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Dissolution of aluminium oxide as a regulating factor in aqueous aluminum corrosion

Description: The solubility of aluminum corrosion product in contact with metallic aluminum in deionized water has been determined over the range 80 to 350/sup 0/C. Evidence is presented to show that oxide dissolution results in the formation of a porous oxide on aluminum exposed in refreshed dynamic systems. Dynamic corrosion rate data have been analyzed on the basis of parabolic film growth and a linear oxide degradation process acting simultaneously on the system. The degradation process has been shown to be a function of refreshment rate.
Date: August 31, 1959
Creator: Dillon, R.L.
Partner: UNT Libraries Government Documents Department
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Production test IP-262-A-11-FP, evaluation of projection fuel elements for use in ribbed process tubes: Demonstration loadings

Description: The objective of this test is to demonstrate the feasibility of projection fuel elements for use In existing process tubing and to determine the reduction in rupture rates or hot-spot incidence so achieved. This test is to authorize, (a) charging 20 columns of bumper type fuel elements and 20 columns of control elements per reactor into B, D, DR, F, and R Reactors for irradiation up to 1200 MWD/T exposure, and (b) irradiation of four columns each of enriched (0.947%) bumper and enriched (0.947%) normal type fuel elements until two ruptures are sustained in each (or until one group shows a significant improvement).
Date: December 31, 1959
Creator: Hall, R. E.
Partner: UNT Libraries Government Documents Department
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Production test authorization IP-217-A-supplement, uranium sample irradiation (HAPO-218)

Description: The object of this production test is to authorize the irradiation of samples of natural uranium for use in the fission product volatilization program. Any empty central zone PCCF and/or process tube at B, D, DR, or H Reactors will meet the requirements of this authorization.
Date: August 31, 1959
Creator: Zima, G. E.
Partner: UNT Libraries Government Documents Department
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The Hard-Core Pinch. I

Description: This report analyzes a linear, hard-core pinch tube built to examine tubes afflicted by small-scale instabilities evident from many observations.
Date: July 31, 1959
Creator: Birdsall, D. H.; Colgate, Stirling A. & Furth, Harold P.
Partner: UNT Libraries Government Documents Department
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The Hard-Core Pinch. II

Description: This report describes a toroidal version of the hard-core pinch and the additional information it can obtain to what the analogous linear pinch can get.
Date: July 31, 1959
Creator: Birdsall, D. H.; Colgate, Stirling A. & Furth, Harold P.
Partner: UNT Libraries Government Documents Department
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SER Temperature Coefficient

Description: Experimentally determine the overall isothermal temperature coefficient of the SER up to the design operating temperatures.
Date: December 31, 1959
Creator: Johnson, J. L.
Partner: UNT Libraries Government Documents Department
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Stored Energy: Growth and Annealing Status of Graphite Moderator in the BNL Research Reactor. Final Report

Description: The present sthtus, past annealing procedures and experiences, future annealing procedures, annealing sehedule, revised annealing procedure (1958), procedure for combating a graphite fire in fuel channel, high-temperature stored energy, and graphite burning experiments are reportcd for the BNL Research Reactor. The following subjccts are discussed in the appendixes: control of radiation damage in a graphitc reactor; annealing of graphite moderator structure in the BNL; annealing operation in BNL graphite reactor; effect of pile radiation on mechanical and other properties of graphite; neutron sensing instrumentation; instrumentation for sensing fuel failures; thermocouple pattern for enriched fuel loading; environmental hazard from a molten fuel element; retention of volatile flssion products on filters; retention of volatile fission products on water tube coolers; retention of volatile fission products in molten fuel plates; and release of the lowtemperature stored energy in the BEPO Pile. (W.L.H.)
Date: October 31, 1959
Partner: UNT Libraries Government Documents Department
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URANIUM ALLOY POWDERS BY DIRECT REDUCTION OF OXIDES

