The Bevatron operated an average of 80% of the scheduled operating time with the exception of one 5-day shutdown period for emergency maintenance of the motor generator sets. The machine was shut off l9% of the scheduled operating time because of component failure and 1% at the request of the user. During and following the Christmas shutdown period an ultrasonic inspection was made of the flywheel keyway on each of the motor generator sets. No faults were apparent. Charpy samples were taken from the inspection bore in the flywheels. Subsequent"vee notch" Charpy impact tests indicated that the flywheels are operated in a brittle region rather than in the ductile range expected. The research program again was devoted mainly to study of pi - and K-meson interactions. (auth)

The distortion behavior of graphite as a function of irradiation temperature is reviewed. The behavior of needlecoke and CSF graphite was examined over moderate exposures in the GETR. Results showed needle-coke to be less contracting than CSF. Details of contraction show a minimum contraction rate per 10/sup 21/ nvt at 600 to 800 deg C for both types. Limitations to be placed on the data presented are listed. (P. C.H.)

The condenser-discharge method of conducting molten metal- water reactions at high temperatures was refined. Two methods to measure energy input to specimen wires and, therefore, to compute initial metal temperatures were developed. Calculated metal temperatures were estimated to be accurate to within 100 deg C. Two reaction cells were designed, one for operation at atmospheric pressure with water at room temperature, and the other for operation at high pressure and with water at elevated temperature. Means were developed to determine the surface area of metal exposed to reaction and to determine the total extent of reaction. Pressure transducers were used to record the rate of reactions. The zirconium- water reaction was studied with initial metal temperatures from 1100 to 4000 deg C with 30 and 60-mil wires in room-temperature water. Initial pressures in these runs were the vapor pressures of water at room temperature (20-30 mm). Runs were made with 60-mil wires in water heated to 200 deg C (225 psi). Results in room-temperature water indicated that the reaction became explosive at an initial metal temperature of 2600 deg C. Below this temperature, 20% or less reaction occurred. At higher water temperatures, reaction ranged from 40 to 70%. Runs in …

Recently a T = 1 resonance in the {Lambda}-{pi} system called Y{sub 1} has been observed with a mass of 1385 MeV. Two types of resonances have been predicted that might relate this observation to other elementary-particle interactions: (1) P 3/2 resonances in the {Lambda}-{pi} and {Sigma}-{pi} systems predicted by global symmetry, corresponding to the (3,2/ 3/2) resonance of the {pi}-N system, (2) a spin-1/2 Y-{pi} resonance resulting from a bound state in the {bar K}-N system. The position and the width of the observed Y{sub 1} resonance agree with both theories, but since the spin and parity have not yet been determined, they cannot distinguish between the two theoretical interpretations.

A loop was operated in the Brookhaven Graphite Research Reactor to determine the effect of in-pile irradiation on the corrosion of various materials by a U-- Bi solution. The loop wws fabricated of 21/4% chrome-1% Mo steel and contained, in the in-pile section, specimens of low-chrome steels, C steel, Mo, Be, Ta, and graphite. The U--Bi solution containing 869 ppm U/sup 235/ 98 ppm U/ sup 238/, 236 ppm Zr, and 346 ppm Mg was circulated at 51/4 gpm. A temperature difference of 75 deg C was maintained on the loop. The in-pile test section ran at 500 deg C and the finned cooler section at 425 deg C. The in-pile test section was exposed to a neutron flux of 4.4 x 10/sup 12/ neutrons/cm/sup 2/-sec which provided a fission density of 5.5 x 10/sup 10/ fissions/cm/sup 3/-sec. Metallographic examination indicated that the corrosion and/or erosion of the steel and graphite specimens was nil. Wetting of the specimens by the U-Bi solution was limited. Results indicate that in-pile and out-of-pile experimental results are similar and that fission fragment recoils did not contribute materially to either wetting or corrosion under the conditions imposed in this test. (auth)

ABS>A statistical evaluation is presented of field data including measurements of alpha activity on work surfaces, in the environmental air; and the results of urinalyses obtained from employees assigned to a UF/sub 6/ manufacturing facility. Two years are studied wherein personnel were subject to relatively low concentrations of both soluble and insoluble normal uranium compounds. Approximately 700 urine specimens were collected and 500 spot-air samples, 4000 shift-length air samples, and 24 surface surveys were made. Statistical correlations of these data are presented, and a method is proposed for estimating the exposure potential of employee groups working in environments having varying uranium concentrations. (auth)

Results of studies of the known parameters for direct calcination of plutonium nitrate in an agitated bed and conversion of PuO/sub 2/ to PuCl/sub 3/ by fluid bed processes are presented. A description of the processes and the demonstration equipment is included. The study of variables is discussed along with the parameters and plans for demonstration runs. (J.R.D.)

None

Organic reactor technology is critically evaluated and areas of research and development work now lacking or inadequate for the successful development of this reactor concept are indicated. The development programs for present organic and heavy water moderated concepts appear generally adequate to reach specific goals. However, the narrow scope of the organic reactor program should be broadened to assure coverage of areas where the application of novel principles might result in marked economic benefits. Further work, principally of a basic nature, is recommended in the fields of chemistry, processing, management, and thermodynamic properties of coolants, in fuel development, and in concept evaluation. (N.W.R.)

None

Measurements of material bucklings for graphite uranium systems are summarized. A comprehensive listing and guide to the original data sources is provided. Complete information on physical and nuclear properties of the lattice and the geometry of the exponential assembly is included, along with some of the auxiliary data taken. The fuel sizes vary from 0.925 to 2.5 in. in diameter for five different fuel geometries. The lattice spacings vary from 4 3/16 to 15 in. Over 300 measurements of material buckling are included. (auth)

Separate abstracts were prepared for 43 sections of this report. (C.H.)

