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630A Maritime Nuclear Steam Generator: Status Report Number 1

Description: From foreword: The primary purpose of this document is to set forth the current status of the 630A Nuclear Steam Generator, under development for the U.S. AEC.
Date: September 12, 1963
Creator: General Electric Company. Nuclear Materials and Propulsion Operation.
Partner: UNT Libraries Government Documents Department
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ANTIPROTON-NUCLEON CROSS SECTIONS FROM 0.5 TO 1.0 Bev

Description: Antiproton-production and nucleon-interaction cross sections were investigated for antiprotons in the energy range 0.5 to 1.0 Bev. The antiprotons were distinguished from other particles produced at the Bevatron by a system of scintillation- and velocity-selecting Cherenkov counters. The excitation function and momentum distribution were recorded for antiproton production in carbon and compared with statistical model expectations. The antiprotons were directed by a system of bending and focusing magnets to a liquid hydrogen target. An array of plastic scintillation counters, which almost completely surrounded the hydrogen target, was used to determine the p-p total, elastic, inelastic, and charge-exchange cross sections. Near 500 Mev the total d-p cross section was about 120 mb, and it slowly decreased to 100 mb near 1 Bev. The inelastic cross section, which is principally due to the annihilation process, represented nearly 2/3 of the total cross section. The elastic scattering distribution was highly peaked in the forward direction and could be fitted by an optical model. The total and partial cross sections were also determined for the collisions of antiprotons with deuterons. The p-d total and inelastic cross sections were found to be approximately 1.8 times the p-p cross sections. Corrections were made for the shielding of nucleons within the deuteron in order to ascertaln the p-n interaction. The results indicate that the p-p and p-n cross sections are very nearly equal in this energy region, and that they satisfy the inequalities required by charge independence. (auth)
Date: December 12, 1961
Creator: Elioff, T.; Agnew, L.; Chamberlain, O.; Steiner, H.M.; Wiegand, C. & Ypsilantis, T.
Partner: UNT Libraries Government Documents Department
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Automatic Temperature Control of Irradiation Capsules by a Variable Binary Gas Mixture

Description: Temperature control was achieved by varying the gas mixture, and therefore the thermal conductivity, in a gas annulus surrounding the irradiation capsule. Control systems were used for over one year and maintained capsule temperature to e due to uraniu 10 l F. The system may be used equally satisfactorily with either fueled or nonfueled capsules. The reliability of the system was extremely high, and all maintenance was limited to the readily accessible instrumentation. (auth)
Date: September 12, 1962
Creator: Drescher, R. C. & Johnson, D. E.
Partner: UNT Libraries Government Documents Department
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B-Plant fission product flowsheets. Part 1

Description: B-Plant is currently being evaluated for use as an integrated fission product plant operating in conjunction with the Purex Plant and a waste calcination system. If the forecasted demands for fission products should increase to rates exceeding present capabilities and if private enterprise continues to remain outside the recovery field, present budget plans are to develop the use of B-Plant in three phases. In Phase 1, the B-Plant canyon would be activated and provisions made for preparing and storing fission product concentrates. In Phase 2, additional equipment would be installed to provide a single-line demonstration system for purifying and packaging fission products. In Phase 3, the plant would be converted to a double-line production system for recovering, segregating and storing, purifying and packaging fission products. The purpose of this document is to present the technical bases for B-Plant project scoping studies, including: Design flowsheets for the preparation and storage of fission product concentrates in the scope design of Phase 1 activities; and conceptual flowsheets for the purification of stored concentrates in the engineering studies of Phase 2 activities.
Date: January 12, 1961
Creator: Beard, S. J. & Judson, B. F.
Partner: UNT Libraries Government Documents Department
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BEVATRON OPERATION AND DEVELOPMENT. XXXIII. Period Covered February- April 1962

Description: Experimental work consisted of one new run started and completed this quarter, and the completion of one of the three continuing runs. Of the scheduled operating time, the beam was on for 69.4% of the time, 2.3% of the time was used for experimental setup, and equipment outage took 29.3% of the time. There were two scheduled and two impromptu shutdowns. During one of the scheduled shutdowns the external-beam extraction magnets were installed in the east and south tangent tanks. The other scheduled shutdown was to readjust the Bevatron magnet elevation to correct for foundation subsidence. Internal magnets were also installed. In the new linac development program the ion source was run at 480 kv with a beam current of 100 ma. The linac tank was partially deplated to provide a clean copper surface, and welds and holes were plated with copper. The r-f losses were thereby reduced 20%. (auth)
Date: February 12, 1963
Creator: Crebbin, K.C.; Wenzel, W.A.; Lothrop, F.H.G. & Johnson, R.M.
Partner: UNT Libraries Government Documents Department
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The Bonding of Molybdenum-and Niobium-Clad Fuel Elements

