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A PROGRAM OF RESEARCH ON MECHANICAL METALLURGY AS RELATED TO FUEL-ELEMENT FABRICATION. Quarterly Progress Report for the Period Ending June 30, 1961
Results of crawing experiments using 3/8-in. bars to provide rod and tube test specimens are reported. In mechanical metallurgy investigations on the relation of formability to imperfection structure the peierls potentials for dislocation of Nb, Mo, Ta, and W were measured. Design and construction of a microextensometer reported. The investigations of the effects of grain size on the stored energy cold work were extended to a new lot OFHC Cu. In studies of formability relations with imperfection structure, a detailed study is reported on the internal-friction relaxation spectrum of Ta relatively unmodified by impurities. A summary of interstitial impurity effects on the internal-friction of Ta is given. (J.R.D.)
Date:
August 31, 1961
Creator:
Trozera, T Z; Koyama, K; White, J L & Chambers, R H
Object Type:
Report
System:
The UNT Digital Library
Biological Effects of Blast. Technical Progress Report
The current state of knowledge relevant to biological blast effects was summarized in a selective manner. Initially, five problems of concern to those who would relate the environmental variations produced by nuclear weapons with biological response and hazard assessment were pointed out. Primary, secondary, tertiary, and miscellaneous blast effects were defined and selected interspecies experimental data of a physical and pathophysiological nature useful in estimating human response were presented. Tentative biological criteria defining safe levels of exposure were set forth as were survival curves for different conditions of exposure in Hiroshima. These were discussed along with the comparative variations in range of the free-field effects as they vary with explosive yield. The fundamental requirement for surviving seconds, minutes, and hours to abet survival for days, weeks, months, and years was emphasized along with the necessity for planning protective measures against all hazardous weapons effects as one attractive alternative for minimizing casualties and maximizing survival in the event of a nuclear war. (auth)
Date:
December 1, 1961
Creator:
White, C. S.
Object Type:
Report
System:
The UNT Digital Library
Response of Dual-Purpose Reinforced-Concrete Mass Shelter
BS>A reinforced-concrete dual-purpose underground parking garage and personnel sheiter designed for a long-duration incident pressure of 40 psi was tested. The sheiter was exposed to shot Priscilla, an approximately 37-kt 700-ft balloon burst (June 24, 1957), at a ground range of 1600 ft (predicted 35-psi peak incident-pressure level). The recorded peak incident pressure at the shelter was approximately 39 psi. Postshot soil borings were made to obtain undisturbed samples for determining soil characteristics. Preshot and postshot field surveys were made to determine the total lateral and vertical displacement of the structure. The test structure provided adequate protection from the effects of the test device at the test GZ distance. Despite failure of the door sealing gasket, a rise in pressure in the interior did not exceed 1.0 psi. The flat-slab roof and supporting structure were more than adequate to resist the 39psi peak incident test loading. (P.C.H.)
Date:
April 1, 1961
Creator:
Cohen, E.; Laing, E. & Bottenhofer, A.
Object Type:
Report
System:
The UNT Digital Library
SOME EXPERIENCES IN THE WELD FABRICATION OF REFRACTORY METALS
Discussion is given on the welding fabrication of tungsten, molybdenum, niobium, and tantalum. Properties which make the four refractory metals important are tabulatcd along with titanium which is given for comparison. Extensive evaluation was conducted using the gas, tungsten arc welding process employing both manual and machine welding. Design data were obtained exclusively from machine welded sheet materials. Flash welding, resistance spot welding and brazing, electron beam welding, and high frequency resistance welding processes were also applied to molybdenum alloys. The oxidation of molybdenum, tantalum, and niobium in flowing air at 2000 deg F is also given. (P.C.H.)
Date:
February 10, 1961
Creator:
Thompson, E.G.
