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Bibliography and Index on Vacuum and Low Pressure Measurement

Description: From Introduction: "This paper consists essentially of a) a bibliography, b) an author index, and c) an index of the subject matter of the bibliography. While the primary objective is to focus on vacuum measurement, it was believed essential to include in the bibliography articles on vacuum technology in some measure accessory or essential to vacuum measurement. For maximum usefulness, an index of the subject matter of the references has been prepared. The abstract publications listed in the previous paragraph have been freely drawn upon in preparing the bibliography."
Date: November 10, 1961
Creator: Brombacher, W. G.
Partner: UNT Libraries Government Documents Department
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Camden-Delaware Valley Area (ARMS-II)

Description: Report regarding an Aerial Radiological Measuring Survey (ARMS) of the Camden-Delaware Valley area that was part of a national program to measure environmental levels of gamma radiation. 6,000 traverse miles were examined around Camden, New Jersey.
Date: November 10, 1961
Creator: Guillou, R. B.
Partner: UNT Libraries Government Documents Department
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CATASTROPHIC OXIDATION OF HIGH-TEMPERATURE ALLOYS

Description: 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.
Partner: UNT Libraries Government Documents Department
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Chemical Processing Technology Quarterly Progress Report, April-June 1961

Description: The Idaho Chemical Processing Plant did not operate on fuel recovery during this period, due to extensive renovation and modiflcation of facilities. Potasslum fluoride was found to be an undesirable additive to the barium precipitating agent used in formation of barium fluozirconate, because of precipitation and loss of uranium, although essentially complete precipitation of zirconium was achieved. Addition of hydrofluoric acid with barium precipitant, to achieve a fluoride/zirconium mole ratio of 5.5, was found to give a total zirconium recovery of 05%, including approximately 10% recovered after concentration of the supernate from the original precipitation. Removal of 97% of the zirconium and fluoride from zirconium -uranium dissolver solutions was achieved by precipitation with two moles of sodium formate per mole of zirconium. Uranium was readily recovered from the concentrated filtrate and wash solution by TBP extraction. The preparation and characterization of zirconium nitrate dibutylphosphate are described, and the solubility in Amsco was found to be similar to that of the uranlum dibutylphosphate complex (2 to 4 x 10/sup -5/ M). Stability studies indicated very little, if any, oxidation of ferrous to ferric iron ln a normal raffinate environment, and ferrous iron has a very low molar extinction coefficsent (0.8) compared with that of uranium (15) in the spectral region near 415 m mu . Pilot plant studies of the fluidized-bed calcination process for reduction of radioactive liquld wastes to the more-easily-stored solid form was continued in the two-foot-square calciner with production, for the first time over a prolonged period of continuous operation, of alpha alumina-free product from a feed contalning substantial sodium. Intra-particle porosities ranging from 0.04 to 0.60 were obtained. Differences in alpha-forming tendency of amorphous aluminas with varied calcination histories were demonstrated but attempts at correlation with known variables in fluid bed calciner operation were not successful. Rapid …
Date: November 10, 1961
Creator: Bower, J. R.
Partner: UNT Libraries Government Documents Department
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PT-IP-344-A-FP, Evaluation of Al-Si bond characteristics

Description: Tests in which aluminum-jacketed, Al-Si bonded uranium fuel elements were baked at various temperatures have shown there is a time-temperature relationship for Al-Si layer decomposition. For heat transfer and secondary coolant barrier considerations, the extent of bonding layer deterioration during fuel element irradiation is important. Currently, Al-Si bonded fuel elements show evidence of spire bond separation, and to a lesser degree, can-bond separation following irradiation. Such evidence has aroused concern for the ability of the currently produced Al-Si bonded fuel elements to withstand future reactor operating conditions. Several potential uranium fabrication and canning process improvements are being developed to further advance fuel element stability and performance. Optimization of process conditions based on these improvements may provide the necessary margin of safety for good bond layer integrity, but before a decision can be made to continue improvement of the present process or convert to a new canning process, more information on the stability of the present fuel element bond is needed. This report presents the irradiation phase of a test which was designed to more fully evaluate Al-Si bond integrity under anticipated future reactor conditions.
Date: November 10, 1961
Creator: Clinton, M. A.
Partner: UNT Libraries Government Documents Department
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D-Reactor Graphite Burnout Interim Report: IP-25A(PT-105-532-E)