Strontium recovery from Purex 1WW was investigated with simulated feeds and tracer activities. Initial experiments demonstrated recovery of over 70% of the strontium by sulfate precipitation from partially neutralized 1WW by either increasing the sulfate concentration to about 3 M or by adding carriers such as lead. Precipitation of iron was avoided by addition of one or more moles of tartrate per two moles of iron. Precipitation at elevated temperatures and addition of lead after pH adjustment were shown to be beneficial. Strontium recoveries of over 95% were achieved by precipitation at about 80 deg C at pH values of 0.4 to 4.0 with sulfate concentrations of 0.67 to 3 M and over 0.02 M lead carrier added. High sulfate concentrations were required at low pH, but the sulfate concentration is not critical above pH 1. Some separation of strontium from cerium was observed at pH 2 to 4, with the degree of separation being dependent on both tartrate concentration and pH. Recovery of strontium from the sulfate precipitate by leaching with nitric acid, sodium acetate, or oxalic acid was unsatisfactory because of the large volumes required. Metathesis with sodium hydroxide removed most of the lead as plumbite and converted …

Critical studies were made with a simulated, large, dilute power reactor having uranium carbide as fuel. The uranium in the core was 30.7% enriched, and the atomic ratio of uranium to carbon was 0.946. The critical mass was 503.01 kg U/sup 235/ and the critical volume 574.47 liters. Central reactivity coefficients, effective fission crosssection ratios, heterogeneity effects, reactivity worth of distributed materials, foil irradiations, and the average prompt neutron lifetime were measured. Multigroup calculations using the Yiftah, Okrent, and Moldauer crosssection set overestimated k for the critical configuration by 4.7%. (auth)

Liquid radioactive wastes are converted into solids, with volume reduction factors of 3 to 8, by flash drying on finely screened, porous, inert solid particles (e.g. alumina) in a fluidized bed at 320 to 550 deg C. The wastes may be either aluminum nitrate-containing wastes from the processing of MTR-type fuel elements, or Purex Process wastes. Ruthenium is found to be the only volatile fission product in this temperature range. Methods are described for its removal from the fluidizing gas. (T.F.H.)

The development of apparatus and laboratory techniques for the study of materials compatibility with 1500 to 2200 deg F potassium was attempted. Techniques for corrosion tab preparation, dry box capsule filling and sampling, and vacuum filling and sampling are described. Apparatus for rotating capsule testing to 2000 F, rotating capsule testing to 2400 deg F, and anisothermal see-saw'' capsule testing is also described. (auth)

Tests were performed to obtain station performance data at various steady-state generator loads. The station 0 was operated for four-hour periods at steady state conditions and levels of 5, 21, 42, and 61 Mw gross generator output. The various readings are presented in tabular form. A list is given of equipment in service during the test. All plant components operated satisfactorily during the test. (M.C.G.)

None

The High Flux Beam Reactor, HFBR, is cooled, moderated, and reflected by heavy water and designed to produce 40 Mw with a total epithermal flux of ~1.6 X 10/sup 15/cm/sup -2/ sec/sup -1/ and a flector thermal maximum flux of 7 X 10/sup 14/ cm/sup -2/ sec/sup -1/, using a core formed by ETR plate-type fuel elements in a close-packed array. The hazards summary is given in terms of site description, reactor design, building design, plant operation, disposal of radioactive wastes and effluents, and safety analysis. (B.O.G.)

A test was performed on June 19, 1959 to determine the capacity of the Core Removal Cooling System for removing reactor decay heat under split-flow'' conditions. The system operated satisfactorily during this test; the pumps developed a flow of approximates 73 gpm at a total head of 254 ft water, as compared with their rated capacity of 75 gpm at a total head of 250 ft water. (D.L.C.)

A program is described which is devoted to heat transfer and pressure drop measurements on single tube sections with coolant flow only on the inside. The tests were conducted on simulated PM-1 fuel elements. Data are included and data reduction methods are discussed. (J.R.D.)

BS>Critical studies of a fast reactor core containing a simulated oxide fuel having an oxygen-uranium atomic ratio of 1: 1 are described. Calculated and experimental critical masses are compared. Experimental results are given for fission ratio, central reactivity coefficient, fuel bunching, and distributed worth measurements. (auth)

tems were cast and cured. Results of chemical tests on aa epoxy curlang exudate are included. Comparison of solvent effects on retention of radioelements by stainless steel was started and data are tabulated for Ac/sup 227/, Th/sup 227/, a nd Ra/sup 22//sub 3/. Work on protactinium was resumed after suspension of this project in 1960. Methods for preparation of small quantities of highly enriched U isotopes are being examined. Included in the survey are chemical exchange, electromagnetic separation, gaseous and liquid thermal diffusion, gas centrifugation, and photochemical techniques. Continued investigation of viscosities of La and Pr for use in Pu alcontinued along with studies of Pu bearing glass fibers. (J.R.D.)

The basic objectives of this program were the determination of the thermal stability of various fission product oxides and nitrates, and an investigation of the gas phase decomposition of ruthenium tetroxide. To accomplish these objectives, a literature survey was flrst made of available physical and chemical data for the oxides and nitrates of Cs, Sr, Ru, Zr, Nb, and Ce. The data were supplemented by a calculation of thermodynamic functions for RuO/sub 4/ vapor from the experimentally determined infrared spectrum and the theoretically calculated raman-active fundamentals. Data are presented graphically. (C.H.)