Object Type:
Report
System:
The UNT Digital Library
CRITICAL EXPERIMENTS FOR THE PRELIMINARY DESIGN OF THE ARGONNE HIGH FLUX REACTOR
Critical experiments were performed with two assemblies simulating a cold clean, and an end-of-cycle,- Argonne High Flux Reactor, core. Data were obtained for flux distributions; cadmium ratios; temperature and void coefficients; and control rod, beam hole, and reflector worths. The data obtained furnished confirmation of theoretical predictions. The peak 2200-m/sec flux per unit power was measured as 3 x 10/sup 7/ n/(cm/sup 2/)(sec)(watt) for both cores. The two cores had internal H/sub 2/O thermal columns, 12.7 cm x 12.7 cm x 50.8 cm. These were enclosed by 100-liter fuel zones. The radial reflector was 90% beryllium containing 10% H/sub 2/0 plus Plexiglas by volume. The top and bottom reflectors were H/sub 2/O. The critical mass was 3.58 kg U/sup 235/ with a 1.16 metal-towater ratio in the fuel zone. The critical mass with a 1.60 metal- to-water ratio, taking into account 34.3 kg Type 304 stainless steel, was 7.15 kg U/sup 235/. (auth)
Date:
June 1, 1961
Creator:
de Villiers, J.W.L. ed.
Object Type:
Report
System:
The UNT Digital Library
Electrolytic Dissolution of Power Reactor Fuels in Nitric Acid
The electrolytic oxidation in nitric acid of stainless steel, zirconium, Zircaloy-2, zirconium- uranium alloy, aluminum, and uranium - molybdenum alloy was demonstrated on a laboratory scale. The rate of chemical dissolution of UO/ sub 2/ in nitric acid was measured. Corrosion of stainless steel by these dissolver solutions was measured and found to be negligible. Electrolytic dissolution was demonstrated to be a practical technique for the first step in processing fuel elements of several types of power reactors. (auth)
Date:
October 1, 1961
Creator:
Clark, A. T., Jr.; Meyer, L. H.; Owen, J. H. & Rust, F. G.
Object Type:
Report
System:
The UNT Digital Library
Adsorption of Krypton and Xenon by Various Materials
The adsorptive capacities of various inorganic adsorbents and activated charcoals for krypton and xenon were determined. Columbia-G activated charcoal had the highest capacity for both krypton and xenon at pressures from 0.01 to 125 mm Hg and temperaturens from 2 to 85 deg C. If a value of 1 is assigned to the capacity of this charcoal at 28 deg C for krypton, other charcoals range from 0.63 to 0.84, molecular sieves (except 4A) from 0.11 to 0.20, and some silica genls from 0.05 to 0.07. Various othenr adsorbennts, including one variety of silica gel and molecular sieve 4A, range from 0.005 to 0.032. Molecular sienve 5A and Columbia-G charcoal adsorbed 11.5 times more xenon than krypton. Adsorption of 7.5% water by either of these adsorbents lowerend their capacity for krypton 25 to 30%, while saturating the sieve material ( approximates 15% H2O) lowered the krypton capacity 80%. (auth)
Date:
December 1, 1961
Creator:
Lloyd, M. H. & McNees, R. A.
Object Type:
Report
System:
The UNT Digital Library
A THEORETICAL STUDY OF SIMPLE MANY-ELECTRON SYSTEMS
None
Date:
May 1, 1961
Creator:
Sachs, L.M.