Description: The oxidation rate of the moderator in D-Reactor has been monitored from samples placed along the length of process tube channel 3478. During the interval from August 8, 1960, to August 10, 1961 were very high, up to 40%/KOD (1000 operating days). From the shape of the front-to-rear burnout profile, the oxidant appears to be oxygen and/or water.
Date: November 10, 1961
Creator: Ryan, B. A.
Partner: UNT Libraries Government Documents Department
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Results of Laboratory Heat Transfer Experiments for C-Reactor Overbore Fuel Channels

Description: The purpose of this report is to present experimental data concerning the heat transfer and fluid flow conditions within a C-overbore geometry process channel for the cases of steady state operation, flow plugging incidents, and inlet piping failure incidents.
Date: November 10, 1961
Creator: Waters, E. D. & Kreiter, M. R.
Partner: UNT Libraries Government Documents Department
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Air Lift Performance at Low Liquid Rates Using Oversized Piping and Lateral Runs

Description: The use of oversized piping in an air lift for transferring solutions at rates less than 5 liters per hour was proven feasible with certain limitations. Reliable operation was also obtained with air lifts containing a lateral run inserted between the point of air injection and the final discharge point. Discharge of the air lifts, especially at low liquid flows, was very erratic under the conditions studied. (auth)
Date: October 10, 1961
Creator: Chamberlain, H. V.
Partner: UNT Libraries Government Documents Department
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CHEMISTRY DIVISION ANNUAL PROGRESS REPORT FOR PERIOD ENDING JUNE 20, 1961

Description: Progress in the fields of nuclear chemistry, isolation and chemical properties of synthetic elements, chemical separation of isotopes, radiation chemistry, organic chemistry, chemistry of aquecus systems, electrochemistry of corrosion, nonaqueous systems at high temperature, and chemical physics for the year ending June 20, 1961, is reported. Separate abstracts were prepared for each topic. (M.C.G.)
Date: October 10, 1961
Partner: UNT Libraries Government Documents Department
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FUEL CYCLE PROGRAM. A BOILING WATER REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Quarterly Report No. 5, July 1961-September 1961

Description: Advanced Fuel Power-Limit Test. Fabrication of the 50 basic fuel assemblies has been completed. The three instrumented fuel assemblies were assembled and calibrated. Equipment and procedures have been prepared for the VBWR stability tests. Forty-two assemblies were irradiated during the last VBWR run. Average exposure for all assemblies which have been irradiated are tabulated. Detailed examination of the centermelt calibration assembly is in progress. Preliminary metallography indicates that centermelting did not occur. The conductivity of UO/sub 2/ is appreciably higher than the design basis of 1.1 Btu/hr-ft/sup 2/ ( f F/ft). This result indicates that the fuel performance limit is higher than the burnout heat flux limit. Heat Transfer and Fluid Dynamics. Use of a rough'' liner in the single rod test section resulted in a l0 to 43% increase in burnout heat flux at various operating conditions. The high-speed movies of the high pressure observational boiling experiments have been edited and assembled. Analytic work is being continued. A theoretical analysis of two-phase pressure drop and density distribution has been made. Experimental Physics. Isotopic analyses of unirradiated fuel pellets show good uniformity. Preliminary calculations of expected isotopic compositions during irradiation are being made. (auth)
Date: October 10, 1961
Creator: Hodde, J.A. ed.
Partner: UNT Libraries Government Documents Department
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HANFORD STUDIES FOR EGCR COMBUSTION CHARACTERISTICS. Summary Report

Description: 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.
Partner: UNT Libraries Government Documents Department
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Neptunium-237 content of E-metal

