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32,891 Matching Results
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The Measurement Of Free Fission Gas Pressure In Operating Reactor Fuel Elements
The experimental program described has had as its objective the determination of the pressure exerted by free fission gas in operating UO2-filled reactor fuel elements.
Date:
January 23, 1963
Creator:
Reynolds, M. B.
Object Type:
Report
System:
The UNT Digital Library
Accurate Nuclear Fuel Burnup Analyses: Fourth Quarterly Progress Report September-November, 1962
Work has continued on the development of accurate nuclear fuel burnup analysis. Work performed during the fourth quarter is summarized here.
Date:
December 1, 1962
Creator:
Rider, B. F.; Ruiz, C. P. & Peterson, J. P.
Object Type:
Report
System:
The UNT Digital Library
Two-Phase Pressure Losses Third Quarterly Progress Report August 12-November 12, 1962
This is the third quarterly report on the work done under Contract AT(04-30-189, Project Agreement No.27, and covers the period August 12 to November 12, 1962.
Date:
December 12, 1962
Creator:
Janssen, E. (Engineer) & Kervinen, J. A.
Object Type:
Report
System:
The UNT Digital Library
[Sioux City Quadrangle: Field Data]
Documentation outlining characteristics of field data samples taken in the Sioux City quadrangle.
Date:
September 22, 1981
Creator:
National Uranium Resource Evaluation Program
Object Type:
Dataset
System:
The UNT Digital Library
[Sioux City Quadrangle: Sediment Site Data]
Data gathered at stream sediment sites in the Sioux City quadrangle, including applicable water chemistry measurements (e.g., pH, conductivity, alkalinity) and elemental analyses.
Date:
September 22, 1981
Creator:
National Uranium Resource Evaluation Program
Object Type:
Dataset
System:
The UNT Digital Library
[Sioux City Quadrangle: Well Water Data]
Data gathered at well water sites in the Sioux City quadrangle, including applicable water chemistry measurements (e.g., pH, conductivity, alkalinity), physical measurements (e.g., temperature, well description, scintillometer readings), and elemental analyses.
Date:
September 22, 1981
Creator:
National Uranium Resource Evaluation Program
Object Type:
Dataset
System:
The UNT Digital Library
Dynamic Corrosion Screening Tests on Inconel and Nickel in NaCl-MgCl2-UCl3 Bath
Nickel is more susceptible to mass transfer in a 100hr non-isothermal dynamic corrosion system than is Inconel when exposed to a NaCl-MgCl2-UCl3 (50.0-33.3-16.0 mole %) bath at a hot zone temperature 1800 F. No nickel mass transfer was observed in a 500-hr test at 1350 F, but Inconel showed some attack under these conditions. Inconel mass transfer was observed in both tests, being more severe at the higher temperature. On the bases of these preliminary tests, it appears that nickel is a more satisfactory container than Inconel for the chloride bath at temperatures in the region of 1350 F. The chromium is more susceptible to selective leaching from Inconel at 1800 F during a short 100-hr test (0.26% Cr in bath) than in a 500-hr test (<0.001% Cr in bath) at a lower temperature (1350 F ).
Date:
June 19, 1957
Creator:
Jansen, D. H.
Object Type:
Report
System:
The UNT Digital Library
Trip to Selas Corporation of America
On May 23, 1957, a visit was made by the writer to the Selas Corporation of American in Dresher, Pennsylvania. The purpose of the visit was to discuss further investigations into methods of tubesheet brazing by direct heating. Original work along these lines has been carried out at ORNL and is covered by a memo (CF-57-4-57) to W.D. Manly, dated April 16, 1957, and entitled : Investigation of Tubesheet Brazing by a Method of Direct Heating.
Date:
June 18, 1957
Creator:
Franco-Ferreira, E. A.
Object Type:
Report
System:
The UNT Digital Library
Fused Salt Compositions
The compositions of the compounds and fused salt mixtures referred to in the ANP project by numbers are given.
Date:
June 20, 1957
Creator:
Barton, C. J.
Object Type:
Report
System:
The UNT Digital Library
The Volatilization of Fission Products by Melting of Reactor Fuel Plates
Experiments in the controlled melting of irradiated fuel specimens, particularly of the APPR, STR, and MTR types have confirmed that prolonged heating in air at temperatures in excess of the melting point results in the release of a large portion of the radioactivity. On the other hand, a moderate amount of heating in air or steam sufficient only to melt a specimen results mainly in the partial volatilization of the rare gases; the halogens, iodine and bromine; and the alkali metals, cesium and rubidium. In the presence of air or water vapor, strontium and other fission products are not released. At trace concentration of fission products, slow melting of the APPR plate at 1525 C in air or steam effected the release of 50 percent of the rare gases, 33 percent of the iodine, 9 percent of the cesium and traces of strontium. After 25% burn-up, the cesium value increased to about 60 percent. Aluminum alloy of the MTR type, also at trace concentration, upon melting at 700 C released up to 2 percent of the iodine, 10 percent of the rare gases, and negligible portions of other fission products. Zirconium alloy of the STR type after 15 percent burn …
Date:
July 15, 1957
Creator:
Parker, Geogre W. & Creek, George E.
