, G.; Hankel, R.; Goldstein, L. & Birman, G.
Object Type:
Report
System:
The UNT Digital Library
CASTING OF LONG AND THIN PLATES OF URANIUM-MOLYBDENUM ALLOYS
The development of procedures for the vacuum induction casting of U--Mo alloys into both thin (0.010 to 0.100-in. thick) plates and long (36 in.) plates is described. Melting and casting cycles were developed, and casting techniques established, which resulted in sound, integral plates. These plates were evaluated by radiographic and metallographic examination, and by chemical analysis. The results indicated the feasibility of the process for the fabrication of fuel plates for nuclear reactors. The process promises a potential reduction in fabrication costs, by eliminating waste. (auth)
Date:
November 1, 1961
Creator:
Katz, N.H. & Binstock, M.H.
Object Type:
Report
System:
The UNT Digital Library
The Thermodynamic Properties of the Alkali Halides
The thermodynamic properties of the alkali halides are reviewed, presenting data supplementary to that of Brewer and Brackett, Chem. Rev. 61: 425-432(1961). (T.R.H.)
Date:
November 1961
Creator:
Brewer, L.
Object Type:
Report
System:
The UNT Digital Library
Evaluation of Hydrocarbon Diluents for the Purex Process
Forty-nine hydrocarbon products were evaluated in the laboratory in a search for a highly stable diluent for the organic extractant used in a radiochemical separations plant. The formation of zirconium ligands during chemical and radiolytic degradation increased with boiling point for isoparaffinic producte but was approximately constant for n-parafiins. Of those materials that met the local flash point specification, the n-parafiins were superior. (auth)
Date:
November 1, 1961
Creator:
Dennis, B. P. & West, D. L.
Object Type:
Report
System:
The UNT Digital Library
AN ANALYSIS TO DETERMINE THE PERCENTAGE OF HELIUM BYPASSING THE CORE DUE TO THE REFLECTOR SEALING SYSTEM DURING NORMAL OPERATION OF THE HTGR
The percentage of helium which will bypass the core if the reflector system shown is used is predicted. It is estimated that nominally about 0.1 to 0.2% of the total flow will bypass the core, which is not considered excessive. The most difficult parameter to determine was Z, the gap between the sealing surfaces. The method used to predict Z is presented. The effect of bowing due to a temperature gradient across the seals is discussed. (auth)
Date:
November 15, 1961
Creator:
Nimtz, F.B.
Object Type:
Report
System:
The UNT Digital Library
AN INVESTIGATION OF THE HIGH-TEMPERATURE PROPERTIES OF THE AISI TYPE 502 STEEL
None
Date:
November 28, 1961
Creator:
Martin, W. R. & McCoy Jr., H. E.
Object Type:
Report
System:
The UNT Digital Library
A Summary of Shielding Constants for Concrete
The present knowledge of the shielding constants of concrete is summarized. The densities, elemental compositions, and mixes, where available, are given for a wide range of concretes. From these data, various constants useful for shielding calculations were computed. These constants include the removal cross sections, total cross sections, average thermal neutron absorption cross sections, thermal neutron diffusion coefficients, reciprocal thermal neutron diffusion lengths, total gamma-ray linear attenuation coefficients, gamma-ray energy absorption linear attenuation coefficients, the effective atomic numbers for the determination of buildup factors, and the capture gamma-ray spectra. Experimental results are included where available. (auth)
Date:
November 1, 1961
Creator:
Walker, R.L. & Grotenhuis, M.
Object Type:
Report
System:
The UNT Digital Library
A Process for the Recovery of Uranium From Nuclear Fuel Elements Using Fluid-Bed Drying and Volatility Techniques
A process scheme for the recovery of uranium from fuel elements has been developed. The scheme combines continuous fluid-bed drying and fluoride volatility techniques after initial dissolution of the fuel element in the appropriate aqueous system, hence the designation ADF, Aqueous Dry Fluorination Process. The application of this process to the recovery of uranium from highly enriched, low uranium-zirconium alloy plate-type fuels is described. ln the process, the feed solution is sprayed horizontally through a two-fluid nozzle and is atomized directly in the heated fluidized bed. The spray droplets are dried on the fluidized particles and form a dense coating. Excessive particle growth was limited by the use of air attrition-jets inserted directly in the bed. Aqueous hydrofluoric acid solutions containing l.2 to 3.6 M zirconiuni, 0.007 to 0.03 M uranium, and free acid concentrations from 1 to about l0 M were successfully processed in a 6-in.-diameter Inconel fluid-bed spray dryer. Rates equivalent to about 3.l kg/hr of zirconium were achieved, 160 ml/min with the most concentrated feed solution. Experiments were successfully carried out from 240 to 450 deg C. A new design for a two-fluid nozzle was developed. Extensive work was done to identify the various zirconium fluoride compounds …
Date:
November 1, 1961
Creator:
Levitz, N.; Barghusen, J.; Carls, E. & Jonke, A. A.
