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12,523 Matching Results
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HRP Dynamic Slurry Corrosion Studies : Quarter Ending April 30, 1957
The assembly of a second thorium oxide slurry corrosion test facility, loop BS, has been completed and 2010 hr of operation on slurry have been logged. This second test loop has proved satisfactory from an .operational standpoint. Corrosion data and operational observations are given for six thorium oxide slurry runs made at 300 C in 100A pump loops BS and CS. A new development model of the rotator for an in-pile slurry toroid is described.
Date:
April 30, 1957
Creator:
Compere, E. L. (Edgar L.); Savage, H. C.; Reed, S. A.; Warner, R. M.; Ulrich, W. C.; Cole, H. D. et al.
Object Type:
Report
System:
The UNT Digital Library
Two-Group Analysis of Thermal, One-Dimensional, Multi-Region Spherical Reactors
This technical report described the formulation of a set of two-group neutron diffusion equations and the solution for the critical fuel cross section in a one-dimensional, multi-region spherical reactor. A subsequent report will describe the ORACLE code developed for survey calculations using this method.
Date:
May 1, 1957
Creator:
Nestor, C. W.
Object Type:
Report
System:
The UNT Digital Library
Comments on the Transportation of Irradiated Fuel and Radioactive Wastes for M Louis Armand, Euratom Group
General considerations involving the transportation of irradiated fuel and radioactive wastes are reviewed. It is assumed that many reactors will supply feed to a few large multipurpose chemical plants which ultimately send radioactive waste to a few disposal sites. General economic considerations of irradiated fuel reprocessing, economic aspects of the nuclear economy complex, growth predictions of the nuclear power economy in the U.S., general requirements for the shipment of fuel and waste, regulations applicable to fuel shipment, and permissible radiation levels are discussed.
Date:
May 6, 1957
Creator:
Culler, F. L.
Object Type:
Report
System:
The UNT Digital Library
A Brief Review of thermal Gradient Mass Transfer in Sodium and NaK Systems
The fact that material transport does occur under conditions of finite temperature difference in a flowing molten metal system was established. The rate mass transfer was thought to be either diffusion limited or solution rate limited. It is believed that the mass transfer of structural materials in Na or NaK systems is solution rate limited. The limiting process has not been qualitatively or quantitatively confirmed for the Inconel-Na or Inconel-NaK system. Increasing the maximum system wall temperature increases the amount of mass transfer, at least above 1300 deg F. The effect of the total temperature difference across the system on the amount of mass transfer was determined.
Date:
February 11, 1957
Creator:
DeVan, J. H. & West, J. B.
Object Type:
Report
System:
The UNT Digital Library
The Effect of Fluctuations in the Widths on Neutron Reaction Cross Sections
The general Wigner-Eisenbud theory is used to develop a method of analysis for the cross sections of fissionable nuclei. The method is employed in giving a reasonable description of the low energy cross sections in U/sup 235/. The single level fit for U/sup 235/ is known to be unreasonable. Many level expressions for the cross sections are derived--the only approximation to the general theory being the neglect of all but a small group of resonances. It is shown that regardless of the number or definition of the fission channels the many-level expressions require few level parameters: the E/sub lambda /, GAMMA / sub lambda n/, GAMMA /sub lambda gamma / and GAMMA /sub lambda F/ of the single level theory for each resonance and a few additional parameters pertinent to the interference between levels. The interference terms are described and shown to be important. The shape and size of the U/sup 235/ cross sections below 2 ev are fitied to within one per cent using (a) only one negative energy resonance of smaller size than in the single level fits (b) no additional levels to fit the shape other than the observed levels at positive energies (c) three interference parameters …
Date:
June 1, 1957
Creator:
Dresner, Lawrence
Object Type:
Report
System:
The UNT Digital Library
Operating Instructions for the UNIVAC Program OCUSOL-A : a Modification of the Eyewash Program
The Eyewash program, written by James H. Alexander and Nancy D. Given, provides solutions of reactor criticality problems in spherical geometry by means of the group diffusion method. It employs thirty lethargy groups (plus one thermal group) in nine regions. The input consists principally of specifying the geometrical scaling factor, boundaries and compositions of the various regions, and temperature level. The output includes the value of vc that would render the system critical, the relative fission density distribution, fissions, absorptions, and leakages in each lethargy group in each region, and, if desired, an edit of the flux at each space point, each lethargy, and an edit of the macroscopic cross sections for each lethargy, each region. OCUSOL-A is a minor modification and extension of Eyewash. It provides for the computation and editing, on the supervisory control typewriter, of the total absorptions in selected nuclides in the various regions. This information is useful in the computation of breeding ratios and the preparation of detailed neutron balances, and in the estimation of flux-averaged cross sections for use in estimating the rate of change of concentration of the various nuclides with burn-up. The program also provides for saving and transferring the final fission …
Date:
June 5, 1957
Creator:
Alexander, L. G.; Carrison, D. A.; Roberts, J. T. & Van Norton, R.
