The work reported deals with the preparation and physical properties, especially thermal dissociation pressures, and densities of hydrides, hydrogen- metal systems, and mixtures of hydrides with other substances. Possible applicatlons as moderators, high-temperature neutron shields, and low-temperature shields are cited and design problems discussed. Most of the data on dissociation pressures cover ranges and compounds not hltherto explored because of experimental difficulties and the basic knowledge of the thermal behavior of hydrides was substantially increased. New hydrldes were prepared and several reported in the literature were shown not to exist. The following compounds, mixtures, and systems were studled: Tl-H, U-H, Ll-H, Na-H, Ca-H, Ba-H, Th-H, Sr- H; NaH-NaF, NaH-NaOH, NaH-CaH/, LlH-LiF, CaH/sub 2/-CaF/sub 2/, CaH/sub 2/-CaC/ sub 2/,CaH/sub 2/-Ca/sub 3/N/sub 2/; FeH/sub 3/ (alleged), NiH/sub 2/ (alleged), Ti(BH/sub 4/)/sub 3/, Th(BH/sub 4/)/sub 4/, WH/sub 4/ (attempted), W(BH/sub 4/)/ sub 4/ (attempted), /sub 4/NBH/sub 4/, (CH , and ydrides are ing an N/sub H/ comparable to water yet stable at red heat, compounds giving a neutron shield weight less than half that of water, and compounds suitable for use as hightemperature moderators containing large amounts of hydrogen. (auth)

This report is the final in a series of preliminary reports that follow the studies of scavenging systems related to radioactive fallout. The project consisted of two phases: preliminary experiments to relate the sizes of particles in air to specific radioisotopes, and preliminary laboratory studies of scavenging of particles by liquid drops, including studies of sticking probability and effects of Brownian motion and water vapor diffusion.

From abstract: "Data obtained in laboratory studies of the 'fumelses dissolving' of irradiated uranium are presented. Procedures for handling dissolver off-gases are proposed."

From abstract: This bibliography consists of references pertaining to "Effects of Nuclear Radiation Upon Electronic and Electrical Components", and to "Electronic and Electrical Components for Critical Environments". References were compiled from the Classified and Unclassified report literature and the published literature. The following indexes and abstract journals were used: Weapons Data Index, AEC Card Catalog, Engineering Index (1950-55), and Office of Technical Services, Research Reports (1956-March '57).

Report discussing the use of a small-source approximation for engineering calculation of thermal piles. Explanation for how this type of approximation is more appropriate in certain problematic situations where a homogenization approximation would be doubtful is given.

This test will determine approximately the relationship between the power rate of rise indicated by the power rate meter and the actual pile power, rising period, and power rate. The actual pile power and power rate cannot be measured accurately during rapidly changing conditions; it is the intent of this test primarily to demonstrate that rate of rise protection offered by the power rate meter an be calculated to the same order of accuracy as the measurements of actual conditions.

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This report describes the second part of an IBM 704 Computer Code to determine transient thermal stress in pressure vessels. This part computes stresses from temperature distribution determined by the first part. A brief account is given of the solution of axially symmetric thermal stresses, using harmonic stress functions. This is followed by a description of the Code and complete input and operating instructions for a number of problems: a. Transient temperature distribution due to sudden change of temperature. b. Transient temperature distribution due to surface temperature changing as a function of time. c. Thermal stresses for temperature problems ( a ) or (b). d. Thermal stresses due to previously known temperature distribution. The code is known as TURF6. It was found that the computer results for thermal stresses in a solid sphere subjected to a sudden change of temperature at the surface corresponded within 5% of the theoretical values. The IBM 704 Computer required for this code has a 16 K memory, three drums and two tape units.

The schedule for the installation of the PAR slurry loop experiment in the South Facility of the ORR has been reviewed and revised. The design, fabrications and Installation is approximately two weeks behind schedule at this time due to many factors; however, indications are that this time can be made up. Design is estimated to be 75% complete, fabrication 32% complete and installation 12% complete.

The purpose of this report is to present the results of laboratory experiments in which coolant water flowed backwards through a ``C`` reactor process tube with operational charge-discharge front fittings under conditions of typical reactor rear header pressures and post-scram rear header water temperature and tube powers. This information is of value in planning and interpreting transient experiments investigating the consequences of the rupture of reactor front face piping to a single tube.

Report documenting the suitability of Alabama for plant locations to produce synthetic liquid fuels, based on raw materials, coal, natural gas, and oil-impregnated strippable deposits.

Herman Lurie's weekly report for the sugar market. Details include prices, distribution, sugar tons, and refiners.

Introduction: The purpose of this report is to describe in further detail the control systems that interconnect the SRE with the steam plant.

