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- Annual Report 1961
- This seventh Annual Report is a summary of some of the progress in scientific and engineering research and development carried on at Argonne National Laboratory during 1961. As is customary in this series, only those portions of the total program that have reached such a stage that they may be of general interest are recorded. Thus, a comparison with the Annual Reports for 1959 (ANL-6125) and for 1960 (ANL-6275) will reveal the description of a generally different set of scientific activities. A more detailed presentation of any work covered in this report or of the many ANL projects not mentioned may be obtained by perusing the various progress and topical reports issued by the Laboratory during 1961. A list of the publications in the scientific journals during 1961 by Argonne personnel has been given as an Appendix.
- Operating Manual for the Argonaut Reactor
- The design of the Argonaut (Argonne Nuclear Assembly for University Training) was initiated by the Reactor Engineering Division of Argonne National Laboratory to satisfy needs for a low-power reactor facility within the Laboratory, and for training uses within the international School of Nuclear Science and Engineering (ISNSE). It was intended primarily for instruction and research in reactor physics. It was also considered as a possibility that it would fulfill the requirements of universities engaged in a program of nuclear science. The cost of the facility was to be kept to a minimum consistent with the high degree of inherent safety and a great amount of flexibility in the system. The basic design stemmed from the Knolls Atomic Power Laboratory Thermal Test Reactor* (TTR), now called Nuclear Test Reactor (NTR). Modification during the course of the work justified the new name "Argonaut".
- Chemical Engineering Division Summary Report for January, February, and March 1958
- Development work was continued on the fused fluoride process for the recovery of enriched uranium from zirconium-matrix fuel alloys. The alloy is dissolved by immersing it in molten sodium fluoride-zirconium fluoride at 600°C and passing hydrogen fluoride vapor through the system.The dissolved uranium tetrafluoride in the melt is then volatilized as uranium hexafluoride by sparging with fluorine. The uranium hexafluoride product is purified and decontaminated by fractional distillation. Additional corrosion tests were made on a variety of metals in an effort to find a material of construction suitable for the fluorination step. All the metals tested, with the exception of Hastelloy B, were attacked rapidly in the fluorinated melt. The attack was particularly severe at the melt-gas interface when tests were made with partially submerged specimens of the metals.
- The Fabrication of a Plutonium Helix for a Doppler Experiment
- A helix constructed of plutonium was made to test the Doppler temperature effect in ZPR-III. The helix, 1 inch in diameter and 6-1/4 inches long, contained 240 grams of delta-phase plutonium alloy encapsulated in titanium tubing. Four plutonium rods were extruded, joined together, and pushed into a titanium tube. This tube was swaged tightly over the plutonium rod, and the assembly was wound into a coil. Electrical leads to the coil were made by swaging copper tubing over the ends of the coil. The helix was tested by cycling about 500 times between 50°C and 190°C. The coil was heated with a current of 130 amperes and cooled with a blast of chilled helium. (1) Several helices of uranium(2) were cycled during the same tests. Despite the severity of the thermal cycles, the helices were undamaged.
- A laboratory Ivestigation of the Fluorination of Crude Uranium Tertrafluoride
- Ore concentrates have been converted directly to crude uranium tetrafluoride by hydrogen reduction and hydrofluorination in fluidized-bed reactors. Small-scale laboratory experiments demonstrated that this process can be extended to the production of crude uranium hexafluoride through fluorination of the uranium tetrafluoride in a fluidized bed. The satisfactory temperature range for the reaction lies between 300°C and 600°C. At 450°C the fluorine utilization is between 50 and 80 per cent. With excess fluorine, over 99 per cent of the uranium is volatilized from the solid material. The fluidization characteristics of certain materials are improved by the addition of an inert solid diluent to the bed.
- Chemical Engineering Division Summary Report July, August, and September, 1957
- Development work continued on a fused salt process for the recovery of uranium from zirconium-matrix fuel alloys. The fuel is dissolved in a sodium fluoride-zirconium fluoride melt at 600°C by hydrogen fluoride sparging. The melt is then sparged with fluorine gas which volatilizes the dissolved uranium as the hexafluoride. The final decontamination and purification of the uranium hexafluoride are accomplished by fractional distillation. The testing of graphite as a container material for the hydrofluorination step was continued. Additional thermal cycling experiments were performed, using a helium sparge in equimolar sodium fluoride-zirconium fluoride melt at 600°C. The extent of penetration of the fused salt into the graphite was determined. No mechanical degradation was present. Dimensional change data were also obtained for graphite vessels in which the fused salt was sparged with hydrogen fluoride.
