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- 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.
- 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.
- Aqueous Corrosion of Uranium and Alloys: Survey of Project Literature
- This report describes corrosion rate of uranium in hydrogen-saturated water appears to be constant with respect to time after a brief induction period and to involve only one type of over-all reaction, in which pitting effects are slight or nonexistent.
- Argonne High-Flux Research Reactor: AHFR Conceptual Design Study
- Report issued by the Argonne National Laboratory discussing high-flux research reactor design studies. As stated in the summary, "this report presents a reactor design to meet the needs of the basic research program at Argonne National Laboratory. The program requires some irradiation in thermal flux...and some beam experiments..." (p. 5). This report includes tables, and illustrations.
- Argonne National Laboratory Annual Report: 1956
- Report issued by the Argonne National Laboratory discussing the variety of work done at the laboratory during the year of 1956. This report includes tables, illustrations, and photographs.
- Argonne National Laboratory Annual Report: 1957
- Report issued by the Argonne National Laboratory discussing the variety of work done at the laboratory during the year of 1957. This report includes tables, illustrations, and photographs.
- Argonne National Laboratory Annual Report: 1958
- Report issued by the Argonne National Laboratory discussing the variety of work done at the laboratory during the year of 1958. This report includes tables, illustrations, and photographs.
- Argonne National Laboratory Idaho Division Summary Report: July-September, 1960
- Report issued by the Argonne National Laboratory covering a summary report of the work conducted by the Idaho Division. Experimental work and progress made on reactors are presented. This report includes tables, illustrations, and photographs.
- Argonne National Laboratory Idaho Division Summary Report: October 1960-March 1961
- Report issued by the Argonne National Laboratory discussing a summary of projects conducted with the Idaho Division. Descriptions of each project are presented. This report includes tables, and illustrations.
- Atomic Energy Commission Division of Reactor Development Reactor Information Meeting. Part VI, Processing; October 7, 8, 9, 1953
- On October 7, 8, and 9, 1953, the Atomic Energy Commission Division of Reactor Development held a reactor information meeting at the Argonne National Laboratory. The objective of the meeting was exchange of information among people actively concerned with the design of reactors for power to the end that the power reactor program would move more speedily and more economically to another milestone of success. In this volume all the papers presented at the meeting are listed. Copies are given of those papers which are available, and references to published reports are indicated where know for those papers not included in this collections. In order to facilitate handling, this volume is being issued in six parts: Part I Power Reactors; Part II Reactor Physics. Critical and Exponential Experiments Measurements; Part III Reactor Components. Reactor Economics Considerations. Reactor Safeguard and Control; Part IV Fuel Element Design and Problems. Corrosion and Chemistry; Part V Heat Transfer; Part VI Processing. The Author Index is being bound and distributed with Part I of this volume.
- Attack on Uranium by Lithium at 600 C
- The tests described in this report were static tests devised to afford a basis for a quick evaluation of the resistance of uranium to attack by lithium. The work was done at the same time as the tests of beryllium, thorium, and various engineering metals in lithium (described in ANL-4990); but the results with uranium are given in the present classified report so that the results of the other tests can be published as an unclassified document. The procedure for carrying out the tests is described in ANL-4990.
- Autoclave Testing of Mechanically Jacketed Thorium Slugs In Water
- Water corrosion tests on mechanically jacketed and pinholed thorium slugs show that these slugs fail in a manner similar to that observed for mechanically jacketed and tested uranium slugs. The proposed mechanism for the water corrosion of these jacketed slugs is analogous to the water corrosion mechanism of jacketed uranium slugs presented in the project lecture. A bare thorium slug appeared to be more resistant to corrosion by water than a mechanically jacketed slug during the first half of the autoclave test. After approximately 90 hours of testing both the bare and the mechanically jacketed thorium slugs were severely corroded by water.
- Calculations on U235 Fission Product Decay Chains
- Report of equations for calculating decay of U235. The introduction states" Calculations have been made on the U235 fission product decay schemes. The results for a typical example, that of a reactor operating at 1000 kilowatts for 180 days, have been tabulated and graphed. General formulae have been used so that the results can be applied for any power level and any time of irradiation" (p. 2).
- A Capsule Design for Experimental High-Flux Irradiations Of Fuel Materials
- New reactors presently in design or construction stages, as well as revised operating procedures for existing reactors, have shown an increasing emphasis on extending the exposure time of the reactor fuel elements. However, operating experience at Hanford, as at other installations, has demonstrated that as the amount of burn-up in uranium metal is increased an increase is also noted in operational difficulties resulting from the dimensional behavior of the fuel. During reactor irradiation uranium slugs or rods have been observed to change in length and diameter, to warp, and to develop surface roughening.
