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Chemical Technology Division, Unit Operations Section Monthly Progress Report for August 1959

Description: The concentration gradients of uranyl ion in aqueous and organic solutions were analyzed by taking a macro photograph of the desired gradient by monochromatic (436 m mu ) light transmitted by the solution normal to the gradient in an appropriate diffusion cell. Two Druhm runs were terminated due to malfunction of the sodium metering system and the third run was terminated when the UF/sub 6/ nozzle ruptured. Calculations of particle temperature versus time relations for the flame denitration-calcination method of preparing metallic oxide from nitrate solutions indicate that the times required for heat transfer are controlled by the rate of radiant heat transfer to particle surfaces instead of by conductive heat transfer within the particles. A completed experimental study indicated that electrolysis in a cell with a mercury cathode and a platinum anode is a practical process for removing nickel from HRT fuel solution. The apparent diffusion coefficient of uranium loading on Dowex 21K was shown to be directly related to the resin size. An explosion of sufficient violence to blow apart the Pyrex pipe dissolver occurred during the fifth Darex dissolution of simulated SRE fuel probably from a rapid gas phase reaction between hydrogen and oxidizing gases such as NO/sub 2/. Materials handling flowsheets were completed for (A) decladding, washing, recanning and storing spent SRE uranium fuel slugs and (B) the shearing and leaching of stainless steel clad UO/sub 2/ and UO/sub 2/- ThO/sub 2/ fuels. A literature survey is being conducted dealing with reactor coolant and coolant loop contamination and decontamination. During run R-17 for calcination of evaporated Darex waste, the same as run R-16 which deformed the bottom calcination vessel except that one of the three added pressure probes was vibrated to keep it unplugged, the bottom of the calcination vessel did not deform, and there was …
Date: December 31, 1959
Creator: Bresee, J. C.; Haas, P. A.; Horton, R. W.; Watson, C. D. & Whatley, M. E.
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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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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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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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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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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ELK RIVER REACTOR. Quarterly Project Report for March-April-May 1959

Description: An alternate design of the Elk River Reactor was completed during the quarter. Three points were emphasized in the study: performance capability to provide 116 Mw(th) output; fuel costs of the reference design versus alternate fuel media and cladding; and reactor control modifications to provide sufficient reactivlty to eliminate soluble hold down'' poisons, permit higher power operation and obtain a greater fuel life. Work is summarized on fuel element and material development program, reactor vessel and internal components, control rods and drive mechanism, process systems, superheater, buildings and facilities, instrumentation and site construction. (For preceding period see ACNP-ERR-3.) (W.D.M.)
Date: October 31, 1959
Partner: UNT Libraries Government Documents Department
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AN EVALUATION OF MERCURY COOLED BREEDER REACTORS

Description: Under the New Reactor Concepts Evaluation Program sponsored by the United States Atomic Energy Commission. Advanced Technology Laboratories (a Division of American Radiator & Standard Sanitary Corporation) has undertaken am investigation of the technical feasibility and economic potential of the use of boiling mercury as a coolant for fast breeder reactors The investigation was performed between January 1, 1959, and October 31. 1959. This is the final report on that investigation and is submitted in compliance with the terms of the program authorization, Contract Number AT(04-3)-109, Project Agreement Number 4. (auth)
Date: October 31, 1959
Creator: Bradfute, John O.; Battles, Donald W.; Clark, George S.; Corridan, Robert E.; Gellenbeck, Edward T.; Kavanagh, Devereux L. et al.
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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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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INFORMATION MEETING ON GAS-COOLED POWER REACTORS, OAK RIDGE NATIONAL LABORATORY, OCTOBER 21-22, 1958

Description: This meeting is one of a series of Civilian Power Reactor Conferences and was held colncident with an AEC invitation to industry to bid on the construction of a gas-cooled facility. Papers are presented on design studles, hazards, components, costs, materials, and design concepts for specific reactors. (W.D.M.)
Date: October 31, 1959
Partner: UNT Libraries Government Documents Department
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IRRADIATION BEHAVIOR OF ThO$sub 2$-UO$sub 2$ FUELS

Description: ThO/sub 2/-UO/sub 2/ fuels have been studied because of certain advantages possessed by this type of fuel over UO/sub 2/. Pellets of ThO/sub 2/- UO/sub 2/ have been found to be highly stable dimensionally under irradiation. Various experimental assemblies which have been irradiated have included Al, Zircaloy, and stainless steel cladding, with the annulus surrounding the pellet filled with air, He, NaK, or Pb. A full-scale core of pellets lead-bonded to Al tube-plates has been used successfully in the Borax-IV reactor. Preliminary defect tests have shown that solid fission products are retained and that only the gaseous products, Xe/sup 138/ and Kr/sup 88/, are detectable in the reactor coolant system. (auth)
Date: October 31, 1959
Creator: Kittel, J.H.
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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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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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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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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Quarterly Report of the Solution Materials Section for the Period Ending October 31, 1958

Description: A simulated HRT fuel solution was unstable when passed through a heated Zircaloy-2 bypass section installed on a Ti loop. Losses of uranyl, cupric, and nickel sulfates were observed when the temperature of the solution emerging from the bypass was as low as 290 deg C. The 0.04 m UO/sub 2/SO/sub 4/ solutions containing Please delete abstract number 7731
Date: October 31, 1959
Creator: Griess, J. C.; Savage, H. C.; Greeley, R. S.; English, J. L.; Bolt, S. E.; Hess, D. N. et al.
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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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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SUMMER LECTURES ON THEORETICAL PHYSICS

Description: Some lectures on theoretical physics are presented, mostly in the field of elementary particles and field theory, delivered during the summer of 1958 by visitors at the Argonne National Laboratory. The lectures are essentially summaries or reviews of published material. Lectures included are on collision theory for composite particles, quantum theory of fields, inquiries into field theory, classical relativistic theory of elementary particles, and bound states of many-particle systems. (W. D. M.)
Date: October 31, 1959
Creator: Haag, R.; Gupta, S. N.; Dresden, M.; Havas, P. & Coester, F.
Partner: UNT Libraries Government Documents Department
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