Report issued by the Oak Ridge National Laboratory discussing the aging characteristics of the alloy Hastelloy B. Materials, equipment, experimental procedures, and results used to determine the characteristics of the alloy are presented. This report includes tables, illustrations, and photographs.

This quarterly progress report details ongoing work at the Oak Ridge National Laboratory as part of the Aircraft Nuclear Propulsion Project. Topics of discussion include reactor theory and design, shielding research, materials research, appendixes with information on ongoing analytical chemical studies.

The first step in a proposed processing method for recovery of uranium from graphite-uranium fuels consists of oxidation of the fuel by oxygen to volatilize the carbon. Residue ash from the combustion step can be treated in a variety of ways to recover and purify the uranium. The combustion step may be caried out by contacting the solid fuel in a fixed or moving bed with a stream of oxygen-bearing gas in a tubular or annular reactor. Oxidizing gas may be introduced to the reactor at several points up the reactor and there may be continuous or intermittent addition of fresh fuel and removal of residue ash.

Report issued by the Oak Ridge National Laboratory discussing work and progress made by the Health and Physics Division during 1963. Instrumentation, personnel monitoring, and laboratory monitoring is presented. This report includes maps, tables, illustrations, and photographs.

Many elements exhibit anisotropic scattering at energies of interest in reactor calculations. A method is presented for the calculation of transport cross sections including the observed anisotropy.

Nitrate contained in raffinates from Canadian Blind River mills interferes seriously with thorium extraction by secondary amines. With primary amines, however, the effect is negligible and these compounds are the preferred process extractants.

The thermophysical properties of Armco iron such as thermal conductivity, electrical resistivity, and Seebeck coefficient have been extensively investigated and reviewed up to 1000 degrees C. Few investigations of such properties have been made on high purity iron. If such a study is made using the same apparatus to determine the properties of two purity levels of iron, then several significant intercomparisons can be made which add meaning to data on a single material. The systemic errors for a single apparatus are the same, therefore comparison of a property of two similar materials is more significant. A comparison of the property changes with temperature and purity can show the effects of impurities on the mechanisms contributing to a property and allows prediction of the properties of iron as a function of purity. For these reasons a study was initiated on the high-purity iron for comparison to Armco iron.

Report presenting the calculation of the costs incurred in shipping irradiated uranium-containing fuel elements. A computer code that designs a cask and calculates the shipping costs is presented. By use of the code, shipping costs were calculated for typical reactor fuels.

This memorandum summarizes the results of an analysis to determine the extent of displacement of the EOCR core due to blowdown in case of several postulated hot main gas coolant pipe failures. Results show that the core will be damaged for any hot pipe double-ended failure. Excepting the improbable case of no coolant flow existing proper to the break, the core will be damaged for any hot pipe fracture exposing a total flow area to the atmosphere equal to that of one pipe. Smaller breaks will probably be safe in this respect.

The purpose of this study was to investigate mathematically the concentration decrease due to particle deposition phenomena in highly concentrated monodispersed aerosols (mean particle size less than 1.0 mu) flowing through ventilation ducts. It was found that, from the standpoint of removal, the decrease in concentration due to deposition on duct walls was insignificant; but, when considering contamination on duct walls, the amount deposited, even though small when compared with the amount in the bulk stream, should not be overlooked.

From summary: "The hazards due to possible chemical explosion and detonation of the dissociation gases in homogeneous-reactor systems have been considered."

Quarterly report discussing progress on the Fast Ceramic Reactor Development Program, "an integrated analytical and experimental program directed toward the development of fast reactors employing ceramic fuels, with particular attention to mixed plutonium-uranium oxide" (p. 1).

In certain phases uranium processing it is poss!ble for operating personnel to acquire internal deposits of uranium. This body burden can be acquired by injection, as in contamination of a wound, by ingestion, of by inhalation. In order to estimate internal exposures, some means of determining the location and extent of these internal deposits is essential.

A partial bibliography of reports on the Fluorox moving-bed for producing UO3, UF4, and UF6 is given.

The first man-made nuclear reactor -- or "pile" as it was then called -- was rather hurriedly improvised and operated in a crowded space under the athletic bleachers of Stagg Field at the University of Chicago on December 2, 1942. Just prior to this time, there began the assembly of a group of physicists with an unusual assignment. They were determined that radiation hazards of unprecedented proportions must be coped with successfully in the conduct of reactor programs as planned. Since these physicists were to be concerned with the health of radiation workers, they were called health physicists. There was no formal instruction available to this first group of health physicists and they perforce received training as they felt their way by firsthand experience and by trial and error. Health physics at Oak Ridge National Laboratory from the very beginning has been organized into three principal areas: applied activities, education and training and research.

