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.

Standard water treatment practices have proven adequate for scale and corrosion control on the secondary side of the ORR pool-cooling heat exchanger. Corrosion rates measured on specimens exposed to secondary water showed maximum corrosion rates of approximately 1 mpy during the first exposure period and these rates decreased during subsequent exposure.

A preliminary study of dispersing agents for thorium oxide has been completed and several of the dispersants have possible uses. Also many of the industrial dispersing agents tested are not usable with thorium oxide due to induced behavior causing balling and caking. The effects of nitric acid concentration have been observed to also effect each dispersing agent.

Temperatures are given at which two liquid phases form in a synthetic homogeneous reactor fuel solution and its concentrates. The data show a two-liquid-phase boundary temperature of 332°C for the Particular HRT Fuel composition and a flat minimum temperature of 305°C for the initial solution concentrated between 300 and 329°C are presented to indicate solution stability in this temperature region. Some related comments on current HRT operations are given.

The method of calculation of various quantities relating the temperature and the rate of heat removal of UO2 test elements is presented. Graphs of the appropriate quantities as a function of the rate of heat removal are included.

An improved heat sink for space vehicles was proposed in ORNL-CF-59-1-21. Subsequent work on the problem has indicated that there is a substantial probability of a puncture of such a radiator by meteors. To meet this problem a modified geometry has been evolved for which the probability of a meteor puncture should be reduced by a factor of at least 100 relative to the radiator of the original proposal at the expense of an increase in radiator weight of possibly 30%. This memorandum outlines the approach to the problem and a series of similar promising solutions.

Preliminary measurements of the corrosion of titanium were made in 13 M HNO3-0.05 M fluoride using 0.1 M H3BO3 as a liquid and vapor phase complexing agent. Titanium Ax-55 was attacked at average rates of 0.58 and 0.33 mils/month in the liquid and vapor. In dissolver solutions containing 0.5 and 1.0 M titanium, all rates were less than 0.1 mils/month.

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.

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.

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.

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.

The measured diffusivity of uranyl nitrate in water at 25ºC was 0.7*10^e cm^2/sec with about 40% average deviation. A program was started to develop nonnuclear uses for depleted uranium. Two continuous DRUHM reaction tuns were terminate due to erratic operation of the sodium metering system. In the second Fluorox run with crude UF, which lasted for 29 hr, a total material balance of 94.8% was obtained and 17.9& of the theoretical amount of UF was collected in cold traps and chemical traps.

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

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.

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.

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.

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.

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.

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.

Grade CGB graphite is a nuclear graphite which is basically an extruded petroleum coke bonded with coal tar pitch. No carbon blacks are used and the low-permeation graphite is finished through a series of impregnations and heat treatments with a final heat treatment of all components to 2800 degrees C. A listing of the results obtained is given in Table 1. The results at 51 degrees C are considered questionable. There was a slight contamination of the 90% Pt 10% Rh-Pt thermocouples at 910 degrees C but it was not sufficient to doubt the validity of the 910 degrees C results. However, the results obtained at 1015 degrees C should be disregarded because of severe thermocouple instabilities. In addition, the electrical resistance of the core heater at 603 degrees C indicated the thermocouples had a -10 to -15 degree error which is sufficient justification to disregard the 605 degrees C data.

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.

The experimental details, mathematical models, and typical data for a rapid comparative method for thermal conductivity measurements are presented. The method consists of measuring the temperature change of a small silver sphere after it is brought in contact with a small disk-shaped specimen which was initially at ta higher temperature. This temperature change was calibrated in the range of 50 to 400 degrees C by making measurements on samples of know thermal conductivity. The accuracy of this technique was shown to be between than +-10% with a reproducibility of at least +-2.5%. Using known transport mechanisms for heat conduction in solids and the temperature dependency of the electrical conductivity, a means to judiciously extrapolate thermal conductivity data obtained between 50 and 400 degree C to high temperature is presented.

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.

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.