Object Type:
Report
System:
The UNT Digital Library
Transient Temperature Distributions in a Thermally Orthotropic Plate With Non-Uniform Surface Heating
ent temperature variation in a thermally orthotropic plate which is subjected to an arbitrary heating rate distribution along one face with all other surfaces being insulated. Dimensionless temperature histories and distributions determined from this solution are presented for the special, but representative, case of a linearly varying heating rate distribution on plates with varying degrees of thermal orthotropy. These results establish quantitatively the value of a material with high planar and low normal thermal conductivities for applications where it is desired to maintain minimum temperatures on the rear or unheated surface of a heat shield when the heated surface is subjected to a very non-uniform heating rate distribution. The applicability of simplifying assumptions in analyzing such a system is discussed. Experimental temperature measurements in a pyrolytic graphite plate heated by an oxyacetylene flame were made to verify the analytical results. Achievement of satisfactory agreement wss found to be dependent upon use of thermal property values differing from those presently available for this material. This is not unusual in that differences in production methods are known to introduce substantial property variations in anisotropic materials such as pyrolytic graphite. (auth)
Date:
June 1, 1961
Creator:
Hornbaker, David Ross
Object Type:
Thesis or Dissertation
System:
The UNT Digital Library
SNAP 2 REACTOR PUMP DEVELOPMENT PROGRAM (RADIAL GAP PERMANENT-MAGNET PUMP)
A compact electromagnetic pump utilizing a rotating permanent magnet with radial gap was developed for possible application to the SNAP 2 reactor coolant system. The pump was designed for circulation of NaK at 1000 deg F and 11.2 gpm with a developed pressure of 3 psi, operation at 40,000 rpm, minimum weight and size, and high reliability. The performance characteristics of four developmental pump models were measured in a 1000 deg F NaK test loop and compared with design predictions. The capability of the pump design concept was demonstrated, though further development work is needed to meet the SNAP 2 pump requirements. A flow capacity of 6.8 gpm of NaK at 1000 deg F with a developed head of 3 psi with attained at a magnet rotor speed of 40,000 rpm. The weight of this pump is 3 pounds. Reasonable agreement was obtained between the actual pump characteristics and the design predictions. (auth)
Date:
September 1, 1961
Creator:
Sudar, S.
Object Type:
Report
System:
The UNT Digital Library
A PROGRAM OF RESEARCH AND CALCULATIONS OF RESONANCE ABSORPTION. Final Report
A direct numerical integration of the integral equation for the average collision density in the absorber was previously suggested in a discussion of resonance absorption. The implementation of this program is considered. The method of calculation, comparison with experimental data, and the computer code developed are described. The method of integration, computation of cross sections, selection of mesh size, integration interval, outside correction, the Dancoff correction, and unresolved resonances are discussed. Resonance integrals for U/sup 235/ and Th2/sup >/s3>s/sup 2/ were calculated and compared with experiment. (M.C.G.)
Date:
August 28, 1961
Creator:
Nordheim, L.W.
Object Type:
Report
System:
The UNT Digital Library
Seal-Shell-a Digital Program to Determine Stresses and Deflections in an Axisymmetric Shell of Revolution
SEAL-SRELL, a FORTRAN II program registered as code number M0077 at Bettis Atomic Power Laboratory, is written for the Philco 2000 computer with two tape units. The program is designed to determine loads, deflections, and stresses in a thin shell of revolution under axisymmetric end loads and pressure. (auth)
Date:
September 1, 1961
Creator:
Friedrich, C.M.
Object Type:
Report
System:
The UNT Digital Library
THE RELEASE OF Kr$sup 85$ FROM UO$sub 2$ IN ORR CAPSULES
In an attempt to determine the validity of the method of predicting the release of fission gases from U0/sub 2/ suggested recently by Cottrell et al., a series of calculations were made of the expected release of Kr/sup 85/ from prototype Experimental Gas-Cooled Reactor (EGCR) fuel capsule irradiated in the Oak Ridge Research Reactor (ORR). The computed values were then compared with measured values of the per cent Kr/sup 85/ released. In the calculations, the thermal conductivity of the U0/sub 2/ was assumed to be 0.028 w/cm- deg C in the temperature range from 700 to 1600 deg C, and in the absence of a precise knowledge of the helium gap, the cases of a 3-mil helium gas and no gap were treated. Values of the release-rate parameter (D) were estimated from BET surface areas of the U0/sub 2/ pellets. Results showed that the measured values of the per cent Kr/sup 85/ released generally fell within or close to the limits set by the 3-mil helium gap and no gap conditions. There was also a definite correlation between the measured values and the 3-mil gap condition when the clad temperature was about 700 deg C. When the clad temperature was …
Date:
October 1, 1961
Creator:
Scott, J.L.