- Casting of Blanket Bricks, Ring, Plug, and Control Rods for the Experimental Breeder Reactor (CP-4)
- The cup assembly of the Experimental Breeder Reactor (EBR) consists of blanket bricks, inner rings, and safety plug, all of natural uranium. The design of the finish machined brick is shown in Figure 1. These pie shaped bricks when stacked together, 12 bricks arranged in circular form and stacked seven rings high, comprise the cylindrical portion of the outer blanket, 17.875" I.D. x 30.875" O.D. after machining and canning. The inner ring, which is shown in Figure 2, fits inside the bottom layer of bricks. The circular opening in the center of the inner ring is closed by the safety plug shown in Figure 3.
- 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.
- 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.
- 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.
- Chemical Engineering Division Summary Report : January, February, And March 1953
- Dissolution of the large number of samples obtained from the natural uranium blanket of the Experimental Breeder Reactor after approximately 485,000 kw.-hr. of operation has been completed, and analysis of these samples for uranium consumed and plutonium formed is well along. An attempt is being made to distinguish quantitatively between uranium-238 and uranium-235 fission in the blanket area by determining the ratio of ruthenium-106 to cesium-137 in the fission products.
- Chemical Engineering Division Summary Report January, February, and March, 1954
- Progress is reported on (1) direct cycle boiling reactor studies, (2) solvent extraction, (3) fluoride volatilization separation process, (4) elevated temperature separations, (5) fluidization studies, (6) development of analytical techniques, (7) processing and utilization of radioactive wastes.
- 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.
- 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 October, November, and December, 1953
- Progress is reported on (1) experimental breeder reactor program, (2) solvent extraction, (3) fluoride volatilization separation process, (4) elevated temperature separations, (5) denitration of uranyl nitrate in a fluidized bed, (6) development of analytical techniques, (7) processing and utilization of radioactive wastes.
- 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.
- Chemistry Division, Section C-1, Quarterly Report July, August, and September 1952
- Report describing the research and development activities related to nuclear chemistry and radiochemistry and basic chemistry conducted by the Argonne National Laboratory Chemistry Division, Section C-1.
- Chemistry Division, Section C-1, Summary Report for April, May, And June 1952
- Report describing the research and development activities related to nuclear chemistry and radiochemistry and basic chemistry conducted by the Argonne National Laboratory Chemistry Division, Section C-1.
- Chemistry Division, Section C-1, Summary Report For October, November and December 1951
- Report describing the research and development activities related to nuclear chemistry and radiochemistry and basic chemistry conducted by the Argonne National Laboratory Chemistry Division, Section C-1.
- Chemistry Division, Section C-II, Summary Report For April, May, And June 1952
- This report was written by different scientist on various experiments of solid state, physical chemistry, radiochemistry and analytical, and special problems.
- Chemistry Division, Section C-II, Summary Report for July, August, and September 1952
- This report deals with the (1.1) physical properties of graphite, (1.2) effects of pile irradiation on the properties of graphite, (1.3) effect of irradiation on "ceramic" materials, (1.4) effects of radiation on ice -- the x-ray induced luminescence of ice, (1.5) investigation of color centers and other optical properties of single crystals. (2.1) radiation chemistry of liquids, (2.2) application of mass spectrometry to chemical problems, (2.3) vapor pressure and heat of vaporization of uranium, (3.1) nuclear properties of Zr93 and Nb93m from fission, (3.2) mass distribution in the spontaneous fission of Cm242, (3.3) Upper limit to the lifetimes of the first excited states of Th236, U234, and Pu236, (3.4) on the one-body model of alpha radioactivity, (4.1) spectrographic analysis, (4.2) chemical analysis, (5.1) paramagnetic resonance measurements, and (5.2) the 60-inch cyclotron.
- Chemistry Division, Section C-II, Summary Report for July, August, and September 1952
- This report deals with the (1.1) physical properties of graphite, (1.2) effects of pile irradiation on the properties of graphite, (1.3) effect of irradiation on "ceramic" materials, (1.4) effects of radiation on ice -- the x-ray induced luminescence of ice, (1.5) investigation of color centers and other optical properties of single crystals. (2.1) radiation chemistry of liquids, (2.2) application of mass spectrometry to chemical problems, (2.3) vapor pressure and heat of vaporization of uranium, (3.1) nuclear properties of Zr93 and Nb93m from fission, (3.2) mass distribution in the spontaneous fission of Cm242, (3.3) Upper limit to the lifetimes of the first excited states of Th236, U234, and Pu236, (3.4) on the one-body model of alpha radioactivity, (4.1) spectrographic analysis, (4.2) chemical analysis, (5.1) paramagnetic resonance measurements, and (5.2) the 60-inch cyclotron.
