From preface: This report was prepared by a task force that met continuously in January and February of 1959 for the purposes of making a relative comparison of the fluid fuel reactor concepts for technical feasibility.

From foreword: This report describes the goal of the overall SDR program to demonstrate the commercial generation of electrical power.

Abstract: In September-October 1958 a six-day reconnaissance in the Tocopilla area, Antofagasta Province, Chile, was made by members of the U.S. Atomic Energy Commission and the Instituto de Investigaciones Geológicas de Chile for the purpose of evaluating reported uranium occurrences.

Based on the results of detailed fabrication studies, an evaluation of the effects of varying the type and size of UO/sub 2/ particles, the type and size of stainless steel matrix powders, blending procedures, compacting pressures, sintering times, temperatures, and atmospheres, roll-clading temperatures and reduction rates, total cold reduction, and heat-treating times and temperatures was made for UO/sub 2/stainless steel dispersion fuel elements. Transverse tensile tests, creep-rupture tests, metallographic examination, radiography, density measurements, and x-raydiffraction studies were used to evaluate the structure and properties of the fuel elements. From these studies a reference fabrication procedure for GCRE fuel elements was established. The fuel element core contains minus 100 plus 200-mesh hydrothermal UO/sub 2/ dispersed in an 18-14-2.5 alloy matrix prepared from minus 325-mesh elemental iron, chromium, nickel, and molybdenum powders. Commercial Type 318 stainless steel is used for cladding. Core compacts are sintered in steps to 2300 deg F after cold compacting at 15 tsi. Evacuated picture-frame packs are hot rolled from a hydrogen muffle at 2200 deg F with a 40% reduction in thickness on the first pass and a 20% reduction in thickness on remaining passes. After annealing at 2300 deg F, the fuel elements are given a …

ABS>Published information is not available for accurate prediction of the electrical insulating characteristics of helium at high pressures and temperatures. In general the breakdown voltage increases as the gas pressure is increased and decreases as the gas temperature is increased. The relatively low breackdown voltage of helium accents the importance of additional investigation in this field. (auth)

To determine the safety of the array in the storage vault for the SM-2 experimental fuel plates, two criticality criteria were applied. A maximum of 18 fuel plates was stored in sthainless steel tubes and the tubes belted to a frame on the wall to prevent movement. No tube could go critical by itseIf. The vauit was then assumed completely flooded by water. In the first calculation, the fuel array was assumed to be distributed uniformly over the wall forming a large slab. This method indicated the array might be critical if the steel tube and cadmium lining were neglected. In the second method, a conservative calculation, wnich included the steel tube and cadmium lining was made. This method indicataed the array was subcritical. Calculations were then made of the criticalty of the SM-2 vault without the steel--cadmium tubes and wcoden blocks. The multiplication factor of the vault was also calculated. In order to determine the accuracy of these calculations, an ORNL critical experimental array was calculated applying the same analytical techniques. (M.C.G.)

A fuel-cycle economic study was made for a 315-Mw(e) graphite-moderated U/sup 235/-Th-fueled fused-salt reactor. Fuel cycle costs of approximately 1.3 mills/kwh may be possible for such reactors when reprocessed for U/sup 233/ and U/ sup 235/ recover y at the end of a 9-year cycle. Continuous removal of fission products during the reactor cycle does not appear to offer any great economic advantage for the converter reactor considered. (auth)

The rate of oxidation of chromium(III) to chromium(VI), catalyzed by ruthenium, was determined at various temperatures and oxygen concentrations. The rate at 300 deg C was too rapid for measurement by aliquot sampling. In the temperature range of 225 to 275 deg C, oxidation was rapid and the rate increased with oxygen concentration. A linear dependence of initial oxidation rate on the reciprocal of chromium(VI) concentration suggested that a rate-controlling step in the reaction mechanism may be desorption of chromium(VI) from the ruthenium catalyst. The activation energy calculated for the reaction is 19 kcal/mole. (auth)

A conceptual study on design of sodium-cooled reactor steam generators was conducted. Included is a detailed description of the preliminary design and analysis, based on the use of known materials and existing methods of fabrication. (See also APAE-41 Vols. I and III.) (J.R.D.)