Object Type:
Report
System:
The UNT Digital Library
Two Group Calculations for Flux Distribution and Critical Mass in Clean Cold ORR Cores
A series of two-group calculations has been made on the Oracle for the purpose of obtaining critical-mass and flux distribution data for various ORR core configurations. The 3G3R code of Bate, Einstein, and Kinney was used, together with the RSP code developed by Nelson. This made it possible to obtain results for the three-dimensional case. The results, which are presented graphically, are intended to serve as a guide for the design of experiments until such time as actual measurements are available. The calculations were performed for the "clean cold" case, and it should be realized that the presence in the core of experiments and of fission products built up during operation will materially alter the flux patterns found. It is believed that the critical-mass data are accurate to within 10%. Within the fuel region it is believed that the thermal-flux patterns are the also accurate to this degree. Comparison of the results with MTR critical experiments, however, indicates that the thermal flux in the reflector in the vicinity of the fuel-reflector interface may have been underestimated by a factor of as much as 1.3. It should also be recalled that in a two-group calculation the "fast flux" is often a …
Date:
March 11, 1958
Creator:
Binford, F. T.
Object Type:
Report
System:
The UNT Digital Library
Evaluation of Coated Al2O3 and Tungsten Carbide Bearing-Journal Assemblies in Westinghouse 100A Pump (Summary of Runs S-96A and S97)
Preliminary results of tests wit the Westinghouse 100A pump indicate that Al2)3 and tungsten carbide coated bearing-journal assemblies prepared by the Linde process are not promising as substitutes for the graphitar-stellite combination. The front Al2O3 assembly failed at start-up with water and both front and rear tungsten carbide assemblies failed after 121 hours with water at 245 C.
Date:
May 3, 1957
Creator:
Kitzes, A. S. & McLaughlin, C. A.
Object Type:
Report
System:
The UNT Digital Library
Summary of Corrosion Data for HRT Mockup Operational Period Ending November 5, 1956
The operation of the HRT mockup was on 0.042m UO2SO4, 0.024m H2SO4, and 0.005m CuSO4 at 280 C and 1400 psi pressure. with the O2 content at near 500 ppm. The pump showed bearing wear and high corrosion. The letdown heat exchanger was removed from the mockup and sectioned. The metallographic examination revealed corrosion. Results of corrosion runs on Ti, zircaloy-2, and stainless steel are given.
Date:
May 22, 1957
Creator:
Wacker, R. E. & Griess, J. C.
Object Type:
Report
System:
The UNT Digital Library
Final Process Design for Leak Detector System for Special Flanges
The leak detector system consists of one gas pressurized reservoir containing heavy water, a tubing manifold connecting the pressurizer to six separate lines each connected to one of the flanges, tubing lines leading from the second hole on each of three flange pairs (dome and heat exchanger flanges) back into the instrument room, plus required valves and fittings. A schematic diagram of the system in included.
Date:
May 29, 1957
Creator:
Mason, Edward A. (Edward Allen), 1926-1994
Object Type:
Report
System:
The UNT Digital Library
Stress Corrosion in the HRT Mockup
Stress corrosion was found in 8 components of the HRT mockup; only of four of these actually shut down the loop. All of the failures have occurred in the high-pressure system of the loop.
Date:
May 20, 1957
Creator:
Harley, P. H.