- Chemical Engineering Division Summary Report for January, February, and March 1957
- A fused fluoride process for dissolution of zirconium-uranium fuel alloys is being developed. The alloy is dissolved in an equimolar sodium fluoride-zirconium fluoride melt at 600°C by sparging the system with hydrogen fluoride. The uranium is volatilized from the melt as the hexafluoride by a sparging operation with fluorine or bromine pentafluoride vapor. This product is then decontaminated and purified by fractional distillation.
- Quarterly Report October, November and December, 1956
- Methods of producing extremely clean surfaces on rolled Zircaloy-2 strip have been investigated. It has been found that the finer abrasives, 400 mesh or finer, are more effective than coarse types because of their ability to penetrate pits and crevices more readily. Two such cleanings, with an intermediate 35 v/o HNO3-5 v/o HF pickle, resulted in a microscopically clean surface. Ultrasonic inspection of the EBWR fuel plates has been completed during this quarter. Approximately 95% of the plates were found acceptable. All subassemblies manufactured from the EBWR plates met dimensional specifications and passed 9-day corrosion tests at 290°C (550°F). All thoria-urania pellets for the loading of Borax-IV have been pressed, loaded into tube plates, and fabricated into subassemblies. The total number of subassemblies made was 82, of which 72 were fuel plates and 10 were blanket plates, more than sufficient for the loading. The reactor has gone critical using this loading.
- A Coated Cast Iron Crucible for use with Eutectic Al-Si Alloy in the Temperature Range 595°-650°C
- The feasibility of the coated metal crucible as a container for eutectic Al-Si alloy has been proven by test. Small, enamel-coated cast iron pots has been proven by test. Small, enamel-coated cast iron pots have successfully withstood the chemically aggressive Al-Si alloy and the adverse influence of an oxidizing atmosphere for a period of 3 months at 725°C. A similarly coated castiron crucible containing 450 pounds of eutectic Al-Si alloy was successfully tested for 144 days in a jacketing operation conducted at 595°-650°C. Under the same conditions, the normal service life of clay-bonded graphite and silicon carbide crucibles rarely exceeds 45 days. The coating material is a commercially available enamel capable of withstanding temperatures up to 790°C (1450°F). It is readily applied to the surface of a variety of ferrous metals and alloys; however, best results are obtained with alloys low in chromium and nickel which also have a low thermal expansion coefficient.
- Chemical Engineering Division Summary Report October, November, and December, 1956
- A final series of runs was made in a four-inch continuous-flow mixing chamber to study the transfer of isobutanol into water and nitrobenzene into ethylene glycol. Satisfactory techniques were developed to provide for the rapid analysis of these systems. In addition, a light-scattering correlation was prepared to provide a measure of the interfacial area of the yellow-colored nitrobenzene-ethylene glycol mixtures.
- Metallurgy Division Quarterly Report [for] October, November, and December 1955
- A total of nine clad plates, containing uranium -5 w/o zirconium 1.5 w/o niobium alloy cores and clad with Zircaloy-II, were rolled in plain carbon steel jackets, heat treated, physically evaluated, and corrosion tested. All these plates were found to be within predetermined dimensional tolerance in width, thickness, length, cladding thickness, and core distribution. Improved control of wielding variables and of the length of the seal pin projecting above the end plugs resulted in the elimination of frequently observed segmented inclusions at the seal pin interfaces.
- Chemical Engineering Division Summary Report
- Measurement of radioactive carry-over was made on borax III operating at 300 psig and at power levels ranging from 4 to 14 mv. Decontamination factors of from 1.5 x 104 (at 14 mv) were obtained. These data are in essential agreement with those predicted by previous laboratory experimental work.