From introduction: "The tube arrangement covered in this report was proposed in June, 1950, as a means of halving the weight and volume required for the heat exchanger. Ideally, the pressure drop across this arrangement should be little more than that required to overcome the skin friction on the tube surfaces. If the pressure drop across the tube spacers and in the cross-flow regions can be kept low, this ideal can be approached. much of the test work covered in this report was devoted to this phase of the problem."

Report containing a series of drawings detailing the Oak Ridge National Laboratory's high flux isotope reactor, its surroundings, and its systems.

Report documenting the ongoing research and developments of the Oak Ridge National Laboratory's Homogeneous Reactor Project.

Report issued by the Oak Ridge National Laboratory discussing quarterly progress made by the Homogeneous Reactor Program. It includes descriptions of progress and studies conducted during the report period, with tables, illustrations, and photographs.

A fifth in-pile loop experiment, L-4-8, was completed. The loop operated in-pile for a total of 1637 hr, during which time the LITR energy output was 4377 Mwhr. The average fission power in the loop based o cesium analyses was 622 w when the LITR was at full power (3 Mw). Based on oxygen data, the generalized corrosion rate for the first 300 hr was 4.0 mpy; the rate for the remaining 1357 hr was 0.7 mpy. The nickel data gave parallel results. The corrosion of the type 347 stainless steel, Zircaloy-2, and Ti-55AX [unintelligible] exposed in the core and in in-line holders was generally consistent with that observed in previous in-pile loop experiments. Some differences with steel were attributed to the fact that this was the first loop containing steel specimens operated with 0.04 m H2SO4 present in the uranyl sulfate charge solution (0.17 m UO2SO4, 0.03 m CuSO4). Stress specimens, made from the alloys Zircaloy-2, type 17-4 PH stainless steel, and Ti-C-130-AM, were exposed in care, in-line, and pressurizer locations. Microscopic examination and average weight loss gave no indication of effects attributable to the stressed condition of the specimens.

Several potential hazards that have been recognized and anticipated in the design and fabrication of the pressure vessel in the Homogeneous Reactor Test are discussed. These hazards results from the high operating pressure and temperature of the reactor, the exposure of the reactor vessel material to potential embrittlement and other affects of fast-neutron irradiation, and the need for containment of corrosive flowing liquids. The steps taken in recognition of these hazards are also discussed. The applicability of present codes to the reactor vessel fabrication is considered. Additional fields are suggested where recommended practices developed by code writing bodies could assist in development-type reactor design and fabrication.

Report presenting in-pile loop experiments DD, EE, FF, GG, L-412, L-413, and L-418, which were seven of a series designed to test the radiation corrosion of zirconium, titanium, and stainless steel alloys in solutions under various conditions of radiation intensity, temperature, solution composition, and velocity of flow. Steel specimens exposed in the loop cores showed increases in corrosive attack over that expected out-of-radiation.

The following is taken from ORNL-2257, Instantaneous Velocity Profile Measurement by Photography, by R. E. Lynch, L. D. Palmer, and G. M. Winn. This report is in in preparation status; and errors, inconsistances, and omissions in he language, as well as in the technical aspects, may exist. The technique of using phosphorescent particles to enable photographic determination of qualitative and quantitative instantaneous velocity profiles is covered by AEC Patent Application No. SN-710, 371 issued to F. E. Lynch, L. D. Palmer, H. F. Poppendiek, and G. M. Winn and entitled " A Method of and Means for Visualizing Fluid Flow Patterns.

An investigation of the materials for use in the water-moderated and cooled Aray Package Power Reactor (APPR) operating at about 500°F was made. The available literature was analyzed, and the results of the different investigators were compared and averaged. Twenty different materials, including stainless steels, nickel alloys, Stellites and others, were investigated from the point of view of physical properties, susceptibility to radiation damage, and corrosion resistance. Corrosion rates were established for all the materials under various conditions, such as irradiation, flow weld, stress, and various water conditions. Type-304 stainless steel was selected as the basic structural material. Operating conditions, to maintain minimum corrosion, were established also.