Object Type:
Report
System:
The UNT Digital Library
Effects of Letdown Rates and Oxygen Injection Rates on Xenon Poison Level and Excess Oxygen Concentration in the HRT
Calculations indicate that it is impossible, even at high oxygen injection rates, to insure an excess of oxygen in the HRT fuel solution if the bubble letdown rate is more than 1 or 2 liters per minute. If, on the other hand, no bubbles are allowed to form, a reasonable excess oxygen concentration can be maintained with an oxygen injection rate which would not tax the capacity of the off-gas system. The xenon poison will be reduced to less than 2% by liquid letdown alone, and if an iodine absorption bed is installed below the catalytic recombiner, the xenon poison should be less than 1% without any bubble letdown. Therefore, it is recommended that sufficient copper be added to prevent the formation of gas bubbles and that the oxygen injection rate be limited to a value which would permit adequate holdup times in the present charcoal adsorption beds, assuming this quantity is sufficient to meet corrosion requirements.
Date:
May 31, 1957
Creator:
Haubenreich, P. N.
Object Type:
Report
System:
The UNT Digital Library
Low Attack Rates Observed in Toroid Tests with 28 Micron 1600 C Fired ThO2 Spheres
Results of previous toroid tests with 28 micro 1600 fired ThO2 spheres are analyzed. Four possible explanations for the essentially zero attack rates are offered and discussed: peculiar motion of toroids or some unknown wall effect; rotational hydrodynamic forces cushioning the particle impact; the bed was not in suspension and not sliding over the walls of the toroid; round particles are not as erosive to oxide film surface as sharp pointed particles.
Date:
May 28, 1957
Creator:
Thomas, D. G.
Object Type:
Report
System:
The UNT Digital Library
In-Reactor Autoclave Corrosion Studies : LITR I. Outline of Methods and Procedures
During the development of in-reactor corrosion experiments three types of bomb designs have been developed for following quantitatively the consumption of oxygen in a bomb which is fabricated from the particular metal under test. The bombs are designed so as to be rocked continuously during their in-reactor exposure, consequently agitating the uranium salt solution contained in the bomb and permitting gentle movement of the solution past metal specimens and other corroding surfaces. The oxygen pressure is produced within the bomb at the beginning of the run either by the withdrawal of gas from an oxygen tank or by the decomposition of hydrogen peroxide added just prior to closure. Temperature measurements are obtained by mans of thermocouples placed in a well within the bomb body or, in the latest design, there will also be thermocouples staked at several points on the outer wall of the bomb. The assembly is inserted in an experimental access hole in the reactor, brought up to a predetermined operating temperature by means of an electric furnace in close contact with the bomb, and allowed to remain for the desired amount of exposure to reactor radiation. Periodic measurements of temperature and pressure are made and the bomb …
Date:
May 22, 1957
Creator:
Warren, K. S. & Davis, R. J.
Object Type:
Report
System:
The UNT Digital Library
Time Behavior of Fuel Concentrations in Single-Region Reactors Containing U-233, U235, Th-232 and Fission Product Poisons
Analytical expressions were obtained for the time behavior of fuel concentrations and fuel-feed rates in single-region, spherical, UO3-ThO2-D2O reactors.
Date:
February 26, 1957
Creator:
Gilbert, Nathan & Kasten, Paul R.
Object Type:
Report
System:
The UNT Digital Library
Capillary Flowmeter
The HRT leak detector system consists of four headers each of which are connected on one side to a common supply of pressurized water and on the other side by individual tubing to the ring grooves of approximately twenty flanges. There are two methods of detecting the loss of water that are particularly applicable to the HRT: (1) By the loss of pressure in a constant volume system; (2) By the measurement of flow from a constant pressure system. It was determined to investigate the second method which requires a flowmeter capable of measuring flows of .5 cc or less of water per hour. The experiment flowmeter constructed performed almost exactly as the design calculations predicted.
Date:
February 20, 1957
Creator:
Hise, E. C.
Object Type:
Report
System:
The UNT Digital Library
Cross Sections for OCUSOL-A Program
The OCUSOL-A program (ORNL-CF-57-6-4) for Univac is a modification of the Eyewash (ORNL-1925) multi-group, multi-region reactor code. The group=energy-lethargy-temperature relationship are given in Table A. The element code numbers are given in Table B. The cross sections now on the sigma-tape are given in tables in the Appendix numbered with the element code number. This technical report explains the bases for choosing the cross sections.
Date:
June 11, 1957
Creator:
Roberts, J. T. & Alexander, L. G.
Object Type:
Report
System:
The UNT Digital Library
Specifications for Cleanliness Requirements High Level Volatility Lab. 4507
Specifications are presented for cleanliness during installation of piping and equipment in the High Level Volatility Laboratory, Bldg. 4507.
