The thermodynamic properties of the alkali halides are reviewed, presenting data supplementary to that of Brewer and Brackett, Chem. Rev. 61: 425-432(1961). (T.R.H.)

Forty-nine hydrocarbon products were evaluated in the laboratory in a search for a highly stable diluent for the organic extractant used in a radiochemical separations plant. The formation of zirconium ligands during chemical and radiolytic degradation increased with boiling point for isoparaffinic producte but was approximately constant for n-parafiins. Of those materials that met the local flash point specification, the n-parafiins were superior. (auth)

The percentage of helium which will bypass the core if the reflector system shown is used is predicted. It is estimated that nominally about 0.1 to 0.2% of the total flow will bypass the core, which is not considered excessive. The most difficult parameter to determine was Z, the gap between the sealing surfaces. The method used to predict Z is presented. The effect of bowing due to a temperature gradient across the seals is discussed. (auth)

None

The present knowledge of the shielding constants of concrete is summarized. The densities, elemental compositions, and mixes, where available, are given for a wide range of concretes. From these data, various constants useful for shielding calculations were computed. These constants include the removal cross sections, total cross sections, average thermal neutron absorption cross sections, thermal neutron diffusion coefficients, reciprocal thermal neutron diffusion lengths, total gamma-ray linear attenuation coefficients, gamma-ray energy absorption linear attenuation coefficients, the effective atomic numbers for the determination of buildup factors, and the capture gamma-ray spectra. Experimental results are included where available. (auth)

A process scheme for the recovery of uranium from fuel elements has been developed. The scheme combines continuous fluid-bed drying and fluoride volatility techniques after initial dissolution of the fuel element in the appropriate aqueous system, hence the designation ADF, Aqueous Dry Fluorination Process. The application of this process to the recovery of uranium from highly enriched, low uranium-zirconium alloy plate-type fuels is described. ln the process, the feed solution is sprayed horizontally through a two-fluid nozzle and is atomized directly in the heated fluidized bed. The spray droplets are dried on the fluidized particles and form a dense coating. Excessive particle growth was limited by the use of air attrition-jets inserted directly in the bed. Aqueous hydrofluoric acid solutions containing l.2 to 3.6 M zirconiuni, 0.007 to 0.03 M uranium, and free acid concentrations from 1 to about l0 M were successfully processed in a 6-in.-diameter Inconel fluid-bed spray dryer. Rates equivalent to about 3.l kg/hr of zirconium were achieved, 160 ml/min with the most concentrated feed solution. Experiments were successfully carried out from 240 to 450 deg C. A new design for a two-fluid nozzle was developed. Extensive work was done to identify the various zirconium fluoride compounds …

An arc furnace is described which employs a nonconsumable tungsten electrode and a water-cooled copper hearth. It is used successfully for melting refractory metals and alloys. The furnace is equipped with a vacuum system, an inert gas supply, and an 800-ampere directcurrent power supply. (auth)

Changes in neutron multiplication caused by voids in the island of the HFlR were calculated and measured experimentally. The results indicated that with only water initially in the island the maximum change in neutron multiplication ( DELTA k/sub max) associated with island voids is 0.032 with a corresponding void fraction of 70%. With a simulated 300 g Pu target in the island DELTA k/sub max/ was 0.0l6, and the corresponding void fraction was 42%. In view of these large changes in neutron multiplication, calculations were made to determine what additional materials could be used in the island to reduce DELTA k/sub max/ and what the associated decrease in peak thermal flux wouId be. The results indicated that of the materials considered the use of beryllium in the water island resulted in the smallest decrease in flux for a specified DELTA k/sub max/. To reduce DELTA k/sub max/ to 0.01 required 26% by volume of beryllium in the island; the corresponding reduction in thermal flux, as compared to an all-water island, was about 10%. In order to reduce DELTA k/sub max/ to 0.0l with a 300 g Pu target in the island, the aIuminum-to-water ratio of the target had to be …

