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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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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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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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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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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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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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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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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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Examinations of Pump Impellers From Sodium and Fused Salt Pump Endurance Tests

Description: Examinations of three Inconel pump impellers were made to establish the extent of cavitation damage and degree of carburization sustained during pump endurance tests. The pumps, two of which circulated fluoride salt and one sodium, operated for the bulk of the test programs in the temperature range 1000 to 1250 deg F. Cavitation damage was manifested in each of the impellers by the formation of deep pits (in excess of 1/4-in.), the location of damaged areas varying with impeller geometry. Pit formation appeared to have occurred by uniform rather than preferential removal of metal components. Each of the impellers exhibited heavily carburized zones along exposed surfaces. The cause of carburization may be associated with the type of gas purge utilized for these pumps. (auth)
Date: April 10, 1961
Creator: DeVan, J. H.
Partner: UNT Libraries Government Documents Department
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Critical Path Scheduling in Maintenance

Description: Summary: The following narrative interspersed with figures and attached reference exhibits is designed to acquaint the reader with the scheduling procedure developed at ORGDP, trial results and evaluation, subsequent improvement, further application, and use in conjunction with our IBM 7090 Computer.
Date: April 10, 1961
Creator: Gritzner, C. L.; Jones, J. P. & Ellis, J. M.
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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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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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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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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Fission-Product Plateout From a Helium Gas Stream

Description: The experimental procedures used and the results obtained in studying the plateout of nongaseous fission products that may escape from a helium gas stream are discussed. Results show that significant quantities of high-activity fission products are volatilized from powdered uranium dioxide at 1800 deg F, and the gross activity for each of the materials (Nichrome, type 302 stainless steel, and quartz) exposed to the fission product vapors follows the same general pattern. Some indication is given that high nickel content promotes plateout at high temperatures as is shown by testing nickel, Nichrome, types 310 and 405 stainless steel, and quartz. Radioisotopes that were identified as being deposited on the specimens were iodine-131 and -133, molybdenum-99, ruthenium- 103, and tellurium-132. Plateout occurred on all of the materials exposed to the helium stream from 1700 to 400 deg F. (N.W.R.)
Date: March 10, 1961
Creator: Johnson, D. E.; Tobin, J. M. & Buchanan, J. D.
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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Tensile Properties of Pyrolytic Graphite to 5000 F

Description: Tensile properties of pyrolytic graphite were measured parallel to the basal planes from room temperature to 5000 deg F. The gage section of the test specimen was 0.06 by 0.10 in. in cross section and3/4-in. long. The specimens were heated in a helium atmosphere by an external graphite heater and were tested at a strain rate of approx 2 x 10/sup -//sup 4/ in./in./sec. Tensile strengths at room temperatare varied from 6,000 to 19,000 psi with elongations of less thsn 1%. At 3000 deg F the strength and elongation were approximately the same as at room temperature. At 4000 deg F there was a very slight increase in the strength and elongation. At 4500 deg F tensile strengths to 35,000 psi and elongations up to 3%, and at 5000"F tensile strengths of 64,000 psi and elongations greater than 70% were measured. At 4500 deg F and above load-deformation curves were recorded. Microstructure and x-ray diffraction patterns showed that marked structural changes accompsny deformation at 5000 deg F. Large changes in room-temperature dimensions, parallel and perpendicular to the basal planes, were measured after heating, with no load, to temperatures in this same range. (auth)
Date: March 10, 1961
Creator: Kotlensky, W. V. & Martens, H. E.
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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The Inhalation Toxicity of Indium Sesquioxide in the Rat

Description: Albino rats were exposed to a In/sub 2/O/sub 3/ dust aerosol of mean concentration 64 mg/m) for 3 months. The retention of In in the tissues and the mobilization of In from the lungs and tracheobronchlal lymph nodes were determined. The results indicate chemical toxicity. (D.L.C.)
Date: February 10, 1961
Creator: Leach, L. J.; Scott, J. K.; Armstrong, R. D.; Steadman, L. T. & Maynard, E. A.
Partner: UNT Libraries Government Documents Department
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