Object Type:
Report
System:
The UNT Digital Library
Arc Melting in the Tungsten Electrode Furnace
An arc furnace is described which employs a nonconsumable tungsten electrode and a water-cooled copper hearth. It is used successfully for melting refractory metals and alloys. The furnace is equipped with a vacuum system, an inert gas supply, and an 800-ampere directcurrent power supply. (auth)
Date:
November 1, 1961
Creator:
Williams, D. E. & Levingston, H. L.
Object Type:
Report
System:
The UNT Digital Library
VOID COEFFICIENT OF REACTIVITY ASSOCIATED WITH THE ISLAND REGION OF THE HFIR
Changes in neutron multiplication caused by voids in the island of the HFlR were calculated and measured experimentally. The results indicated that with only water initially in the island the maximum change in neutron multiplication ( DELTA k/sub max) associated with island voids is 0.032 with a corresponding void fraction of 70%. With a simulated 300 g Pu target in the island DELTA k/sub max/ was 0.0l6, and the corresponding void fraction was 42%. In view of these large changes in neutron multiplication, calculations were made to determine what additional materials could be used in the island to reduce DELTA k/sub max/ and what the associated decrease in peak thermal flux wouId be. The results indicated that of the materials considered the use of beryllium in the water island resulted in the smallest decrease in flux for a specified DELTA k/sub max/. To reduce DELTA k/sub max/ to 0.01 required 26% by volume of beryllium in the island; the corresponding reduction in thermal flux, as compared to an all-water island, was about 10%. In order to reduce DELTA k/sub max/ to 0.0l with a 300 g Pu target in the island, the aIuminum-to-water ratio of the target had to be …
Date:
November 15, 1961
Creator:
Cheverton, R.D.
Object Type:
Report
System:
The UNT Digital Library
OPTIMUM FILL VOLUMES IN POT CALCINATION OF RADIOACTIVE WASTES
The 15,000 MW nuclear economy assumed for the long range study of pot calcination costs reported earlier was used as a basis for calculating optimum fill volumes. An algebraic expression was developed for cost as a functmon of the normalized radius of the central void space in a partially filled vessel. Minima of this expression were found for acmdmc and neutralized wastes in 6, 12, and 24in.-diameter vessels. Optimum fill volumes decreased as vessel diameter increased, varying for acidic wastes from 99.8% for 6-in.-diameter vessels to 92.5% for 24-in.diameter vessels. Decreases in costs by using optimum fill volumes instead of the 90% fill volume assumed for all cases in the long range study were small, the largest being an 8% decrease for neutralized wastes in 6- in.-diameter vessels. (auth)
Date:
November 17, 1961
Creator:
Perona, J.J.
Object Type:
Report
System:
The UNT Digital Library
Single Element Flow Tests for Type 3 (SM-2) Fuel Elements in SM-1, SM-1A, and PM-2A Cores
Channel-to-channel flow distribution within Type 3 (SM-2, stationary and control rod fuel elements modified for use in the SM-1, SM1-1A, and PM-2A core support structures and control rod tubes was measured in single element flow testing. Plots of channel-to-channel flow distribution and element pressure drop at various element flow rates are given. Flow distribution for the top-orificed SM-1A and PM-2A stationary elements was within the plus or minus 12% deviation from element average utilized in previous thermal analyses of these cores. Testing of the bottom-orificed SM-1 stationary element and the SM-1, SM-1A, and PM-2A control rod assemblies showed flow distribution exceeded plus or minus 12% devation from average. Simple modifications to the SM-1 stationary element indicated the possibility of improvng fiow distribution in that element. (auth)
Date:
November 27, 1961
Creator:
Krause, P. S.
Object Type:
Report
System:
The UNT Digital Library
Development of a Large Metal Ultrahigh Vacuum Simulation Chamber
A large ultrahigh vacuum chamber was built for environmental testing of components for the SNAP program at temperatures as high as 1000 deg F. The chamber employs diffusion, electronic, and cryogenic pumping to handle high gas loads at high temperature and ultrahigh vacuum. A unique internal heating system, connections, assemblies, flanges, and test set-up jigs are described in detail. (auth)
Date:
November 1, 1961
Creator:
Kamensky, F. J.