Object Type:
Report
System:
The UNT Digital Library
Experiments on the Release of Fission Products from Molten Reactor Fuels
Experiments in the controlled melting of irradiated fuel specimens, particularly of the APPR, STR, and MTR types, have confirmed that prolonged heating in air at temperatures in excess of the melting point results in the release of a large portion of the radioactivity. On the other hand, a moderate amount of heating in air or steam sufficient only to melt a specimen results mainly in the partial volatilization of rare gases, iodine, bromine, cesium, and rubidium. In the presence of air or water vapor, strontium and other fission products are not released. At trace concentration of fission products, slow melting of the APPR plate at 1525 C in air or steam effected the release of 50% of the rare gases, 33% of the iodine, 9% of the cesium, and traces of strontium. After 25% burn-up, the cesium value increased to about 60%. Aluminum alloy of the MTR type, also at trace concentration, upon melting at 700 C released up to 2% of the iodine, 10% of the rare gases, and negligible portions of other fission products. Zirconium alloy of the STR type after 15% burn-up, when melted at 1850 C, released up to 95% of the rare gases, 90% of …
Date:
March 11, 1958
Creator:
Parker, George W. & Creek, George E.
Object Type:
Report
System:
The UNT Digital Library
Determination of Trace Amounts of Sulfur in Fluoride Salts
A method has been developed for the determination of total sulfur in fluoride salts using the methylene blue procedure. Reduction of sulfate to hydrogen sulfide is achieved through the use of a new reducing mixture consisting of stannous chloride dissolved in concentrated phosphoric acid. The new mixture is effective on microgram amounts of sulfate and offers a major advantage over the red phosphorous reducing mixtures in that larger samples may be taken for analysis. The procedure has been applied to fluoride salts containing from 1 to 500 ppm of sulfur. The coefficient of variation the method is 10 percent.
Date:
June 24, 1957
Creator:
Gilbert, T. W. & White, J. C.
Object Type:
Report
System:
The UNT Digital Library
The Relationship of Aqueous ThO2 Slurry Physical Properties of the Engineering Design of a Reactor System
In a reactor system the principal components affect by slurry properties are the blanket vessel, pressurizer, heat exchanger, and dump tant. The particular properties that affect the operation of these components are: caking, degree of flocculation, foaming, and slime formation. these properties are related to the characteristics of compounds in a reactor system through experience gained in the operation of slurry loops. It is pointed out that the optimum slurry for one component may not necessarily be the optimum for another.
Date:
June 17, 1957
Creator:
Thomas, D. G.
Object Type:
Report
System:
The UNT Digital Library
In-Reactor Autoclave Corrosion Studies II : Autoclave Z-19
In order to test the effect of chromate ion on the corrosion of Zircaloy-2 a 0.04 m uranyl sulfate solution (enriched) containing added acid, copper sulfate and 557 ppm Cr (VI) was autoclaved with rocking at 280 C for about eight days in the HB-5 facility of the LITR. The average corrosion rate established by the a rate of consumption of oxygen was 9.9 mpy at a power density of 4.9 w/ml. It is indicated by comparison with a previous corrosion study under LITR radiation that the presence of Cr (VI) had no significant effect on the radiation corrosion of Zircaloy-2 by enriched uranyl sulfate solutions. However, the data are not conclusive and may be interpreted as showing a low corrosion rate for a limited period (i.e. a short term inhibition) followed by a correspondingly rapid corrosion.
Date:
March 22, 1957
Creator:
Warren, K. S.; Davis, R. J. & Jenks, G. H. (Glenn Herbert), 1916-
Object Type:
Report
System:
The UNT Digital Library
Hedstrom Plot for the Calculation of Laminar Flow Pressure Drop for the Bingham Plastic Materials with Hedstrom Numbers from 0 to 10(15)
The results of a machine calculation of a modified Fanning-friction-factor Hedstrom plot for Hedstrom numbers from 0 to 10(10) are presented in graphical and tabular form.
Date:
June 20, 1957
Creator:
Thomas, D. G.
Object Type:
Report
System:
The UNT Digital Library
Compilation of Various Undocumented Classified Memoranda on Sherwood Program
This compilation includes the following subjects: (1) Spectroscopic studies, (2) Neutral carbon in the vacuum carbon arc, (3) Anode effects, doppler blast effects, and stark broadening, (4) Neutrals in the high-current carbon arc; (5) Photon breakup of N2 in the high-current carbon arc, (6) Ion density in the high current carbon arc, and (7) Recombination cross-section for fast hydrogen ions and slow electrons. Minor revisions have been made in the subject memoranda in incorporating them in the compilation.
Date:
June 28, 1957
Creator:
McNally, J. Rand (James Rand), 1917-
Object Type:
Report
System:
The UNT Digital Library
Testing of Adsorptive Capacity of Charcoal Beds : HRT Test No II-A 10 b
During the pre-startup phase of the HRT operations, moisture was accidentally introduced into the charcoal bed adsorbers in the off-gas system. Tests have been made to determine the effect of wetting upon the adsorptive properties of the charcoal. The work was divided into two phases, testing of fresh charcoal in the laboratory and testing of the HRT charcoal beds in situ. It is recommended that the beds be dried by heating them to about 40 C and purging each with one to two liters/min of dry instrument air.