Description: An analytical program was carried out to obtain an accurate single point measurement of neptunium-37 content versus exposure for E-metal. Two large volume dissolver solution samples representing E-metal from the KW, KE, C, and DR reactors were obtained from the Redox Plant for the basis of analysis. The neptunium-237/U ratio was determined by direct analysis, and the exposure was estimated from the measured Pu/U ratio and the uranium-235 burn-out.
Date: October 10, 1961
Creator: Schneider, R. A.
Partner: UNT Libraries Government Documents Department
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NEUTRON FLUX AND Cd RATIO MEASUREMENTS IN THE HN-1 BEAM HOLE FOR THREE FUEL LOADINGS OF THE OAK RIDGE RESEARCH REACTOR

Description: Neutron flux measurements were made in the Oak Ridge Research Reactor beam hole HN-l shield plug. at low reactor power (N/sub L/) with three fuel configurations. The purpose of these tests was to determine the most favorable fuel arrangement in the region of the experimental hole in order to permit minimization of exposure time of an in-pile slurry loop experiment using pure thoria. It was found that the perturbed thermal neutron flux decreased by factors of 2, each 1.4 in., at the forward end of the beam hole. Maximum and average fluxes observed for three fuel configurations were: high, 9.7 x l0/sup 13/ , 5.6 x 10/sup 13/; intermediate, 8.0 x 10/sup 13/, 4.7 x l0/sup 13/; and present operating, 7.4 x l0/sup 13/, 3.8 x 10/sup 13/. In the high and intermediate configurations fuel elements were located in the outer row of the lattice adjacent to the beam hole. Cadmium ratios were generally high (22 to 111) implying low available epi-cadmium flux under any of these configurations. (auth)
Date: October 10, 1961
Creator: Shor, A.J.f Mauney, T.H.
Partner: UNT Libraries Government Documents Department
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SUMMARY OF HRT RUN 21

Description: The HRT was operated experimentally during run 2l at powers up to 5 Mw to explore the limiting conditions of fuel stability and to demonstrate the reliability of the system. The effect of core pressure on fuel stability was investigated over the range from l250 to 1750 psig. Stable operation at 5 Mw (2.6 Mw in the core) was demonstrated at 1250 psig. At 1600 and 1750 psig, fuel instability accompanied by rapid loss of reactivity occurred at powers down to 2.5 Mw. The threshold power for reactivity loss at intermediate pressures was raised by increasing the fuel acid/sulfate ratio from 0.28 to 0.34. In other studies the fuel temperature was varied from 240 to 275 deg C at sev eral different pressures. In some instances the reactor appeared more stable at the lower temperatures. The effects of suspended solids and oxygen concentration were examined briefly without conclusive results. At times during operation at low pressure and high power, an increase in reactivity, indicating deposition of uranium on the core tank, was observed. During an experiment to investigate this effect, a hole was melted in the core near the equator. The reactor was shut down for examination and modifications to improve the core hydrodynamics. Experiments on internal recombination showed solution recombination- rate constants significantly higher than were previously measured in out-of-pile experiments. Equipment performance was generally satisfactory. A diaphragm failure in one head of the fuel feed pump, minor leakage through four valves, low efficiency of the low-pressure recombiners and rupture of the air-cooled condenser by freezing were the principal difficulties. There was one period of 105 days of continuous operation. During run 21 operations, which extended from October 4, l959, to January 23, l960, the reactor was critical for 2455 hours and produced 5598 Mwh(th). (auth)
Date: October 10, 1961
Creator: Haubenreich, P.N.; Bauman, H.F.; Bradley, N.C.; Engel, J.R.; Kolb, J.O.; Piper, H.B. et al.
Partner: UNT Libraries Government Documents Department
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Final report on production test IP-289-I, Supplement 1, H reactor export flow test

Description: The raw water export system forms the last ditch water supply system to the ``O`` and ``C`` type reactors; in the event of electrical and steam power failure, the export system is designed to supply enough raw water coolant. After the original export orifice was modified twice, the export system was retested.
Date: July 10, 1961
Creator: Cremer, B. R.
Partner: UNT Libraries Government Documents Department
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Study of Factors Influencing Ductility of Iron-Aluminum Alloys. Monthly Letter Report No. 1, June 1, 1961-November 30, 1961