Description: A solid-state bonding technique involving the use of gas pressure at elevated temperatures was utilized for the self-bonding of molybdenum and niobium. Bonding conditions and surface preparation as a function of the integrity of the bond achieved were evaluated for each material. Optimum self-bonding of niobium was achieved by bonding parameters of 2100 to 2300 deg F at 10,000 psi for 3 hr with surfaces which had been prepared by etching in a nitrichydrofluoric acid solution prior to bonding. The process as developed was used to prepare niobium- clad flat-plate- and rod-type fuel elements and flat-plate subassemblies. Niobium tubing was also fabricated by this technique. (Molyb denum self-bonding was most readily achieved by gaspressure bonding at temperatures of 2300 to 2600 deg F at 10,000 psi for periods of 3 hr. With these bonding conditions a number of different surface preparations were satisfactory. Directional ductility of the molybdenum was encountered after bonding and methods to eliminate this were evaluated. Cross rolling with respect to the original rolling direction was shown to improve the ductility of molybdenum-clad specimens. (auth)
Date: July 12, 1960
Creator: Paprocki, S. J.; Hodge, E. S. & Gripshover, P. J.
Partner: UNT Libraries Government Documents Department
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C Reactor fuel failures tubes 2562 and 3361

Description: C Reactor experienced two severe fuel failures, one in process tube 2562 on 10-27-67, and one in process tube 3361 on 12-27-67, both resulting in major efforts to effect their removal. The intent of this document is to present a summary of the action taken for the removal of the failed elements.
Date: April 12, 1968
Creator: Marx, E. R.
Partner: UNT Libraries Government Documents Department
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Calculated Equilibrium Distributions for the Uranyl Nitrate - Tributyl Phosphate - Dilute Nitric Acid System for Temperatures Between 25 and 75 C

Description: Report discussing the "equilibrium uranium distribution between an aqueous nitric acid solution and a 30 per cent by volume solution of tributyl phosphate in a hydrocarbon diluent" (p. 2). This includes the necessary equations.
Date: August 12, 1960
Creator: Wilburn, N. P.
Partner: UNT Libraries Government Documents Department
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Calibration of Liquid-in-Glass Thermometers

Description: Report discussing important elements of thermometer design. Factors affecting the use of common types of liquid-in-glass thermometers are included together with tables of tolerances and reasonably attainable accuracies. The calculation of corrections for the temperature of the emergent stem is given in detail for various types of thermometers and conditions of use.
Date: February 12, 1965
Creator: Swindells, James F.
Partner: UNT Libraries Government Documents Department
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Carbon Transport and Corrosion in High-Temperature Gas-Cooled Reactors

Description: It is noted that in high-temperature He-cooled graphite reactors, sufficiently high levels of gaseous impurities can lead to transport and corrosion effects. The possible effects of these reactions in graphite-moderated reactors designed to operate at a He-coolant pressure of a about 20 atm. were investigated. Results are included on C transport, steam-graphite reactions, and deposition of C on surfaces. (J.R.D.)
Date: April 12, 1962
Creator: Zumwalt, L. R.; Burnette, R. D. & Riedinger, A. B.
Partner: UNT Libraries Government Documents Department
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Chemical Technology Division, Chemical Development Section B Monthly Progress Report, June-July 1960