- Quarterly Report January, February and March, 1956
- The EBWR loading requires a total of 888 plates. It is anticipated that approximately 1000 plates will have to be produced to obtain the number of acceptable plates required for the loading. To the end of this quarter, 568 cladding billet cores acceptable with respect to chemical composition and physical soundness had been cast; this number represents 78% of the total number of cores cast. Approximately 75% of the Zircaloy-II stock required has been rolled, and about 55% of the cladding components required have been finished. The anticipated number of 495 cladding billets required for the thin (0.210") natural and enriched plates have been assembled, welded, sealed, and jacketed in steel. A total of 310 cladding billets have been rolled to fuel plates; of this number, 142 have been completely finished, and the remaining 168 are in the finish processing stages. The stability of the equipment for measuring the clad thickness of EBWR fuel plates has been improved by placing the phototube and the anthracene scintillator crystals in an insulated box with a temperature regulation of the order of 0.1°F.
- The Fabrication of Prototype Fuel Elements for the Experimental Boiling Water Reactor and the Experimental Breeder Reactor
- The purpose of this program was to develop techniques and methods for producing fuel elements for the Experimental Boiling Water and Experimental Breeder Reactors. Methods for fabricating large tubes, flat plates, and small pins were investigated. The tube and plates contained U-5 w/o Zr-1.5 w/o Nb alloy and were designed for the EBWR. The pins contained U-2 w/o Zr alloy and were designed for the EBR. Cladding and end seal material of Zircaloy-2 was required for the water-cooled EBWR elements. Unalloyed zirconium was specified for cladding on the sodium-cooled EBR elements.
- Quarterly Progress Report on Reactor Development 400 Program
- Physics calculations have been made for various combinations of the four types of fuel assemblies to be used in the EBWR core. Two thicknesses of plates, 0.205 in. and 0.274 in., including the two 0.020-in. cladding layers, are to be made of both natural U and U containing 1.44% U235. A total of 148 assemblies, 74 natural and 74 enriched, are to be fabricated with six identical plates each. Various configurations of these fuel assemblies will be used to (1) change the critical size of the core, (2) change the power distribution in the core, and (3) change the amount of reactivity corresponding to a given stream volume in the core. The physics calculations show that uncertainties in critical mass are adequately covered by the number and variety of fuel assemblies and that the possible changes in core characteristics with the different fuel assemblies should provide valuable information about the factors affecting maximum power density and stability in a boiling water reactor.
- Chemical Engineering Division Summary Report July, August, and September, 1956
- Additional runs have been made in the six-inch, continuous-flow mixing chamber to study the rate of mass transfer between isobutanol and water. These runs were inconclusive because the effluents were mutually saturated. A new four-inch cell has been designed and is being fabricated; this will permit a reduction in the time available for mass transfer. Consideration has been given to other liquid pairs which may transfer more slowly than isobutanol-water. The system nitrobenzene-ethylene glycol appears attractive.
- Reactor Engineering Division Quarterly Report Section I January, February, March. 1956
- Physical calculations have been performed for various combinations of the four types of fuel assemblies to be used in the EBWR core. Two thicknesses of plates (0.205 in. and 0.274 in., including two 0.020-in. cladding layers) are to be made of both natural uranium and uranium containing 1.44% U235. Any given fuel assembly contains six identical plates. A total of 148 assemblies, 74 natural and 74 enriched, are to be fabricated. Various configurations of these fuel assemblies can be used to (1) change the critical size of the core, (2) change the power distribution in the core or (3) change the amount of reactivity corresponding to a given steam volume in the core. Physics calculations show that any uncertainties in the required critical mass are adequately covered by the number and variety of fuel assemblies, and that the changes in core characteristics possible with the different fuel assemblies should provide valuable information about the factors affecting maximum power density and stability in a boiling reactor.
- The Manufacture of Enriched ZPR-III Fuel Plates
- This report is essentially a procedural account of the fabrication of certain enriched ZPR-III fuel plates for use in the ANL fast critical experiments at Arco, Idaho. A total of 208.92 kilograms of fully enrich, unalloyed uranium was processed. Of this amount 202.74 kilograms was received in the form of Oak Ridge type reduction buttons and 6.18 kilograms as pressed-powder plates. The completed fabrication consisted of 720 rectangular fuel plates having the nominal dimensions 3in. x 2in. x 1/8in. Their combined weight of 159.21 kilograms represents 76.22% of the weight of enriched material processed. The final distribution of the enriched material was as follows: [figure not transcribed].