Description: A process is outlined for the production of uranium alloy powders by co- reduction of mintures of uranium oxide and alloy element oxides. The reduction of mechanical mintures of the oxides of uranium and alloy element with calcium in a sealed reaction vessel is shown to produce powder wtth a variation in particle composition, although of consistert composition over various size fractions. The particular alloy systems which are considered are uranium--nickel, uranium-- chromium, uranium --molybdenum, and uranium--niobium. The uranium-molybdenum and uranium--niobium powders are single phase (metastable gamma), which is of consequence in the production of dimensionaHy stable nuclear fuels. Potential applications of some of these alloys are discussed. (auth)
Date: October 31, 1959
Creator: Myers, R.H. & Robins, R.G.
Partner: UNT Libraries Government Documents Department
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Fuel Elements Conference Held at Gatlinburg, Tennessee, May 14-16, 1958

Description: The fuel element conference provided a favorable medium for presentation and discussion of recent developments in the field of solid fuel elements. The conference was designed to replace the more general Metallurgy Information Meetings held annually in the past. The scope of the meeting embraced the design fabrication, performance, and material problems of fuel elements. (W.D.M)
Date: October 31, 1959
Partner: UNT Libraries Government Documents Department
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PROCEEDINGS OF THE FRENCH-AMERICAN CONFERENCE ON GRAPHITE REACTORS, HELD AT BROOKHAVEN NATIONAL LABORATORY , NOVEMBER 12 TO 15, 1957

Description: Twenty-eight of the thirty-five papers presented at the conference are included with discussions. Abstracts of the remaining seven papers are given. The seven sessions were devoted to: radiation effects on graphite, nuclear properties of graphitc, graphite lattice reactivities, chemistry of graphite, chemical reactions between liquid Na and Zr, slug canning fer the ORNL Graphite Reactor, and critical assemblies. Separate abstracts have been prepared for each of the twentyeight papers. (T.R.H.)
Date: October 31, 1959
Partner: UNT Libraries Government Documents Department
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Quarterly Progress Report for the Period July 1 to September 30, 1958

Description: An evaluation and the resulting conclusions of work performed are given for each project in which definitive progress was made. The principle progress consisted of: obtaining promising fuel fabrication results from production scale investigations of simplified UO/sub 2/ preparations methods, presintering operation elimination, increasing pellet densities and reductions of UO/sub 2/ losses; successfully brazing a ninety-five inch long experimental fuel subassembly in which all joints, including outside ferrules and control rod rubbing strips, had brazed fillets; completing the fabrication of the last group of MTR process water irradiation samples and all of the in-pile test loop samples except for the specimens containing 27% enriched pellets; performing a nuclear analysis of various fuel assembly designs as part of an over-all evaluation of fuel assembly bowing in the Yankee reactor; completing corrosion studies of primary plant materials in static autoclaves to aid in the selection of a pH control agent; developing a Compromise Design'' for the fueh assembly which incorporates a series of small design changes to eliminate the possibility of restricting control rod motion by interference due to a superposition of bowing of the fuel assembly and an adverse accummulation of mechanical tolerances; completing the calculation of the moderator temperature coefficient, the Doppler temperature coefficient, the void coefficient, neutron lifetime, and delayed neutron fraction; obtaining flux profiles, flux peaking, flux spectrum, peripheral fuel rod worth, void coefficient, temperature coefficient, and control rod worth data from the Critical Reactor Experiments with a 3: 1 water-to-uranium metal ratio core at the Westinghouse Reactor Evaluation Center; continuing instrumentation and individual component operational testing of the in-pile test loop; and analyzing the results of the post-irradiation examination of the second group of process water samples and delivery of the last group of samples to the MTR. (For preceding period see YAEC-87.) (auth)
Date: October 31, 1959
Creator: Garbe, R. W. & Walchli, H. E.
Partner: UNT Libraries Government Documents Department
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Sodium Reactor Experiment (Sre) Shielding Evaluation for Thermal Neutron Streaming at Reactor Vessel Coolant Pipe Penetrations