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A study of the radiation effects of fast neutrons (O.5 ev to about 15 Mev) and gamma rays (fission spectrum) on various pieces of aircraft instrumentation is re ported. Components studied were: fuel gage vacuumtube amplifier and indicator, transistorized (germanium and silicon) servo amplifier system, and 4-in. diam. gyro. The fuel gage was subjected to a total neutron flux of about 6 x 1O/sup 5/nvt and about 5 x 1Osup 8/ r gamma with no adverse electrical effect. Physical damage was evident. Four transistorized amplifiers were irradiated in a fast flux of about 1.0 x 10/sup 10/cm/sup 2/ sec. Rapid loss in gain, rapid increase in d-c current drain, and complete failure occurred when the integrated flux reached about same flux. Studies on the gyro were inconclusive due to equipment failure. (auth)

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Abbreviated press feed preparations of UO/sub 2/ powder appear suitable for the automatic pressing of pellets on a production scale. Progress was made in achieving compartment uniformity in disc-compartmented stretch-formed fuel tubes and controlling the free space for pellet expansion within the compartments. The designs for additional in-pile and out-of-pile test specimens were completed. A reference control rod program was developed on the basis of outside-in'' rod withdrawal in order to approach uniform fuel burnup and fission product buildup in the core. Stainless steel specimens suffered no detrimental attack as a result of their exposure to borated water under simulated reactor conditions in a dynamic test loop. The successful fabrication of a full-size strap design fuel assembly model demonstrated the feasibllity of welding tubes lnto bundles. Pressure drop versus flow curves were experimentally determined for the reactor vessel model Various fuel loadings in the CRX core were used in materials buckling measurements, control rod worth determinations, and instrumentation evaluations. The latest revisions of mechanical, thermal, and nuclear design data are appended. (For preceding period see YAEC-52.) (D.E.B.)

A brief summary, with reference literature for details, of pertinent and important process flowsheet conditions which are in use in the Purex Process is presented. (auth)

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Studies concerned with the deposition of salts from simulated fuel solution under boiling conditions were continued. Additional deposition tests were carried out in loop L-2-23 in which the core section was heated. Several reagents were tested to determine their ability to dissolve stainless steel corrosion products. A simulated HRT fuel solution was stable with regard to U concentration when diluted to 23 ppm at 250 C; however, some Cu was lost from solution. Preliminary tests in a uranyl sulfate solution indicate that an 18-8 stainless steel alloyed with either small amounts of Pt or Cu is more corrosion resistant at low flow rates than a conventional 18-8 stainless steel; however, the addition of either Cu or Pt did not reduce the corrosion rate of 18stainless steel at high flow rates. A study was made to determine the susceptibility of off-specification type 347 stainless steel to intergranular attack by uranyl sulfate solutions. Continued testing with a single heat of cast type 347 stainless steel has confirmed previously reported data to the effect that the cast alloy is more resistant a stress-corrosion cracking than is the wrought alloy. Six commercial grades of Al/sub 2/O/sub 3/ suitable for use as check balls in …

The CUREBO code for the IBM 704 is described. The code is divided into three parts including the calculation of nuclear cross section of the various physical components of a reactor (WOX7), the solution of the multigroup diffusion equations in two-space dimensions in order to find neutron fluxes and sources for an operating reactor containing these components ( CURE), and the calculation of fuel and poison depletion as a result of operating this reactor under steady- state conditions (BO2). (auth)

6 9 2 4 8 < E 5 0 2 5 H 5 of a solution containing 0.04 M UO/sub 2/SO/sub 4/, 0.025 M D/sub 2/SO/sub 4/, 0.03 M CuSO/sub 4/ and 750 ppm of nickel as nickel sulfate at 300 deg C in D/sub 2/0 but both nickel and copper were lost from solution when the nickel concentration was increased to 1700 ppm. The chemical stability of a solution containing 0.036 M UO/sub 2/SO/sub 4/, 0.022 M D/sub 2/SO/sub 4/, 0.009 M CuSO/sub 0.013 M NiSO/sub 4/ and 0.0016 M MnSO/sub 4/ in D/sub 2/0 was also demonstrated at 300 deg C. The corrosion of stainless steel in 0.04 M UO/sub 2/S/sub 4/O/sub 4/ containing 0.025 M H/sub 2/SO/sub 4/ and 0.0 1 M CuSO/sub 4/ was found to be relatively severe above the critical velocity at 250 and 300 deg C. The critical velocities observed were 10 to 20 and 30 to 40 fps, respectively at 250 and 300 deg C. Loop runs in which oxygen was removed from solution have shown that as uranium precipitated excess acid, ferrous and nickelous ions formed. On addition of oxygen to such a solution, uranium dissolved and ferrous ions were oxidized …

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