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SOME EXPERIENCES IN THE WELD FABRICATION OF REFRACTORY METALS

Description: 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.
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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THE RECOVERY OF FISSION PRODUCT RARE EARTH SULFATES FROM PUREX 1WW

Description: Cerium- and 144 promethium-147, accompanied by rare earths resulting from fission or decay can be removed from Purex 1WW in>90% yield as an insoluble, crystalline sodium-rare earth double sulfate. Precipitation is initiated by a one-to-three hour equilibration at 90 deg C and centrifugation at 90 deg C to take advantage of the lower solubility of the double sulfate salt at a higher temperature. The sulfate concentration should be one molar and the solution pH at the time of precipitation should be 0.5 to 1.5. The addition of tartrate ion to complex the iron allows the use of a higher pH and sulfate concentration, gives a more complete separation from iron, and a quantitative recovery of the rare earths. The double sulfate precipitate can be dissolved in dilute nitric acid or converted to the carbonate and then dissolved to yield a solution for further processing. The double sulfate precipitation of the rare earths, with tartrate added, gives a good separation from impurities. One-cycle decontamination factors of 150 for Zr-Nb and 1100 for Ru-Rh have been achieved in laboratory tests. Tests in the Purex head-end equipment with up to twomegacurie batches of cerium have corroborated the laboratory results. Decontamination factors of 70 for iron, 10 for zirconium, 20 for niobium and 25 for ruthenium have been obtained. It was found wise to limit the batch size because decay heat leads to partial calcination in the centrifuge and to difficulty in redissolution. (auth)
Date: May 10, 1961
Creator: Wheelwright, E.J. & Swift, W.H.
Partner: UNT Libraries Government Documents Department
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A Method of Determining the Intermediate Energy Neutron Dose

Description: The intermediate energy neutron flux existing outside the biological shielding of reactors has not been studied to any great extent previous to this time, because of the lack of an instrument capable of detecting neutrons in the intermediate energy range. The instrument used at the MTR utilizes polyethylene spheres of various sizes to give different amounts of moderation and absorption to the impinging neutrons. A procedure for the approximate determination of the relative number of intermediate energy and fast neutrons is given. By graphical methods the following information is obtained: (1) fraction of intermediate neutrons, (2) fraction of fast neutrons, and (3) the approximate average energy of the fast neutrons. Since the instrument described can be used to determine the thermal neutron flux independent of intermediate and fast fluxes, only one instrument is required to determine the neutron flux in all three energy ranges. Dose calculations indicate the intermediate range neutrons give a dose greater than the dose delivered by fast neutrons around the MTR-ETR reactors under normal operating conditions. (auth)
Date: March 10, 1961
Creator: Hankins, D. E.
Partner: UNT Libraries Government Documents Department
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CHARGING AND DISCHARGING OF DEMINERALIZER RESINS. CORE I, SEED 2. Test Results (T-612085). Section 1

Description: An investigation was conducted to flush the 1BD coolant purification system demineralizer of resin and to measurc the radiation level at pornts on the resin discharge line. It was found that the system demineralizer was satisfactorily flushed of resin. It was also found that the amount removed is measured by the amount required to recharge the demineralizer. (J.R.D.)
Date: January 10, 1961
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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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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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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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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E-N and blanket conversions from analysis of tubes irradiated at H

Description: A block-loading of striped columns and tubes simulating a blanket loading were analyzed for product (Pu, tritium, E-metal) after irradiation in IP-255-A-9-FP. Results are rationalized to full-pile values; pile conversion ratios and pile gains are given.
Date: February 10, 1961
Creator: Lang, L. W. & Nechodom, W. S.
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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Supplement A to PT-IP-183-A-98-FP: Evaluation of projection fuel elements for use in K process tubes

Description: The objective of this supplement is to authorize charging of ten tubes of ``bumper`` fuel elements and controls into each K Reactor. The test is designed to reevaluate the reduction in hot-spot incidence associated with fuel alignment within K Reactor ribbed process tubes for both natural and enriched uranium I&E fuel elements of the KIV geometry.
Date: April 10, 1961
Creator: Clinton, M. A.
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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Proposal for charging heat treatment test elements GEH-10-44 & 45

Description: The objective of this irradiation is to determine the differences in irradiation behavior; typified by dimensional changes, surface roughness, and overall distortion; resulting from elements of similar fabrication history but different beta heat treating schedule. The fuel will be the inner tube only of an NPR fuel assembly. Both elements were heated in chloride salt at 730C; one was rapidly quenched and the other air cooled to obtain a wide variation in grain size and structure and residual stress.
Date: January 10, 1961
Creator: Kemper, R. S. & Young, F. E.
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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