The three isomeric (5-phenyl-2-oxazolyl)pyridines have been shown to be sensitive fluorescent pH indicators which show a pronounced change to increased visible fluorescence as the pH is lowered. Absorption and fluorescence spectral data and pK values are given. Selective excitation of fluorescence from the conjugate acid in the presence of the free base was found possible. The sensitivity of the 4-isomer was demonstrated to be adequate for determining the small amounts of acid produced in certain chemical dosimeter systems. (auth)

In the production of plutonium in a reactor, plutonium-238, 240, and 241 are formed as well as Pu{sup 239}. It is well known that the specific alpha activity of the plutonium varies as the percentages of these isotopes are changed. Kinderman, et al have worked out the relationship between isotopic content and MWD/ton exposure. Their findings are reported in this document.

The conceptual design of a 100 Mw(e) nuclear power plant is described. The plant utilized a sodium-cooled graphite-moderated reactor with stainless- steel clad. slightiy enriched UO/sub 2/ fuel. The reactor is provided with three main coolant circuits, and the steam cycle has three stages of regenerative heating. The plant control system allows automatic operation over the range of 20 to 100% load, or manual operation at all loads. The site, reactor, sodium systems, reactor auxiliaries, fuel handling, instrumentation, turbine-generator, buildings. and safety measures are described. Engineering drawings are included. (W.D.M.)

Present thinking and planning appears to be developing from the following factors as concern the 13.8 KV temporary construction power limit. 1. It is understood that the present intent is to supply 100-N operating requirements from a single stub source in the 230 KV loop. 2. The original thoughts were to obtain construction power over a 13.8 KV line from 151-D substation. 3. Construction load requirements are now less than originally planned since steam has been substituted for electrical drive of primary loop pumps and 5500 hp motor tests are no longer necessary. 4. An extreme emergency backup source for the K plants has always been of concern, although minimized in recent planning. It is desirable to review the temporary construction line requirements from a future operating viewpoint to determine if the line could be useful to the operating plants after completion of construction. It is highly desirable to provide T.C. power source from K plants rather than 151-D and then leave the line and breakers in place for future maintenance assistance and as extreme emergency backup to K plants.

The continuous calciners at UO{sub 3} Plant are of a new design which was developed at HAPO and placed in service late in 1956. The heat transfer troughs are considered to be the most vulnerable parts of the calciners because of their high operating temperatures. Thermal stresses are calculated to be quite high, and when added to the direct mechanical stresses from powder load etc., come close to the limiting safe stress for stainless steel at the operating temperature. It is felt that trough failure will be progressive as a result of creep type distortion and cracks at stress concentration points. The higher the stress (and temperature), the more rapid the rate of failure will be. All possible steps have been taken to limit maximum trough temperature, rate of change, and variations from one part of the trough to another. These steps were not always successful and various troughs have been subjected to rather severe temperature shocks as well as high mechanical stresses due to agitator failures. Despite these difficulties, no signs of failure could be detected by visual inspection. It was decided, therefore, that a more complete examination should be made. This examination was made to determine the present …

Enriched beams of 450 Mev/c K/sup -/ mesons and antiprotons have been produced by separation with the coaxial static electromagnetic velocity spectrometer. Characteristics of the final separated beams as observed in the 15- inch hydrogen bubble chamber are given together with a detailed description of the beam optics and apparatas. (auth)

Two types of hypothetical reactor catastrophe are considered. In the first of these, the Boiling Accident,'' it is assumed that a fraction of the radioactive material in a reactor is released to the atmosphere at a steady rate over a period of hours. In the second, the Puff Accident,'' it is assumed that the release of the radioactive material takes place instantaneously.'' The following concepts are used as measures of the hazard existing outside the controlled plant area. Danger Distance,'' defined as that distance beyond which the fission product cloud becomes so dilute that it cannot cause death; Probabiiity of Death per Capita per Accident,'' which is a measure of the hazard to any individual; and Expectation Number of Deaths per Accident.'' which is a statistical measure of the hazard to the entire off-site populace. Three mechanisms for each type of catastrophe were considered: direct irradiation from the fission product cloud, inhalation of the air in the cloud, and rainout from the cloud followed by irradiation from the ground. Failout is not considered. for it requires that a very energetic explosion be assumed. It is concluded that the size of the plant should be set by the hazard of irradiation …

Energy distributions and energy release rates of fission product gammas and energy release rates of the fission product betas are presented in tabular and graphical form, and the computation methods are outlined in detail. The data given for beta and gamma decay rates pertain directly to the thermal fission of U/ sup 235/. The possible effects of neutron energy spectra, other than thermal, and neutron flux level upon the fission product decay rates were examined, and the results are reported. The importance of bremsstrahlung caused by beta particles from fission products and the effect of photoneutron production due to the photonuclear reactions of the fission product gammas with D or Re were studied. Methods of evaluating these effects in the reactor are briefly described. (J.E.D.)

>The analytical procedures for wet adsorption methods for surface area measurements are described. (J.S.R.)

This report addresses the advanced energy sources and conversion techniques.

A concpept for an advanced enineering test reactor is described. Thpe results of various investigations with regard to alternative design possibilitipes are presented. Retailed information is given rearding the nuclear and enineering calculations. (W.D.M.)