Object Type:
Report
System:
The UNT Digital Library
Reactor Development Program Progress Report, August 1961
Progress is reviewed on the following reactors: EBWR; Borax-V; ZPR-III- ZPR-VI; ZPR-IX; EBR-I; and EBR-II. An outline of fast and slow reactor safety studies in TREAT is presented. Progress is also reported in applied nuclear and reactor physics; development of reactor fuels, materials, and components; heat engineering technology; separation processes; and advanced reactor concepts. (T.F.H.)
Date:
September 15, 1961
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
TRANSIENT RADIATION EFFECTS IN CAPACITORS AND DIELECTRIC MATERIALS
Measurements of dielectric leakage, capacitance, electric strength, andd charge scattering phenomena were performed at the Kukla and Godiva III critical assemblies for tantalum and aluminum electrolytic, wax- and oilimpregnated paper, mylar, mica, and ceramic capacitors, and for mylar and Vitamin B-impregnated paper. Leakage data indicate that gamma induced conductivity in capacitor dielectric varies directly with gamma DELTA , where gamma is the gamma radiation rate and DELTA is 0.9 for mylar, 0.7 for Vitamin Q-impregnated paper, and approximately 1.0 for the other dielectrics. A small portion of the tantalum oxide conductivity induced by gamma radiation exhibited a recovery time of approximately 150 mu s. Transient capacitance changes due to radiation were non- existent within plus or minus 0.1% for mica and Vitamin Q capacitors. Transient charging of tantalum capacitors was noted during irradiation with no applied voltage. No drastic changes in electric strength were noted during irradiation of mylar and Vitamin Q-impregnated paper. Results are compared with a summary of data previously collected by others. The use of test data in parametric form as a tool for predicting transient radiation effects is discussed. (auth)
Date:
August 15, 1961
Creator:
Wicklein, H. W. & Dickhaut, R. H.
Object Type:
Book
System:
The UNT Digital Library
Chemical Technology Division, Unit Operations Section Monthly Progress Report, May 1961
The experimental results on the oxidation of H from a He stream with CuO pellets were very close to the predicted behavior based on the mathematical model. Experimental measurements of uranyl sulfate loading rates on chloride equilibrated resin showed little variation with solution concentrations. A tentative flowsbeet was proposed for cost analysis of processing a Pebble Bed Reactor. A U-Zr plate was dissolved in nitrate-free Zirflex solution. An authentic TRIGA prototype was processed in engineering-scale equipment. Three 4- stage leacher model dissolution runs were made, two of which used 8 M HNO/sub 3/ and one used 4 M HNO/sub 3/. Flooding rates and holdup data were obtained for sieve plate pulse columns under 5% TBP - l.8 Mi Al(NO/sub 3/)/sub 3/ flowsheet conditions. A Purex waste calcination run (R-37) was made using sodium anid imagnesium to reduce sulfate volatility. (auth)
Date:
December 26, 1961
Creator:
Whatley, M. E.; Haas, P. A.; Horton, R. W.; Ryon, A. D.; Suddath, J. C. & Watson, C. D.
Object Type:
Report
System:
The UNT Digital Library
Annotated Bibliography of Theories of the Equation of State of Ionized Gases and Strong Electrolyte Solutions
This bibliography lists 297 references on the equation of state of ionized gases and electrolyte solutions, including calculations of closely related quantities such as free energy, partition functions, o smotic pre ssure, activity coefficients, and equilibrium compositions of partially ionized systems. A subject index and a supplementary list of 42 bibliographies on plasma physics and similar topics are included. (auth)
Date:
August 1, 1961
Creator:
Brush, S. G. & Wensrich, C. J.