Date:
June 4, 1957
Creator:
Van Winkle, R. & Wiethaup, R.R.
Object Type:
Report
System:
The UNT Digital Library
Performance of HRT Charcoal Beds
The expected performance of the HRT carbon beds was calculated for various reactor operating conditions. the calculations indicated that the flow rate of sweep gas will have to be limited to prevent excessive activity discharge. Data on activity discharge are included.
Date:
June 4, 1957
Creator:
Weeren, Herman O. & Lee, John (W. John)
Object Type:
Report
System:
The UNT Digital Library
A Preliminary Study of Pre-Solvent Extraction Treatment of Stainless Steel-Uranium Fuels with Dilute Aqua Regia
The continuous dissolution of 304 stainless steel and stainless steel-UO2 alloy in dilute aqua regia was studied with subsequent stripping of the dissolver product to remove chloride ion. The process has the advantage of producing, by means of a simple head end treatment, a solvent extract feed in a conventional nitric acid medium so that existing solvent extraction processes, materials of construction and waste disposal methods can be used. The purposes of this study were to investigate the variables affecting the dissolution process and to obtain dissolver scale-up data, and to investigate the removal of chloride from the dissolver product and the variables affecting the stripping operation. A continuous flooded pot dissolver was used. It has the advantages of stability of operation and ease of control in comparison with column dissolvers and requires a minimum of mechanical processing prior to dissolution. Stripping of the dissolver product to remove chloride ion was studied in a 4-in. diameter Pyrex bubble-cap column containing 12 single bubble cap plates. Continuous dissolution rates and dissolver product stainless steel loading were correlated with aqua regia feed composition, acid feed rate and surface area exposed to reaction. Profiles of chloride concentration down the stripping column were obtained …
Date:
October 11, 1957
Creator:
Kitts, F. G. & Perona, J. J.
Object Type:
Report
System:
The UNT Digital Library
Effect of Core Corrosion Sample Assembly on HRT Critical Concentration
An estimate has been made of the critical fuel concentration in the HRT, taking into account the effect of the core corrosion sample assembly. The estimate is based on a number of previous calculations of critical concentration in an un-poisoned reactor and one calculation of critical concentration as a function of poison level. The makeup of the first core corrosion sample assembly was used in calculating equivalent neutron poisoning effects. Figure 1 shows the estimated critical concentration as a function of temperature with the corrosion sample assembly in place. At 280°C, the assembly raises the critical concentration by 0.6 g U-235/kg D2O. This effect is equivalent to a uniformly distributed poison equal to 4.1% of the fission cross section. The equivalent poison is greater at lower temperatures, where the uranium concentration is lower.
Date:
July 18, 1957
Creator:
Haubenreich, Paul N.
Object Type:
Report
System:
The UNT Digital Library
Nuclear Computations for HRE-3 Design : Equilibrium Results
Various nuclear characteristics of two-region spherical homogeneous reactors have been computed in order to provide information for the design of HRE-3. Equilibrium isotope concentrations were established using an ORACLE code, and a two-group model was used to obtain critical concentrations and flux distributions. Breeding ratio is plotted as a function of reactor size, blanket thorium concentration, and other design and operating parameters, and the time required for a demonstration breeding is discussed. Tables of results, including neutron balances, are given for selected reactors. a number or relations are presented for estimating the effects of fission products, copper, corrosion products, H2O, and the core tank on breeding ratio.
Date:
July 10, 1957
Creator:
Rosenthal, M. W. & Fowler, T. B.
Object Type:
Report
System:
The UNT Digital Library
Radiation Level in the Stator Region of the HRT Fuel Circulation Pump
The gamma dose rate in the motor region of the HRT fuel circulation pump was measured with the pump scroll full of radioactive solution. Extrapolation of the data to the solution activity expected in the pump under normal operation gives a dose rate well below that which would result in excessive gas production in the stator can within the life of the pump. The above dose rate does not include the effects of fast neutrons from the fuel solution or of the general cell radiation level in the vicinity of the pump. It appears that the possibility of gas production in the stator from the cell background radiation is sufficiently great to warrant the installation of a shield around the outside of the motor end of the fuel circulating pump.
Date:
July 3, 1957
Creator:
Engel, J. R.
Object Type:
Report
System:
The UNT Digital Library
Thermal Stress Testing of SM-2 Fuel Elements : Final Report January 1, 1959 to July 1, 1960
Thermal stress testing was performed on portions of reactor fuel elements.
Date:
September 20, 1960
Creator:
Christenson, J. A. & Kortheuer, J. D.
Object Type:
Report
System:
The UNT Digital Library
Summary of Reactor Design Information from Three Years Operation of a Small PWR
Reactor design information obtained from three years' operation of a small pressurized water reactor, the SM-1 (formerly APPR-1), presented and discussed
Date:
September 9, 1960
Creator:
Gallagher, J. G.
Object Type:
Report
System:
The UNT Digital Library