- Table of Sin θ and Sin2 θ for Values of θ from 2° to 87°
- The table of sin θ and sin2 θ, to five decimal places for every hundreth of a degree from 2°-87°, has been prepared for the use of Professor W. H. Zachariasen in his X-ray diffraction studies. [Tables not transcribed]
- Summary Report of the Hazards of the Internal Exponential Experiment (ZPR-V)
- The Internal exponential Exponential Experiment (ZPR-V) will be constructed by loading up to 49 of the fuel cans, containing up to 155 kg of U235, of the present Fast Exponential Experiment in a 22-in. square iron tank, surrounded by an annular thermal region of fully enriched light water lattice 10 to 15 cm thick. This assembly will be placed in a 5-ft diameter tank which will, in turn, be located in the 10-ft diameter ZPR-II tank, the annular space between the outer tanks containing water for shielding. The new experiment will be a well-shielded, strongly coupled fast-thermal system. It will be possible to make measurements that cannot be made on the present Fast Exponential Experiment. One category of such determinations is the study of reactivity effects produced in the fast core, including control scheme studies and danger coefficient and oscillator measurements of such effects as Doppler coefficients and effect of lumping and streaming. The higher flux and excellent shielding will make beam studies of energy spectrum practical. Additional foil activations will be possible. Characteristics of mixed fast-thermal systems, which are of potential importance as power breeders, can be studied.
- Reactor Engineering Division Quarterly Report [for] October, November, December 1955. Section I
- The gastight steel building (400,000 cu ft) in which all radioactive components are to be housed has been completed by the Graver Tank Company. This structure was tested for strength at 18.75 psig (20% above design pressure) and then tested for leaks. No leaks were found in soap bubble testing of all welded seams. Continuous measurements of temperature and pressure over a ten-day period showed the leakage, if any, to be less than the 500 cu/ ft/day at 15 psig specified. The gastight cylinder was, therefore, accepted. General construction work by the Sumner Sollitt Company on the remainder of the plant has begun.
- ALPR Preliminary Design Study (Argonne Low Power Reactor) Phase 1
- A preliminary design study, Phase I of the ALPR project, has been made in accordance with the Army Reactors Branch specifications for a nuclear "package" power plant with a 200-260-kw electric and 400 kw heating capacity. The plant is to be installed at the Idaho Reactor Testing Station as a prototype for remote arctic installations. The "conventional" power plant as well as the exterior reactor components are described in the accompanying report and cost estimate by Pioneer Service and Engineering Company, Architect-Engineers for the project."Nuclear" components of the reactor are designed by Argonne National Laboratory as described in the present report.
- Report for July, August, and September 1950
- A quarterly report on various physics research projects conducted at the Argonne National Laboratory
- Quarterly Report: December 1, 1949 Through February 28, 1950
- Covers a quarterly period of reactor development performed by the Argonne National Laboratory operated by the University of Chicago
- Summary Report for October, November, and December, 1949
- A quarterly summary report of activities conducted at the Argonne National Laboratory.
- Mass Transfer from a Vertical Plate of Naphthalene to a Surrounding Air-Fluidized Bed
- A computer program was devised for the IBM 704, which can be used to correlate the mass transfer rate data from submerged surfaces to a surrounding fluidized bed.
- Comments on "Application of Fluidization to Nuclear Technology"
- The technique of solid fluidization has recently applied to the thermal conversion of uranyl nitrate into uranium trioxide. Two tests have been used to report the reactivity of uranium trioxide.
- Initial Superconductor Experiments
- The purpose of this note is to summarize superconducting experiments as an aid to future superconducting investigations
- Literature Survey of the Properties of Fission Product Oxides, Fluorides, and Oxyfluorides
- Interest in the reactions and properties of oxides and fluorides of the fission products prompted a literature survey of these substances
- The Problem of Oxidation of Uranium in ZPR-6
- A examination of having plates of unalloyed uranium exposed to the air in ZPR-6
- Metal Oxidation and Ignition : Parts I and II
- A study of the circumstances surrounding ignition did not definitely indicate a cause. This study has been undertaken as a basic laboratory investigation
- Dissolution of Thorium Oxide-Uranium Oxide Fuel Elements
- A promising new fuel element consists of urania in a thoria matrix is discussed.