Description: The experimental program performed in the SRE auxiliary and main primary galleries was part of a program to determine the adequacy of the shielding configuration for the SRE. The work discussed in this report is concerned with analysis of neutron streaming at coolant pipe penetrations of the reactor vessel, analysis of the shielding required, testing and evaluation of recommended shielding, and measurement and correlation of neutron streaming in labyrinths with theory. The activation analysis method using zinc sheets which was developed for the program of determining thermal neutron streaming in the SRE primary galleries was proven to be versatile, accurate, and reliable. A modified form of the theoretical method of Price, Horton, and Spinney, used to determine neutron scattring through labyrinths, was found to agree favorably with the experimental results obtained from the SRE primary galleries. The theoretical attenuation method used no determine the neutron shield configuration installed in the auxiliary primary gallery was found to give an overestimate of the actual attenuation properties of this shield. The neutron shield configuration installed in the auxiliary primary gallery proved to be adequate in reducing the thermal neutron streaming flux to an acceptable level. It is concluded that both SRE primary galleries are now adequately shielded to prevent excessive neutron- induced activation of the components and equipment located therein. (auth)
Date: October 31, 1959
Creator: Anderson, F. D.
Partner: UNT Libraries Government Documents Department
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Development of Uranium Nitride-Stainless Steel Dispersion Fuel Elements

Description: In research in support of the GCRE, procedures for the fabrication of stainless steel-clad flat-plate fuel elements having a core of 28 w/o UN dispersed in an iron-18 w/o chromium--14 w/o nickel--2.5 w/o molybdenum matrix were developed. The preparation of UN and its compatibility with the components of the matrix alloy were studied. The UN for the program was prepared by nitriding uranium metal at 850 C and then dissociating the U/sub 2/N/sub 3/ produced to UN in a vacuum at 1300 C. In compatibility studies, UN reacted with nickel alone at 1800 F, but no reaction with nickel was observed when the other matrix components were also present. The effects of fabricational variables were evaluated by metallographic examination, longitudinal and transverse tensile tests, bend tests, and corrosion tests. Studies indicated that minus 200 plus 325-mesh UN dispersed in a minus 325-mesh elemental-powder matrix rolled green vith a 30 per cent initial reduction at 2100 F and then annealed 3 hr at 2300 F produced the best fuel core. (auth)
Date: July 31, 1959
Creator: Paprocki, S. J.; Keller, D. L.; Cunningham, G. W. & Foulds Jr., A. K.
Partner: UNT Libraries Government Documents Department
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Monthly Progress Report for the Period September 1 to 30, 1958

Description: A 95-inch experimental Yankee sub-assembly was successfully brazed at 1890 plus or minus 10 for three hours then furnace cooled. All of the joints, outside ferrules, and control rod rubbing strips showed a brazed flllet. Dimensional surveys and joint strength studies indicate that two brazing techniques appear feasible for subassemblies. Codings for computers for use in core design problems are described. A series of criticality calculations for the three water-to-metal volume ratios of the Yankee critical experiments was concluded with the calculatlon of the critical mass of a 4:1 water-touranium metal volume ratio core. Calculations were performed to determine the Mwd/t burnup in terms of unperturbed nvt for 2.7 and 5.4% enriched process water test specimens. Temperature stability experiments were conducted on Rohm and Haas XE- 150 resin by means of water baths at the puriflcatlon system operating temperature of 14O F and at an elevated temperature of 170 F. Data were obtained on corrosion of construction materials. A study was made of the reaction of soluble oxygen with hydrazine in borated water. Performance of criticality experiments on stainless steel clad UO/sub 2/ fuel elements at various water-to-metal ratios are reported. (For preceding period see YAEC95.) (W.D.M.)
Date: October 31, 1959
Creator: Garbe, R. W. & Walchli, H. E.
Partner: UNT Libraries Government Documents Department
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APPLIED HEALTH PHYSICS SEMI-ANNUAL REPORT FOR JULY 1956-DECEMBER 1956