The 15,000 MW nuclear economy assumed for the long range study of pot calcination costs reported earlier was used as a basis for calculating optimum fill volumes. An algebraic expression was developed for cost as a functmon of the normalized radius of the central void space in a partially filled vessel. Minima of this expression were found for acmdmc and neutralized wastes in 6, 12, and 24in.-diameter vessels. Optimum fill volumes decreased as vessel diameter increased, varying for acidic wastes from 99.8% for 6-in.-diameter vessels to 92.5% for 24-in.diameter vessels. Decreases in costs by using optimum fill volumes instead of the 90% fill volume assumed for all cases in the long range study were small, the largest being an 8% decrease for neutralized wastes in 6- in.-diameter vessels. (auth)

Channel-to-channel flow distribution within Type 3 (SM-2, stationary and control rod fuel elements modified for use in the SM-1, SM1-1A, and PM-2A core support structures and control rod tubes was measured in single element flow testing. Plots of channel-to-channel flow distribution and element pressure drop at various element flow rates are given. Flow distribution for the top-orificed SM-1A and PM-2A stationary elements was within the plus or minus 12% deviation from element average utilized in previous thermal analyses of these cores. Testing of the bottom-orificed SM-1 stationary element and the SM-1, SM-1A, and PM-2A control rod assemblies showed flow distribution exceeded plus or minus 12% devation from average. Simple modifications to the SM-1 stationary element indicated the possibility of improvng fiow distribution in that element. (auth)

A large ultrahigh vacuum chamber was built for environmental testing of components for the SNAP program at temperatures as high as 1000 deg F. The chamber employs diffusion, electronic, and cryogenic pumping to handle high gas loads at high temperature and ultrahigh vacuum. A unique internal heating system, connections, assemblies, flanges, and test set-up jigs are described in detail. (auth)

A method is presented for translating inspection measurements of fuel plate spacing to obtain minimum coolant channel clearances under reactor operating conditions. Considerations of fuel plate ripple growth and the inspection procedure used are included. The method is applied to establish dimensional tolerance specifications used for the procurement of SM-1A and PM-2A Core II. (auth)

GAMMA I is a FORTRAN-compiled program for the IBM 704 Electronic Data- Processing Machine. It embodies a certain general, uniform procedure H of mathematical logic for seeking out a proof of any theorem within any mathematical theory which is given in formal axiomatic form. An extended discussion is provided of the underlying method and of the necessary background of mathematical logic. The program is described in detail. (M.C.G.)

The growth of nonprotective, crust-like oxide films was encountered in high-temperature alloy systems that contain molybdenum, vanadium, or tungsten as strengthening additions. The cause of accelerated oxidation in such alloys appears to be associated with the characteristically low melting temperatures of oxides of these refractory elements. The factors that contribute to a breakdown of oxidation protection in these systems are outlined and remedial methods which may be used to avoid catastrophic oxidation are discussed. Commonly encountered service failures that have resulted from catastrophic oxidation are also described. (auth)

An investigation was made to determine the effect of rolling temperature, roll pass design, amount of reduction, and heat treatment before and after rolling on the anisotropic growth rate of uranium under irradiation. The growth rate was found to increase with decreasing rolling temperature and with increasing reduction of area at 300 deg C. The rate of elongation was proportional to the amount of (0l0) component present or, where shortening occurred, to the amount of (l00) component. Oval-edgeoval roll passes resulted in somewhat higher irradiation growth rates than did round roll passes. Recrystallization after rolling effectively reduced the irradiation growth rate of uranium rolled at temperatures of 500 deg C and lower. Irradiation caused length shortening in uranium which was beta quenched after being round-rolled at temperatures of 400 deg C and above, and which was beta quenched after being oval- rolled at temperatures of 300 deg C and above. (auth)

Work is being carried out to develop and demonstrate on pilot plant scale integrated processes for treatment and disposal of radmoactive wastes. High-level waste calcination, low-level waste treatment, economic and hazards evaluation, engineering evaluation, disposal in deep wells, disposal in natural salt formations, Clinch River studies, fundamental studies of minerals, and White Oak Creek basin study are discussed. (M.C.G.)

Studies carried out on corrosion associated with the hydrofluorination- dissolution phase in the fused-salt Fluoride Volatility Process are summarized. Corrosion for hydrofluorination-dissolver vessels used in bench-scale and semiworks-scale process development at ORNL is discussed. The results of a study on construction materials for the dissolution phase are presented. Corrosion studies at ANL are described for comparison purposes. A full-size hydrofluorinator dissolver is described. (M.C.G.)