- Separation of Molybdenum Hexafluoride from Uranium Hexafluoride--Simple Distillation of Uranium Hexafluoride-Molybdenum Hexafluoride Mixtures
- The separation of UR6 from MoF6 is important for ore processing and for the reprocessing of nuclear fuel by the fluoride volatility method
- Pyro-Met Plutonium Flowsheet
- To aid in outlining some of the problems this preliminary flowsheet for the core processing of a plutonium breeder was prepared
- Effect of Processing Losses on Breeding Gain and Doubling (Converting) Time for EBR-II and PBR Reactors
- Due to interest in the break-even point at which losses in the overall fuel cycle neutralize the breeding gain, the effect of losses on doubling (converting) time has also been calculated in this study.
- Literature Search on Chemical Reduction of Uranium Oxides
- Reduction of the oxide formed in the slagging operation has been suggested as a possible means of increasing the overall ingot recovery. The various reduction methods which have been investigated are reduction with calcium, magnesium, and carbon. Details of each method are discussed in this paper.
- Revised EBR-II Flowsheet for Fuel Cycle with Fuel Purification by Oxidative Slagging
- This flowchart supersedes the first one developed in ANL-LB-SL-693 for the purification of EBR-II core material by oxidative slagging
- Measurement of Neutron Flux in the EBR
- A study of the neutron flux distribution in the EBR was made utilizing the neutron activation of Au and P foils
- Nuclear Fuse Element Transient Heating
- A report on a test run of a nuclear fuse element in boiling reactor.
- Analysis for Pu(III) and Pu(IV) on a Tracer Scale
- A tracer scale analysis of the type of the zirconium phenylarsonate procedure is applicable to many types of solutions of plutonium which can not be analyzed spectrophotometrically and for that reason is very important.
- Drawing of Tubes filled with Particulate Solids
- The principal parameters involved in drawing of tubes filled with metallic, mineral and organic powders were investigated on small copper, mild steel and 304 stainless tubing.
- New Fabrication Procedures for High-Quality Tubing
- Small tubing used in nuclear reactors and other critical applications must represent the ultimate in structural integrity. Freedom from cracks and surface defects is particularly important
- Sintering and Properties of Uranium and Thorium Monosulfides
- Uranium and thorium monosulfides have been under investigation because of their potential as reactor fuel materials.
- Preliminary Irradiations of PuC and UC-PuC
- The irradiation of plutonium monocarbide is of interest because of the potential use of this fuel in small high temperature fast reactors.
- Plutonium Fuel Programs at Argonne National Laboratory
- Most of the plutonium fuel programs at ANL are aimed at the utilization of plutonium in fast reactor power systems. The principal incentive and instrument for the development of plutonium-bearing fuels is the second loading of KBR-II.
- Theory and Some Applications of Pulsed Current Fields to the Problems of Nondestructive Testing
- The velocity associated with a current field propagating in a good conductor that is of interest for the purposes of nondestructive testing is termed the "signal velocity."
- Neutron Radiography as an Inspection Technique
- The differences in absorption characteristics between neutrons and X-rays have been pointed out by a number of investigators. Such inspection situations would include the possibility of greater radiographic discrimination for certain materials, the inspection of heavy metals with reduced exposure times, and the relatively easy inspection of light materials by themselves or even when they are in an assembly with much heavier materials.
- The Corrosion of 1100 Aluminum in Oxygen Saturated Water at 70° C
- In oxygen-saturated distilled water at 70°, the rate and amount of corrosion during short exposure are influenced by experimental conditions. One noteworthy effect is that contamination of the water by the reaction increases the corrosion rate.
- Effects of Oxygen and Surface Treatment on the Corrosion of Stainless Steel in Superheated Steam
- Exploratory corrosion tests performed with stainless steels in steam at 650°C, 600 psi suggested a sharp dependence of corrosion behavior on surface treatment of the samples.
- Neutron Radiography : A 1962 Progress Report
- Since the relative absorption in materials for thermal neutrons and X-rays is very different the use of neutron radiography as a complimentary inspection method to X-radiography has many potential advantages.