Description: Routine monitoring procedures are summarized. Based on samples collected by continuous air monitors, the wcekly average value for air contamiration in the laboratory area was 1.98 x 10/sup -12/ mu c/cc. Radioactive fall-out was determined by measuring the total activity and total number of particles collected on gum paper fall-out trays. Two peak periods of activity in July and September were concluded to be the result of weapons tests. Meteorological data are tabulated. Data are included from a survey of natural radioactive elements in TVA lake waters, activity levels in samples of river and lake waters in the laboratory drainage area, rain water, and laundry decontamination measuremcnts. The combination film dosimeter-personnel identification badge was redesigned. A drawing of the badge and its component parts is included. Design modifications are described which wcre made in an automatic air sampler rized from radiation surveys. (C.H.)
Date: October 31, 1959
Partner: UNT Libraries Government Documents Department
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ELK RIVER REACTOR. Quarterly Project Report for December 1958 January- February 1959

Description: Design and construction are proceeding satisfactorily. Fuel capsule irradiation studies are progressing, and preparation of specifications for fuel elements and assemblies is almost complete. Other design, development, and construction progress is reported. Instrumentation developments are described. Reactor physics studies are reported. (For preceding period see AECU-4013.) (T.R.H.)
Date: October 31, 1959
Partner: UNT Libraries Government Documents Department
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HIGH TEMPERATURE PROPERTIES OF URANIUM AND ITS ALLOYS

Description: Data are presented on the mechanical and physical properties of high- purity and normal uranium and for a range of dilute uranium alloys at temperatures up to 700 deg C. The results of thermal conductivity and thermal expansion determinations for a range of alloys are discussed, and some results on the latent heat of fusion and volume change on melting are presented. Results of dilute alloy creep tests are reported and correlated with thermal cycling data and resistence to swelling under irradiation. (J.R.D.)
Date: October 31, 1959
Creator: McIntosh, A.B. & Heal, T.J.
Partner: UNT Libraries Government Documents Department
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VARIABLE MODERATOR REACTOR DEVELOPMENT PROGRAM. Quarterly Progress Report No. 1

Description: Development of the boiling water UO/sub 2/ fueled Variable Moderator Reactor (VMR) is conducted under contract for the USAEC. The initiation and progress of work under Phase I of the contract, Physics and Kinetic Analysis and Initial Evaluation,'' and the preparation for Phase II, Critical Experiment and Analysis of Results,'' are reported. A hydrodynamic flow sheet representing the sequence of calculations for the BOCH program was prepared. A preliminary block diagram of the kinetics model of the VMR was prepared. Work is reported on the PUREE code which is designed to give an accurate representation of the physics of the VMR core. A fuel element fabrication speciftcation was prepared and released for quotations. A study was made to select the most appropriate material for void simulation throughout the range of interest in the VMR. (W.D.M.)
Date: August 31, 1959
Partner: UNT Libraries Government Documents Department
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NUCLEAR BATTERY--THERMOCOUPLE TYPE. Quarterly Progress Report No. 8 for October 1, 1958 to December 31, 1958

Description: A demonstration battery containing an 880-curie polonium-2l0 heat source was constructed and tested. The battery initially had an open circuit voltage of 23.6 volts, a maximum output of 161 milliwatts, and an overall efficiency of 0.57%. Details for the construction of the battery generator are given. (For preceding period see MLM-CF-58-4-59.) (auth)
Date: October 31, 1959
Creator: Blanke, B.C.
Partner: UNT Libraries Government Documents Department
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Quarterly Report for October, November, and December 1957