Description: Studies are being made on the effects of variation of aluminum content, heat treatment, surface preparation, and other metallurgical factors on the room temperature ductility of Al-- Fe alloys. The variation of Fe/sub 3/Al order as a function of temperature for 13.9 Alfenol was redetermined under constant instrumental conditions. The variation of the electrical resistivity of the three alloys under consideration with temperature on slow cooling is illustrated. An anomaly in the disordering process near the Fe/sub 3/Al -- FeAl transformation temperature was manifested in the stoichiometric Fe/sub 3/Al alloy. Above a temperature of 490 deg C there was a rapid decrease in the integrated intensity of the (210) FeAl superlattice reflection. The best ductilities obtained corresponded to heat treatments involving slow cooling to produce a high degree of Fe/sub 3/Al order and subsequent annealing of the ordered material for short periods of time near the Fe/sub 3/Al -- FeAl transformation temperature. (M.C.G.)
Date: July 10, 1961
Creator: Rauscher, G. P., Jr. & Nachman, J. F.
Partner: UNT Libraries Government Documents Department
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Summary of the Operational Status of Reactor Control Instrumentation, Report No. 2

Description: The purpose of this review is to report the operability status of the reactor control instrumentation. The status of the instruments was determined twice during the first six months of this year, April 1 and June 1. The information contained in this report is not intended to be a complete description of the control instruments, but only as they apply to reactor control. The assigned Pile Physicist at each reactor reported the status of instrumentation at his reactor. Chart I summarizes the operability status of the various instruments. Chart II shows the relative range of reactor power over which these control instruments apply. Appendix II contains a functional description of the instruments and Appendix III lists how each instrument is used during reactor operation.
Date: July 10, 1961
Creator: Stewart, S. L.
Partner: UNT Libraries Government Documents Department
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Supplement A to PT-IP-263-A-FP evaluation of chemically nickel plated fuel elements

Description: Irradiation of the initial test in this program involving ten tubes of alternately charged nickel-plated C-64 alloy clad test elements and X-8001 alloy control elements has been successfully completed. The test indicated that the nickel-plate spalling problem has been resolved as no significant spalling or flaking was observed during the post-irradiation examination. The second test in this program will be to verify the performance of nickel-plate with a pilot loading (up to 100 charges) of fuel elements which have been nickel-plated on a production basis. The objectives of this test are to demonstrate with a larger scale test that nickel-plate performs satisfactorily and that reducing the nominal plate thickness from .6 mil to .2 mil will not affect the performance of the nickel-plate fuel element. This test authorizes the irradiation of up to 100 columns of OIIN, chemically nickel-plated, C-64 alloy jacketed fuel elements to 200% of normal goal exposure to extend the evaluation of nickel-plated fuel elements on a pilot scale at DR Reactor. Seventy columns will be plated to a nominal thickness of .6 mil and thirty columns to a nominal .2 mil thickness. Twenty measured columns, ten representing each plate thickness, will be charged to monitor the irradiation performance. Effluent samples will be obtained during the course of the test from a pair of tubes, each tube containing a measured monitor charge representing each plate thickness.
Date: July 10, 1961
Creator: Clinton, M. A.
Partner: UNT Libraries Government Documents Department
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Thermoelectric Nuclear Fuel Element Quarterly Progress Report, April-June 1961

Description: Uranium-bearing thermoelectric compounds are now being prepared by tantalum bomb melting and by the hydride process. Tests of devices made up from these compounds indicate that the main fabrication problems are densification and contact bonding. Data from a hot-swaged pellet and a swaged device of US/sub 2/ indicate some promise for that compound. Improvements in techniques of thermoelectric parameter measurements include programming of automatic test data recording at desired intervals around the clock; increased accuracy and versatility of measurements through use of a newly-constructed adjustable precision resistor; and a method for measuring which should lead to an experimental means for determining the thermoelectric figure of merit, Z. Potential profile studies on PbTe pelleta are yielding important information on contact resistance parameters. A fission-fired thermoelectric generator is being prepared for the next in-pile test. (auth)
Date: July 10, 1961
Partner: UNT Libraries Government Documents Department
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