Object Type:
Report
System:
The UNT Digital Library
Supporting Analysis and Derivation of Dimensional Tolerance Specifications for Core II of SM-1A & PM-2A
A method is presented for translating inspection measurements of fuel plate spacing to obtain minimum coolant channel clearances under reactor operating conditions. Considerations of fuel plate ripple growth and the inspection procedure used are included. The method is applied to establish dimensional tolerance specifications used for the procurement of SM-1A and PM-2A Core II. (auth)
Date:
November 1, 1961
Creator:
Brondel, J. O.
Object Type:
Report
System:
The UNT Digital Library
Gamma I. A General Theorem-Proving Program for the IBM 704
GAMMA I is a FORTRAN-compiled program for the IBM 704 Electronic Data- Processing Machine. It embodies a certain general, uniform procedure H of mathematical logic for seeking out a proof of any theorem within any mathematical theory which is given in formal axiomatic form. An extended discussion is provided of the underlying method and of the necessary background of mathematical logic. The program is described in detail. (M.C.G.)
Date:
November 1, 1961
Creator:
Robinson, J.A.
Object Type:
Report
System:
The UNT Digital Library
CATASTROPHIC OXIDATION OF HIGH-TEMPERATURE ALLOYS
The growth of nonprotective, crust-like oxide films was encountered in high-temperature alloy systems that contain molybdenum, vanadium, or tungsten as strengthening additions. The cause of accelerated oxidation in such alloys appears to be associated with the characteristically low melting temperatures of oxides of these refractory elements. The factors that contribute to a breakdown of oxidation protection in these systems are outlined and remedial methods which may be used to avoid catastrophic oxidation are discussed. Commonly encountered service failures that have resulted from catastrophic oxidation are also described. (auth)
Date:
November 10, 1961
Creator:
DeVan, J. H.
Object Type:
Report
System:
The UNT Digital Library
Effects of Rolling and Heat Treatment on Anisotropic Irradiation Growth of Uranium. Final Report-Metallurgy Program 6.1.15
An investigation was made to determine the effect of rolling temperature, roll pass design, amount of reduction, and heat treatment before and after rolling on the anisotropic growth rate of uranium under irradiation. The growth rate was found to increase with decreasing rolling temperature and with increasing reduction of area at 300 deg C. The rate of elongation was proportional to the amount of (0l0) component present or, where shortening occurred, to the amount of (l00) component. Oval-edgeoval roll passes resulted in somewhat higher irradiation growth rates than did round roll passes. Recrystallization after rolling effectively reduced the irradiation growth rate of uranium rolled at temperatures of 500 deg C and lower. Irradiation caused length shortening in uranium which was beta quenched after being round-rolled at temperatures of 400 deg C and above, and which was beta quenched after being oval- rolled at temperatures of 300 deg C and above. (auth)
Date:
November 1, 1961
Creator:
Kittel, J. H.
Object Type:
Report
System:
The UNT Digital Library
Waste Treatment and Disposal Progress Report for August and September 1961
Work is being carried out to develop and demonstrate on pilot plant scale integrated processes for treatment and disposal of radmoactive wastes. High-level waste calcination, low-level waste treatment, economic and hazards evaluation, engineering evaluation, disposal in deep wells, disposal in natural salt formations, Clinch River studies, fundamental studies of minerals, and White Oak Creek basin study are discussed. (M.C.G.)
Date:
November 29, 1961
Creator:
Blanco, R. E. & Struxness, E. G.
Object Type:
Report
System:
The UNT Digital Library
Corrosion Associated With Hydrofluorination in the Oak Ridge National Laboratory Fluoride Volatility Process
Studies carried out on corrosion associated with the hydrofluorination- dissolution phase in the fused-salt Fluoride Volatility Process are summarized. Corrosion for hydrofluorination-dissolver vessels used in bench-scale and semiworks-scale process development at ORNL is discussed. The results of a study on construction materials for the dissolution phase are presented. Corrosion studies at ANL are described for comparison purposes. A full-size hydrofluorinator dissolver is described. (M.C.G.)
Date:
November 15, 1961
Creator:
Goldman, A. E. & Litman, A. P.
Object Type:
Report
System:
The UNT Digital Library