Date:
June 6, 1957
Creator:
Ruch, J. B.
Object Type:
Report
System:
The UNT Digital Library
Effect of Slurry Physical Properties on Heat Exchangers and Pump Characteristics
Design calculations were made for a system consisting of a pump, one hundred feet of pipe, and a heat exchanger to remove 1 Mw of heat from various aqueous thorium oxide slurries. The rheological properties of the slurries were varied over a range of yield stresses from 0 to 1.5 lb/sq ft and of coefficients of rigidity from 1/2 to 2 centipoise. Two different cases were studied: a heat exchanger having fixed axial and radial delta T in which the tube length was allowed to vary and a heat exchanger having fixed tube length in which the axial and radial delta T were allowed to vary. It was shown that the pump power must be increased by a factor of 15 to 30 in order to maintain satisfactory operation of the heat exchanger as the slurry yield stress is increased form 0 to 1.5 lb/sq ft. However the pump power is essentially independent of heat exchanger tube diameter for any given slurry. The rated capacity of a slurry heat exchange is essentially independent of slurry yield stress and coefficient of rigidity, provided that the tube velocity can be suitably increased as the slurry yield stress in increased.
Date:
June 10, 1957
Creator:
Thomas, D. G.
Object Type:
Report
System:
The UNT Digital Library
Observed performance of the Fuel Sample Cooler
Measurements of flow rate through the fuel high-pressure system sampler indicate that the average flow rates is about 0.29 gpm (145 lb/hr) plus or minus 50%, which affords an adequate purge of from 12 to 36 volumes through the sample line if the full fifteen minutes of purging is allowed before isolating the sample. The fuel sample cooler was fund to have adequate capacity to reduce the temperature of the fuel solution form about 275 to 70 C, using pre-heated cooling water at 70 C. Uncertainties in temperature measurements make it impossible to estimate an observed over-all heat transfer coefficient.
Date:
June 3, 1957
Creator:
Van Winkle, R. & Wiethaup, R. R.
Object Type:
Report
System:
The UNT Digital Library
Observed Net Heat Loss from the HRT High-Pressure System
An estimate has been obtained of the heat that should be generated in the HRT core in order to hold the system at operating temperature under no-load conditions. This estimate was made by measuring the feed-water rate to the package boiler during an oxygenated water rung. Results are summarized.
Date:
June 4, 1957
Creator:
Van Winkle, R. & Wiethaup, R. R.
Object Type:
Report
System:
The UNT Digital Library
The Disposal of Power Reactor Waste Into Deep Wells
For various reasons it is not possible to leave the uranium or other nuclear fuel in a power reactor until all of it has been "burned up" by fission. In the case of liquid fuel (homogeneous) reactors a small part is continuously bled out, purified and returned. In the case of solid fuel reactors, fuel elements are periodically removed, reprocessed and the "unburned" fuel put back into service. In both cases the purification produces wastes which contain radioactive fission products and transuranic elements, and it is with the disposal of these wastes that we are concerned. For technical reasons, we will limit our consideration to the wastes from the processing of solid fuel elements, and from the processing of the very similar solid "blanket" elements in which fissionable fuel is made from non-fissionable isotopes of uranium and thorium by interaction with neutrons in the outer regions of the nuclear reactor.
Date:
June 13, 1957
Creator:
De Laguna, Wallace, 1910- & Blomeke, J. O.
Object Type:
Report
System:
The UNT Digital Library
HRT Letdown Valves
To supply information about the several letdown valves which have been in HRT service, a typical valve plug examined by the Metallurgy Section confirms the belief that chemical attack on the Stellite #6 was occurring. It appears that most of the corrosion was due to rinse solutions, since this phenomenon has not been noted on letdown valves in the HRT Mockup where over 90% of the operating time has been with UO2SO4 solutions.
Date:
June 5, 1957
Creator:
Billings, A. M.
Object Type:
Report
System:
The UNT Digital Library
September, 1956, Measure of Radiation level of HRE Fuel System Components (After Storage for 27 Months)
Radiation level measurements of various HRE fuel system components, made in September, 1956, after 27 moths of storage, are compared with the June, 1954, readings before storage. Measurements were made with a standard cutie pie and results tabulated.
Date:
June 5, 1957
Creator:
Haynes, T. E. & Van Winkle, R.
Object Type:
Report
System:
The UNT Digital Library