It was demonstrated that neptunium(IV) can be readily absorbed onto anion exchange resins from 6 M HNOsub 3/ containing ferrous sulfamate and hydrazine or semicarbazide, separated from plutonium, uranium, and common metallic impurities by washing the resin at 25 deg C with 6 M HNO/sub 3/ containing ferrous sulfamate and hydrazine or semicarbazide, separated from fission products and thorium by washing the resin at 60 deg C with S M HNO/sub 3/- 0.01 M HF containing hydrazine or semicarbazide, and eluted at concentrations greater than 40 g Np/l with 0.35 M HNO/sub 3/ at 25 deg C. Decontamination factors of greater than 10,000 from uranium, plutonium, and common metallic contammants, greater than 25,000 for fission products normally expected in the feed (mainly Zr-Nb with some Ru-Rh), and greater than 1000 for thoriuin are obtainable under proper operating conditions. Because of the low processing rates, the necessity for carrying out the absorption cycle at 25 deg C and the absence of radiation daraage problems, Dowex 1, X-4 (50-100 mesh) or Dowex 21K (50-100 mesh) resins are considered the best choices for this application. Gassing occurs with the use of ferrous sulfamate - semicarbazide reductant but is not a serious problem and …

A laboratory study of a gelatin headend process for feed from silicon- containing aluminum fuels and plant salvage solutions is described. The optimum conditions for the gelatin treatment of fuel solutions were to boil a 0.1 to 0.5N nitric acid solution with 100 milligrams of gelatin per liter for 30 minutes. This treatment improved filtration rates and decreased the surface activity of the filtrate for TBP extraction. A number of possible flowsheets for fuel solutions are presented using gelatin treatment and filtration. An adequate treatment was not found for salvage solutions of unknown composition because a gelatin dosage which was satisfactory for all solutions could not be selected. The optimum treatment for a salvage solution which was grossly contaminated with zirconium, soluble and colloidal silica, and dibutyl phosphate was to boil a 1N acid deficient solution with 600 milligrams of gelatin per liter, filter, and use a Hexone extraction system. A silicic colloid in fuel processing solutions was characterized as a surface active material by this study. (auth)

From introduction: This report is a summary of the detailed geologic and engineering information that is available on three sites for underground explosions on the Nevada Test Site.

From introduction: The papers comprising the various parts of this report contain the preliminary results of the U. S. Geological Survey investigations in the Ul2e.05 tunnel at the Atomic Energy Commission's Nevada Test Site, Nye County, Nevada (fig. 1). Reports on electrical resistivity, natural radioactivity, and heat required to raise the rocks to 100C will be issued later. A preliminary report on the geologic effects of the Blanca event is being prepared.

Abstract: The theoretical response of a low-geometry scintillation counter to any photon-emitting nuclide whose decay scheme is known is developed. It is used to compute the counting rate, as a function of time, of (1) individual and total fission products resulting from. 10^4 simultaneous slow-neutron fissions of U^235 and (2) several other nuclides for 10 initial atoms. The calculations extend from ~45 min to 301 days after fission.

Several functions used in the calculation of x-ray crystallographic intensities are tabulated over large ranges. These tabulations include Lorentz- polarization factors as a function of Bragg angle, the Debye function as a function of THETA /T, and the Debye-Waller temperature factor as a function of B for selected sin theta / lambda values. (auth)

The equations and nomographs most frequently used intended calculating behavior of stack effluents are given and explained. (T. R. H.)

Nuclear fuel cost of 1.25 mills/kwh would make nuclear power competitive with conventional power in lowcost coal areas if capital and operating costs can be brought to within about 10 percent of those of coal-fired plants. Substantial decreases in fuel fabrication cost are anticipated by 1970: other costs in the fuel cycle are expccted to remain about the same as at present. Unit costs and irradiation levels that would be needed to give a fuel cost of 1.25 mills/kwh are believed to be attainable by 1970. (auth)

Foreign research and power reactors are tabulated. Nuclear power buildup goals are given for each nation on which information is available. (J.H.D.)