- 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.
- Comparative Analysis of ANL High Purity Uranium
- In the course of the development at Argonne of high purity uranium metal in ingot form, some questions arose as to the validity of the chemical analyses of some of the impurities (particularly those for carbon, boron, and silicon), with one analytical laboratory reporting concentrations in some instances of an order of magnitude greater than another laboratory. Since the low concentrations of impurities in this material involved, in some cases, the development of modified analytical procedures and standards, it was decided to check these discrepancies by having identically prepared samples analyzed by several AEC and associated laboratories. This report is a compilation of the results obtained.
- Condensation of Metal Vapors: Mercury and the Kinetic Theory of Condensation
- Report issued by the Argonne National Laboratory discussing condensation theories of metal vapors. As stated in the introduction, "the objectives of this research then are critical analysis of condensation theories and data for metal vapors and experimental evaluation of the resistance to condensation for a representative metal such as mercury" (p. 18). This report includes tables, illustrations, and photographs.
- Corrosion and Stability Tests on Chemical Poisons in Higher-Temperature Water
- Corrosion-stability tests have been made in static autoclaves at 500 and 600F on solutions of compounds having high neutron cross sections to evaluate their usefulness for shutdown purposes. The only compound tested which appeared to be completely stable in 600F water was boric acid. Limited corrosion data did not show it to cause excessive corrosion of zirconium or stainless steel.
- Corrosion of Plutonium Alloys in NaK
- A plutonium-aluminum alloy containing 4 atom per cent aluminum showed no attack after exposure to purified NaK for one month at 400 C in the absence of any oxide. The same specimen and other plutonium alloys, including pure plutonium, showed marked deterioration in shorter exposure in the presence of oxide films from a welded stainless steel container. Pure uranium was resistant even in the presence of such oxides.
- Cost Study of a 100-Mw(e) Direct-Cycle Boiling Water Reactor Plant
- Report issued by the Argonne national Laboratory discussing a technical and economic evaluation of a direct-cycle light-water boiling reactor designed for natural circulation and internal steam-water separation. The reference 100-Mw(e) reactor power plant design evolved from the study should have the best chance (compared to similar plants) of approaching the 8 to 9 mill/kwh total power-cost level. This report includes tables, and illustrations.
- Critical Studies with G. E. Type Fuel Elements
- The ZPR-I is a facility to study low power critical assemblies using enriched uranium as fuel, having a light water moderator and an essentially infinite water reflector on all sides. The fuel is held in elements 43" long with a 1" square cross section. Any of these elements may be placed in or removed from any position in the reactor tank. Thus, any desired core configuration may be easily obtained.
- The Decomposition of Light and Heavy Water Boric Acid Solutions by CP-3' File Radiations
- The behavior of light and heavy water solutions of boric acid toward pile radiations has been investigated as a function of boric acid concertation. A study has also been made of the effect of hydrogen, hydrogen peroxide, and potassium iodide on the radiation stability of boric acid solutions.
- Deposition of Corrosion Products by Cataphoresis
- This report is a record of experimentation conducted intermittently over a period of two years and directed toward preventing deposition of transport corrosion products on fuel elements and other critical components in high-temperature, circulating water nuclear reactor. It includes the postulated mechanism for deposition, a description of experimental equipment, experimental data, results obtained from the experiments, and recommendations for future study.
- Design and Testing of a High-Heat Flux Electron-Bombardment Heater
- Report issued by the Argonne National Laboratory discussing the testing of an electron-bombardment heater. As stated in the abstract, "the applications of electron-bombardment heating to liquid-metal heat transfer and reactor safety experiments are discussed. The design of a high-heat-flux, electron-bombardment heater (EBH) is presented" (p. 7). This report includes tables, illustrations, and photographs.
- Determination of Zirconium and Total Fluoride Ion in Zirconium - Hydrofluoric Acid Solutions
- An analytical method has been developed for zirconium and fluoride ions in the system resulting from the dissolution of fuel elements in hydrofluoric acid. The method is based on determination of the density and electrical conductivity of the dissolved metal solution.
- Development of a Process to Produce Zirconium Hanford Type Process Tubing by Roll Forming And Inert Arc Welding
- The development of methods which were successful in producing zirconium Hanford type process tubing by roll forming and inert are welding (He) flat strip to which appropriate rails had ben previously attached by resistance welding is described in this report. Grade 2 drip arc melted crystal bar material was used.