A general survey is presented of information in the uranium alloy field. Emphasis is centered on alpha uranium-rich alloys of special interest as uranium- base fuel element materials. The systems treated include uranium-molybdenum, uranium-silicon, uraniumzirconium, uranium-niobium, and uranium-zirconiumniobium (high uranium compositions). The uraniumfissium alloys are discussed in relation to their projected applications as fast reactor fuels. Equilibrium diagrams, transformation kinetics, and other pertinent data are presented for the uranium plus fissium'' element systems, e.g., uranium-ruthenium, uranium-rhodium, uraninum-palladium, and uraniummolybdenum-ruthenium. The properties covered include constitution phase diagrams, metallographic structure, physical and mechanicaly properties, transformation kinetics, crystallographic structure, thermal cycling, ad irradiation stability (where pertinent). Correlations between microstructure, heat treatment, and dimensional stability are emphasized, with particular reference to the U-2 wt.% Zr, U-5 wt.% Zr, and 1 1/2 wt.% Nb alloys. A discussion of the role of alloying and heattreatment in improving the dimensionaly stability and corrosion resistance of uraaium is presented, and an evaluation is made of the present status in attaining these objectives. (auth)

Shielded enclosures equipped with viewing devices and remote-hauling equipment for use in experiments and processes involving radioactivity are referred to as hot cells. The hot cell includes the biological shield enclosing the working space, viewing devices, special ventilating equipment, and special equipment for use in the hot cells, such as manipulators, cranes, machine tools, and measuriag devices. A hot cave is the same as a hot cell. A junior hot cave is a small-sized hot cave. A summary is presented of pertinent data on hot cells in use at various AEC installations. (C.H.)

The creep properties of Ag-base alloys and specifically of the Ag-15% In- 5% Cd wrought alloy are presented. (W.L.H.)

On June 6, 1958, KW Operations had HCR Channel No. 16 borescoped to determine why this rod could not be inserted on May 3. Observations revealed 3X balls in the channel and horizontal separation between graphite blocks ranging from 1/2 inch to 2 1/4 inch. The separations were noted only in the first fifteen feet in from the outer skin and in the last five feet of the channel. As a result of these findings and past operational difficulties with certain HCR`s at both KE and KW Reactors, a program of measurements is in progress to determine the extent and causes of the stack displacements. From measurements and observations so far obtained, the following conclusions may be drawn about separations at locations of consequence to the loss of 3X balls from channels: Side to side horizontal separations totaling two to three inches have been observed at both reactors. The significant separations lie outside of the VSR pattern and in the lower half of the stack. The VSR pattern does not include the six outer ball 3X channels. There are probably a few small separations (< 1/2 inch) at the outer VSR`s. There are probably no separations large enough to admit …

ABS>A method for determination of submicrogram quantities of Al in reactor cooling water by neutron activation analysis is described. Data obtained in analyses of samples from the OKR and the Bulk Shielding Reactor are included. (J.R.0.)

Data are presented on the results of acid and carbonate leaching studies on samples of ore from the LaBajada Mine of the Lone Star Mining Company, Santa Fe County, New Mexico. (auth)

An annlog computer study of the ORR gas-cooled loop was made. The effects on fuel and gas temperatures of changing gas (nitrogen) flow rates, turning the heater on and off, and inserting or removing the fuel were determined. It was found that the temperature of the sample, if fuel, could be raised more rapidly than it could be reduced, and that the temperature could be reduced faster with high rates of flow. All of the equations used in building up the computer are given, and some of them are derived. (auth)

One of the major factors restricting reactor power levels is the incidence of ruptured slugs. The primary purpose in studying ruptures is to determine how reactor operating variables affect rupture rates. With this knowledge reactor operating conditions may be adjusted or controlled in the manner that will optimize reactor production. In addition, knowledge of rupture rate relationships are useful in fuel element development and in overall economic studies of existing and proposed reactors and reactor processes. This report is a compendium of various types of rupture information largely developed during the past eighteen months. Plant rupture experience for CY-1957 is reviewed; rupture rate correlations with reactor variables for solid natural uranium material are presented; a comparison between solid and cored natural uranium material rupture rates is made; the basis for current discharging practice of rupture-prone metal lots is discussed. 11 figs., 2 tabs.

ABS>Experimental power-reactivity feedback transfer functions were calculated from the EBWR power transfer function measurements. A simplified model of the EBWR kinetics was developed, using an analog computer, and an analytic expression was obtained for the feedback function. The analytic solution was fitted to the experimental functions to obtain power coefficients and time constants for various modes of operation. These data were extrapolated, and a power transfer function was predicted for 40 Mw. The reactor fumction was measured and compared with the prediction. A stability study was carried out, using open loop transfer functions containing the experimental feedback functions. Extrapolation of the gain and phase margins indicated stability to at least 66 Mw. The reactor was successfully operated at 61.7 Mw following this, with power limited by the capacity of the feedwater pumps. The use of the simplified model for parameter studies is demonstrared by a series of calculations to evaluate the effect of heat transfer time constant on stability. (auth)

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The development of a matrix method of solving multigroup diffusion equations in nonmultiplying regions is described. The method is applied to a three-region shielding problem, and comparison is made with experimental results. Equations obtained by this technique can be solved with a desk calculator. (auth)

Twenty charges each of low hydrogen dingot fuel and standard ingot fuel elements were irradiated at DR-reactor. Distortion data from 22 tubes are analyzed in this document. The dingot fuel elements are prone to greater tube filling capacity (TFC) and greater daimeter growths at both center and ends of the fuel elements (the growths at center are larger than at ends). The difference between average warp values for the two types of fuel elements is not statistically significant, indicating that the greater TFC values shown by th dingot fuel elements are due to their larger diameter growths at the slug center rather than to larger warp values.

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