Fifty-seven papers and l7 abstracts of papers are presented in the report. Fifty two of the papers are abstracted separately; in addition, a single abstract is written to cover the section on Plasma Physics Theory, which contains 3 papers and 8 abstracts of papers. The two brief papers not abstracted separately are concerned with fast neutron detection, and homogeneous critical assemblies of 3%enriched UF/sub 4/-paraffin systems. (T.F.H.)

A bibliography on pion-pion interactions is presented. The 241 references are those available through November 7, 1961. A chronological listing is given together with excerpts or comments. In another part the references are grouped according to subject. An author index is included. (M.C.G.)

>An assembly was constructed which simulated the moderator coolant annulus in the Experimental Gas-Cooled Reactor. This assembly was preheated to various temperatures and air was passed through the coolant annulus. Under certain conditions it was demonstrated that self-sustained combustion of the graphite could occur. Rates of temperature rise were generally less than 1 deg C/ min until the graphite temperature exceeded 700 deg C and then rise rates became very high. In these cxperimcnts, the assembly was not operated in such a manner as to give ignition temperatures eharaeteristic of the EGCR. These tests were designed only to investigate the effects of changing such parameters as the rate of coolant flow, air humidity, prior oxidation on the graphite, and air inlet temperature. A later series of experiments to more closely duplicate the EGCR operating conditions was completed and will be reported in a second report, HW- 71182. (auth)

Existing experimental data on the variation of reactivity with core geometry are reviewed. Four typical fast neutron systems are analyzed to predict: (l) the variation of critical mass with cylindrical core geometry (core and reflector composition are held fixed); (2) the reactivity worth of fuel at the radial core boundary as a function of cylindrical core geometry; and (3) the geometric variation of heat removal parameters; these include the ratio of: (a) Maximum power density to average power density in the core. (b) Maximum power density to average radial power density in the core. (c) Total reflector power to total core power. The absolute values of all of these parameters are determined by the core and reflector compositions of the four systems. These were chosen to simulate typical constituents of interest to reactor analysis. Two systems represent a typical fast reactor and a typical fast critical experiment. The other two systems represent compositional combinations of the two basic systems. The results of the analyses show that the significant geometric varia tion is in items (2) and (3b). Itenm (1) is almost constant for small variations near the optimum geometric configuration. Outside of this range, the variation of critical mass with …

A method is presented for the determination of uranium (as U/sup 238/) in uraniuni-loaded graphite fuel elenients by a non-destructive, direct radioactivity analysis technique. A 200-cbannel pulse-height analyzer, equipped with a 3 in. x 3 in. NaI(Tl) crystal, is used to measure the Np/sup 239/ radioactivity of the neutron-irradiated samples. The amount of U/sup 238/ in the test samples is deterimined quantitatively by comparing the Np/sup 239/ radioactivity induced in each sample with the Np/sup 239/ radioactivity induced into known standards of U/sup 238/ processed under the same conditions as the test samples. The limit of detection for U/sup 238/in samples of normal uranium composition for this method is about 1.0 x l0-4 ug. (auth)

A pump was developed for pumping carbon dioxide in a closed loop without introducing impurities. This pump will give flow rates of up to 3 liters/min and will develop a working pressure of over 70 mm Hg. No wear was observed after 2000 hr of testing. It is felt that this pump is more desirable for this application than those developed by other experimenters for two reasons: relatively inexpensive construction of the pump and the associated electronic circuit and the low coefficient of friction between the piston and the cylinder wall. Various modifications are suggested which will make this pump satisfactory for other applications. (auth)

In laboratory studies the dissolution of iron in dilute nitric - hydrochloric acid mixtures shows an apparent reaction order of -0.62 with respect to HCl. No apparent order value for HNO/sub 3/ can be determined over the concentration ranges studied. Nickel dissolutions show apparent orders of 1.4 with respect to the HCl and 4.2 for HNO/sub 3/. Activation energies determined from 50 to 80 deg C are not constant, ranging from l0 to 20 kcal per mole for both metals. (arth)