- The Development of Equipment and Methods for Centrifugally Casting Reactor Fuel Slugs
- This technical report describes the design and construction of equipment and the development of methods for multiple mold, centrifugal casting of reactor fuel slugs. Advantages of the centrifugal casting method over the conventional fabrication methods were found to be (1) fewer operations, (2) fewer and more easily recovered residues, (3) less expensive equipment, and (4) the production of fuel slugs in shapes and in alloys not well adapted to other methods of manufacture. The method consisted of vacuum melting the alloy in stoppered crucibles and bottom pouring into a spinning rotor carrying 16 radially arranged copper molds. The castings so produced were used without further processing, except for cropping the sprue end to obtain the specified length.
- Development of Zirconium Clad Uranium Plates for Reactor Fuel
- In this investigation a method has been developed for production of Zrclad uranium plate based upon previous investigations at Argonne and Oak Ridge National Laboratories, and preliminary in investigations in their present work. The present preliminary investigations included experimental studies of the effect of interface atmosphere upon roll bonding and led to the conclusion that thin layers of air between bonding surfaces could be tolerated.
- Differential Thermal Analysis of Irradiated Diamond and Silicon Carbide
- It was demonstrated by differential thermal analysis (DTA) that: 1. Catastrophic amounts of energy can be stored in diamond. 2. Even at low irradiations, the release takes place over serval hundred degrees, indicating a spectrum of activation energies. 3. At higher irradiations, the stored energy release is considerably less than the increased energy contents and seems not to have been completely released even at the highest temperatures reached. 4. There is some indication of an increased heat capacity below the temperature of stored energy release. It was shown by DTA that large amounts of energy can be stored in silicon carbide on irradiation. The release was found to be spread out over a greater range of temperatures than in diamond and indicated a larger and higher group of activation energies. Catastrophic release was not achieved. The amount of stored energy released over the range of temperatures used was 140 cal/g in a sample irradiated in a water-cooled test hole at HEW for an exposure of 265 Mwd/aT.
- Dimensional Stability of Uranium Powder Compacts Upon Thermal Cycling
- Thermal cycling tests on uranium have shown that the dimensional changes that occur on cycling in the alpha range are directly related to both the texture of the material and its grain size: cold rolled rods generally elongate in the direction of rolling, while the same rods, after a beta-treatment, grow at rates several orders of magnitude lower. This considerable improvement by beta-treatment has been attributed to the texture randomization accompanying the heat-treatment. In the course of this heat-treatment, however, considerable grain growth occurs, which ahs the effect of causing surface roughening on cycling (also referred to as "bumping"); fine grained material generally retains a smooth surface. These observations led to the speculation that the most desirable structure in uranium, from standpoint of dimensional stability, is one that combines both a random texture and a fine grain size. Heat treatment of rolled rod offered no easy method to obtain such a product; powder metallurgical techniques, however, appeared ideally suited for the purpose. To this end, early in 1949, the Sylvania Electric Products Company initiated a program to develop suitable techniques for producing uranium powder compacts having the above-mentioned desired characteristics. Because of the availability of thermal cycling equipment at Argonne, the Metallurgy Division of the Laboratory has undertaken to evaluate the stability of the various experimental compacts produced in the developmental phases of the program. This report contains the results of these evaluations. The data in this report indicate that compacts of nearly theoretical density and fine grain size can be obtained by hot pressing uranium or uranium hydride powders in the high alpha temperature range.
- Dissolution of Uranium Oxide Arising From Slug Failure
- The purpose of this work was to study reagents which might be effective in dissolving uranium oxide produced during slug failures in water-cooled reactor systems. An aspect of this problem which has subsequently become of primary importance is the solubility or transportability of the oxide in pure water.
- The EBWR: Experimental Boiling Water Reactor
- Report issued by the Argonne National Laboratory discussing the Experimental Boiling Water Reactor (EBWR) power plant. Designs of the final EBWR power plant are presented. This report includes tables, illustrations, and photographs.
- Eddy Current and Ultrasonic Testing of CP-6 Fuel Elements
- The fuel element to be used in the Savannah River reactors is a natural uranium slug 1.00 in. in diameter and 8 in. long, encased in a 2S aluminum can 1.080 in. O.D. having a wall thickness of 0.035 in. The slug is bonded to the can with an aluminum silicon alloy, using the Hanford Al-Si process.
- Effects of Metal Purity and Heat Treatment on the Corrosion of Uranium in Boiling Water
- Corrosion rates of present reactor grade uranium were measured in boiling distilled water and were found to have higher values almost by a factor of two then previously reported corrosion rates of uranium. Mallinckrodt biscuit metal showed corrosion rates in the same medium somewhat lower than reactor grade uranium, and high purity metal prepared at Argonne National Laboratory corroded considerably less rapidly than the biscuit metal.