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
HANFORD STUDIES FOR EGCR COMBUSTION CHARACTERISTICS. Summary Report
The temperature, geometry, and flow conditions which exist in the EGCR were duplicated in a mock-up designated as the EGCR Burning Rig to establish the combustion conditions in the reactor. The conditions under which the EGCR Burning Rig will ignite were established and an analytical model was developed which predicts these conditions. Because the Burning Rig cannot exactly dupIicate the reactor situation the final prediction of the safety of the EGCR must rest on computer calculations employing the above analytical model. No advantage in retarding combustion was found in using silicon carbide coated fuel sleeves. The negative results of these tests are due both to the particular geometry of the EGCR moderator and sleeves as well as to the fact that all sleeves tested contained imperfections in the coatings. Chlorine was demonstrated to be an effective agent for extinguishing graphite fires. Concentrations in air of about 1% were observed to extinguish graphite fires at temperatures as high as 1000 deg C. (auth)
Date:
October 10, 1961
Creator:
de Halas, D.R.; Dahl, R.E. & Jackson, J.L.
Object Type:
Report
System:
The UNT Digital Library
In Vivo Gamma Lung Measurements--a Mathematical Model
A low-background facility is described for rneasuring lung burdens of U, Th, and other nuclides in vivo. Problems associated with this method of radiation measurement are discussed. A mathennatical, computer-oriented simulation was devised to gain insight into the relation of the net observed radiation spectrum to the burden of radioactivity in the body or its organs. Chest cavities for persons of three sizes were synthesized in a three-coordinate space comprised of one-inch cubes and including a 9-in.-diameter crystal detector. Data, describing the tissue composition of each cube in the body and the characteristic radiation attenuation for each tissue-type, were coded for use with a program on a high-speed digital computer. Efficiencies for measuring radiation emitted by numerous point sources of enriched uranium were calculated. Data on in vivo measurement efficiency were obtained assuming uniform distribution of radioactive material throughout the lungs and also for nonuniform deposits. The effects of individual size and geometry, and of detector position on the measurement efficiency were determined for these twvo categories and radiation flux distributions on the detector face were computed in some cases. Data are appended and a flow diagram of the computer program is included. (C.H.)
Date:
October 20, 1961
Creator:
Ammann, P. R.; Wilson, C. W. & Mohr, C. M.
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
SPERT IV HAZARDS SUMMARY REPORT
Spert IV is a large pool-type experimental facility for reactor kinetic studies. These studies will include power excursion and instability tests for a variety of reactor designs. Since the Spert IV experimental program requires the performance of tests which will approach, and may exceed the threshold of reactor destruction, the probability of occurrence of the maximum possible accident is not negligible compared with that of other possible accidents. The maximum possible accident for this facility is considered to be a severe nuclear excursion which results in the destruction of the reactor building and the release of 100% of the accumulated fission product inventory of the atmosphere in a steam cloud. The fission product source assumed in the analysis of this accident is an upper limit in view of the nature of the tests to be performed and the heat removal capacity of the system. This postulated accident is independent of the details of core and control system design and is valid for all cores anticipated for use in the experimental program. The major hazards present in the operation of this facility, the precautions to be taken to reduce the probability of an accident, and the consequences of the maximum possible …
Date:
July 1, 1961
Creator:
Bentzen, F. L. & Crocker, J. G.
Object Type:
Report
System:
The UNT Digital Library
OXIDATION OF GRAPHITE UNDER HIGH TEMPERATURE REACTOR CONDITIONS
A kinetic study was conducted to provide information on oxidation of reactor graphites in the temperature range of 450 to 675 deg C and on the effects of reactor environment on oxidation rates. Among the parameters studied were chemical reactivity of the graphite, prior oxidation, a high intensity gamma flux during oxidation, variation of the surface-to-volume ratio of the graphite specimens, neutron bombardment prior to oxidation exposure, and gas flow rates. Rate equations showed apparent activation energies of 50 kcal/mole in the absence of radiation and 30 kcal/mole in the presence of a 1 x 10/sup 6/ r/hr gamma flux. (auth)
Date:
July 1, 1961
Creator:
Dahl, R.E.
Object Type:
Report
System:
The UNT Digital Library