Description: 3 9 8 4 3 3 3 6 9 6 5 5 have been started on possible urania-ceramic fuels with magnesia, alumina, zirconia, magnesia-- zirconia and calcia--zirconia. Mixtures of 45 magnesia--45 zirconia and 9 magnesia--81 zirconia by weight appeared to give good preliminary promise. Work on ThO/sub 2/--10 wt.% UO/sub 2/ has centered about improvement in density of pellets and elimination of cracking and open porosity. Efforts to increase the percentage of urania in thoria--urania have met with considerable success up to 70 wt.% urania in air-firing tests. Some attempts have been made to prepare UC by reaction of UO/sub 2/ with carbon, but so far an incomplete reaction has resulted. As to metal fuels, a sample of U--3 wt.% Ti--1.5 wt.% Nb has shown good corrosion resistance in 245 deg C water, slightly better than the EBWR fuel alloy. Fast Power Breeder Program. An improvement of the average yield of EBR- IItype fuel pin injection castings has been attained, reaching over 90% as compared with yields of 50 to 60% in the recent past. The uranium-plutonium- fissium alloys proposed for the fuel of EBR-II are being given further examination. The thermal conductivity of U-- 20 wt.% Pu--10 wt.% Fs alloy was determined at 0.053 (cal/seccm- deg C) at 600 deg C and thermal expansion at about 16.7 10/sup -6/ per deg C in the range 20 to 510 deg C. Reactor Systems Research and Development. A calibration test of the ultrasonic inspection technique was made on known defect Croloy tubing to check its effectiveness in the examination of fuel jacketing material. Development is continuing on a beta ray method for measuring clad thickness and on the scintillation counting method of measuring uranium content of fuel plates. An improved design of probe coil for eddy current measurement of cladding thickness …
Date: October 31, 1959
Partner: UNT Libraries Government Documents Department
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The Transformation Behaviour of Some Dilute Uranium Alloys

Description: Metallographic and dilatometric studies of the heat treatment behavior of technical uranium and certain alloys of interest for use in natural-uranium fuelled reactors are described. The micro-structure and secondary phases present in high purity and technical purity uranium are first discussed and the time- temperature-transformation data presented. Results on binary alloys comprising U/ Al, U/Cr, U/Fe, U/Mo, U/Nb, U/Ti, U/V, and U/Zr, containing up to about 2 at.% of the alloying element, are described. Cr, Fe, and Zr were found to be the most effective grain refining agents. Al and V were satisfactory under some conditions whereas Mo, Nb, and Ti were ineffective. The grain-size results together with the time-temperaturetransformation curves indicated that grain refinement is associated with depression of the temperature of transformation on heat treatment, the extent of this depression for a given cooling rate being dependent on the alloy content. In practice it was found that alloys such as U/ Zr and U/Fe responded to water quenching treatments, and optimum refinement of others (e.k., U/Cr) resulted from isothermal treatment at 500 to 600 deg C. (auth)
Date: October 31, 1959
Creator: Jepson, M. D.; Kehoe, R. D.; Nichols, R. W. & Slattery, G. F.
Partner: UNT Libraries Government Documents Department
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Diffusion of Uranium with Various Transitional Metals; DIFFUSION DE L'URANIUM AVEC QUELQUES METAUX DE TRANSITION

Description: The diffusion process in uranium and its alloys was studied from 550 to 1075 deg C with diffusion couples of U with Zr, Mo, Ti, and Nb and with the alloys U--Nb and U--Mo. A brief description is given of the experimental methods. Results relative to the concentration-penetration curves are presented, and the coefficients of diffusion are calculated. The equilibrium diagram was established for the U--Zr system. The results obtained by micrographic examination, microhardness measurements, and autoradiography are compared with each other. The mechanisms of diffusion are investigated by studying the Kirkendall effect and calculating the Darken intrinsic coeffi cients in the U--Zr and U--UMo diffusion couples. (J.S.R.)
Date: October 31, 1959
Creator: Adda, Y. & Philibert, J.
Partner: UNT Libraries Government Documents Department
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