Description: The effect of two neutron poisons, baron and cadmium, on the rate of dissolution of high-density 95% ThO/sub 2/-5% UO/sub 2/ pellets in the Zirflex Process was determined. Dissolution of U-10% Mo alloy in boiling HNO/sub 3/ resulted in a precipitation of uranyl molybdates. Air caused greater uranium and thorium losses during decladding of ThO/sub 2/-UO/sub 2/ fuel than irradiation. Processing of U-Mo fuel by a Zircex type process is discussed. Two leaches of graphitized fuel with 90% HNO/sub 3/ recovered more than 99% of the uranium. Irradiation of synthetic ThO/sub 2/-UO/sub 2/ fuel solution to 5 and 10 watt-hr/l in a Co/sup 60/ source resulted in about a 50% decrease in decontamination factor using the acid-Thorex flowsheet. Corrosion of titanium, tantalum, and Ni-o-nel in Thorex solution and titanium corrosion in various molybdenum core alloy solutions were investigated. The solubilities of ferric mono- and dibutyl phosphates in HNO/sub 3/ and 30% TBP-Amsco-HNO/sub 3/ solutions were determined. Fission product concentrations expected in Purex waste from processing Yankee Atomic Reactor fuel were calculated. Chemical applications of nuclear explosions to H/sup 3/ exchange, reduction of CaSO/sub 4/, and Gnome sampling are discussed. (For preceding period see CF-60-6-108.) (M.C.G.)
Date: December 12, 1960
Creator: Blanco, R E
Partner: UNT Libraries Government Documents Department
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Chemical Technology Division, Chemical Development Section C Progress Report for April-May 1960

Description: An economical process was successfully demonstrated in bench-scale continuous equipment for stripping U from amines with ammonium carbonate solution. A continuous countercurrent mixer-settler extraction system was set up for further testing of the process for recovery of Te, Np, and U by tertiary amine extraction from UF/sub 6/ transfer cylinder was solutions. The effect of Purex aqueous feed adjustment procedures on Pu extraction by 1 M di-secbutyl phenylphosphonate (DSBPP) was studied. Work was continued on plutonium(IV) nitrate extraction with TBP and phenylphosphonate esters. The response of Ru/sup 106/ extraction to variations in the treatment of TBP-Amsco 125-82 solvent was tested. Two solvents have shown ability to extract cesium. (For preceding period see CF-80-3-136.) (W.L.H.)
Date: July 12, 1960
Creator: Brown, K B
Partner: UNT Libraries Government Documents Department
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Clad thickness variation N-Reactor fuel elements

Description: The current specifications for the cladding on {open_quotes}N{close_quotes} fuels were established early in the course of process development and were predicted on several basic considerations. Among these were: (a) a desire to provide an adequate safety factor in cladding thickness to insure against corrosion penetration and rupture from uranium swelling stresses; (b) an apprehension that the striations in the zircaloy cladding of the U/zircaloy interface and on the exterior surface might serve as stress-raisers, leading to untimely failures of the jacket; and (c) then existing process capability - the need to maintain a specified ratio between zircaloy and uranium in the billet assembly to effect satisfactory coextrusion. It now appears appropriate to review these specifications in an effort to determine whether some of them may be revised, with attendant gains in economy and/or operating smoothness.
Date: May 12, 1966
Creator: Smith, E. A.
Partner: UNT Libraries Government Documents Department
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Compact thermoelectric converter. Quarterly progress report, July 1, 1968- -September 30, 1968. Phase II-C

Description: Declassified 30 Aug 1973. An initial analysis of experimental heatup data was compiled to determine the success of module experiments designed to triple the voltage power ratio of standard TEM-9 modules. Accomplishment of this design improvement is extremely significant in the development of tabular module systems having lower power applications. With the increased voltage to power ratio, power conditioning devices will not be required to step up the module output voltage to a more usable level. Heatup data from TEM9AE S/N-1, employing 0.030-inch lead telluride washers and 0.0015-inch mica insulators, correlate very closely with performance calculations. A study was initiated to determine an optimum design for a module to be internally fueled using cobalt60. In addition, alterations to the calculation model TEMOD were made to handle the effects of gamma heating within the lead telluride washers. An experimental test program was defined which would verify the validity of the mathematical model. Additional analytical work was accomplished to correlate predicted axial heat transport rates of heat pipes with experimental data. A study of sodium heat pipe performance data revealed that a sonic vapor velocity can occur in the region between evaporator and condensor sections of the heat pipe to produce a limit to the axial heat transfer rate. Evaluation of niobium contacts show them to be considerably more compatible with p-type lead telluride than iron contacts. However, the niobium does not yield absolute electrical stability and, as a result, no further work beyond the present analysis is contemplated. Six modules containing solid tungsten conductors were placed on test at elevated temperatures. Although degradation rates measured were the lowest yet recorded for modules employing TEGS-2P material, for temperatures above 1100 deg F the most stable operation is achieved through the substitution of TEGS-3P.material for TEGS-2P material. (auth)
Date: October 12, 1968
Partner: UNT Libraries Government Documents Department
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Comparative nuclear effects of biomedical interest. Civil effects study

Description: Selected physical and biological data bearing upon the environmental variations created by nuclear explosions are presented in simplified form. Emphasis is placed upon the ``early`` consequences of exposure to blast, thermal radiation, and ionizing radiation to elucidate the comparative ranges of the major effects as they vary with explosive yield and as they contribute to the total hazard to man. A section containing brief definitions of the terminology employed is followed by a section that utilizes text and tabular material to set forth events that follow nuclear explosions and the varied responses of exposed physical and biological materials. Finally, selected quantitative weapons-effects data in graphic and tabular form are presented over a wide range of explosive yields to show the relative distances from Ground Zero affected by significant levels of blast overpressures, thermal fluxes, and initial and residual penetrating ionizing radiations. However, only the ``early`` rather than the ``late`` effects of the latter are considered.
Date: January 12, 1961
Creator: White, C.S.; Bowen, I.G.; Richmond, D.R. & Corsbie, R.L.
Partner: UNT Libraries Government Documents Department
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Consistent Properties of Composite Formation Under a Binary Relation

Description: In this paper we study the properties of a relation, H, between two sets, X and Y, H ⊆ X x Y. We wish to determine when the existence of H between some elements of X and Y implies the existence of H between other composite elements of X and Y. The aim is to characterize a relation by its extension or restriction to composite elements.
Date: August 12, 1969
Creator: Schwebel, John C & McCormick, Bruce H
Partner: UNT Libraries Government Documents Department
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COOLING OF THE HFIR BERYLLIUM REFLECTOR FOLLOWING A REACTOR SCRAM OR AN ELECTRICAL POWER OUTAGE

Description: Thermal stresses in the HFIR beryllium reflector were computed for the unlikely case where the reactor is scrammed with a simultaneous loss of coolant flow and for the case following an electrical power outage where the reactor power level and the coolant flow rate are reduced simultaneously. For the case where the reactor is scrammed with a sudden loss of the coolant flow, the resulting maximum tensile thermal stress following the scram is 22,500 psi. In case of an electrical power outage, the maximum tensile thermal stress following a reduction of the fission power level from 100 Mw to 10 Mw with the lowering of the coolant flow rate to 10% of the normal value is 12,800 psi. (auth)
Date: December 12, 1961
Creator: McLain, H. A.
Partner: UNT Libraries Government Documents Department
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Corrosion of Zircaloy-2 by pH 10 LiOH in heated crevices

Description: Both the inner and outer tubes of the N-Reactor fuel elements will have self supports spot welded to the lateral heat-transfer surface of the element. A crevice a few mils thick will exist around the weld between the support tab and the cladding. Because of the heat flux through the cladding at this point and the insulating effect of the support tab, the temperature in this crevice will be higher than that on the free surface away from the support. This can result in boiling in the crevice leading to concentration of LiOH (or impurities in the water) to a level where it can cause severe corrosion of the Zircaloy-2 cladding. The tests described in this report were conducted to determine whether such attack might be encountered in N-Reactor.
Date: November 12, 1963
Creator: Dickinson, D. R.
Partner: UNT Libraries Government Documents Department
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DESALINATION OF SEA WATER

Description: No Description Available.
Date: April 12, 1965
Creator: Harty, H.
Partner: UNT Libraries Government Documents Department
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Design and Test of a Diaphragm Pump for Liquid Metals

Description: Technical report. From Abstract : "Details are provided on the construction and operation of a two-stage diaphragm pump successfully used for the first time in liquid metal service. From the results of 5,376 hr. test of the pump it was concluded that it is well suited to the pumping of liquid metals at low flow rates where pulsating flow can be tolerated and also where remote operation is required. Operating temperatures and pressures are limited only by the availability of suitable materials of construction."
Date: April 12, 1963
Creator: Westerheide, D. E. & Clifford, J. C.
Partner: UNT Libraries Government Documents Department
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Design criteria for coolant backup three remaining smaller reactors

Description: This document defines the objectives, bases, and functional requirements that shall govern the preparation of design of the coolant backup system for three remaining smaller reactors. This project will increase the reliability of the coolant backup facility at B Area by providing an independent last-ditch coolant system to B and C Reactors. The reliability of the last-ditch system for D Reactor will also be improved in that the present F and H leg of the export system will no longer be a part of the new export system.
Date: August 12, 1964
Creator: Brinkman, L. B.
Partner: UNT Libraries Government Documents Department
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