The energy distribution of ions and electrons in DCX are being studied by means of the Fokker-Planck approximation to the Boltzmann equation. An IBM- 704 code, called FOPP, was constructed to solve simultaneously the coupled Fokker-Planck equations for each of the two species of particles. This report discusses the difference scheme employed and derives the boundary conditions necessary in order that this difference scheme conserve energy and particles in the absence of sources and sinks. In particular, detailed discussion is given of problems arising from the use of two grid sizes, which proved advantageous on account of the great difference in the mass of ions and electrons. (auth)

A process has been developed for converting nitricacid-insoluble uranium in ore concentrates into soluble form. Ore concentrates are treated with a reducing agent such ss carbon monoxide or hydrogen at temperatures or 670 to 730 C. In the laboratory, retention times nsoluble uranium vary inversely with the concentration or the reducing agent. Laboratory studies leading to the development of the process are reported. (auth)

Progress on various theoretical studies are reported. DC model magnet studies to determine the optimum size and position of correcting holes drilled in the magnet yoke are described The major design criteria for the ring magnet are given. The current in the magnet of the 12.5 Bev proton synchrotron was calculated. The power supply characteristics and operating conditions are described. The side coil assembly phase of the ring magnet ooil program is described. The radiofrequency system consisting of accelerating cavity, rndiofrequency program, noise-measuring systems, power amplifier, and beam inanction electrodes is discussed Detailed studies were made of the BNL Linac design and such modifications were made as are required to meet the specific needs of the ANL injector system. Two methods of fabrication of the ring magnet vacuum chamber are discussed. A preliminary study and desige was made of several vacuum pumpimg systems for the ZGS accelerator. Investigations on the vacuum properties of epaxy plastics, evapor-ion pumps, water flow control, and foundations are reported. (For preceding period see ANL-5564.) (W.D.M.)

A fuel-element assembly from the first loading of the OMRE was examined in detail after experiencing an average uranium burnup of between 1 and 2 at.%. The rate of decay heat generation was evaluated by temperature monitoring of the shipping-cask coolant. Temperatures of the fuel-element-box assembly and the fuel plates were measured with thermocouples and tempilstiks. Structurally, the fuel-element assembly was affected very little by either radiation or the organic coolant-moderator. Although there was some distortion in the side and end plates of the assembly, the coolant channels between the fuel plates were free from major fouling and obstructions. The channel cross sections were reduced at specific points less than 5 per cent. The plates studied were subjected to complete gamma scanning. Specimens removed from selected areas of the scanned plates were radiochemically analyzed for burnup and the results correlated with the gamma-scan data. Burnup profiles were constructed for each of the scanned plates. The gamma-scan data were also utilized to determine the average plate burnup. (auth)

A study was made of hydraulic characteristics of Yankee-type fuel rod assemblies using experimental and analytical methods. Two scale model fuel assemblies utilizing both ferrule and strap type arrangements were constructed and tested at atmospheric pressure and room temperature. Analytical methods using semiempirical relationships are substantiated by experimental results for both the fuel assembly having strap-type spacers and the fuel assembly having cylindrical ferruletype spacers. The experimental pressure drop across the assembly model using either straps or ferrules correlated within 5% of the value calculated by means of equations based on the equivalent diameter concept for flow inside pipes. The individual frictional drops along the rods and across the end plates and straps correlated within 15% of the predicted pressure drops. The indlvidual pressure drops across both the staggered ferrule sections and the full ferrule section correlated to within 17% of the predicted pressure drops. Comparison of the ferrule and the strap pressure drops indicates that the pressure drop across a level of straps was more than four times the pressure drop across a full ferruled section. It is concluded that the analytical methods based on the equivalent diametcr concept can be satisfactorlly used to calculate pressure drops for flow …

This annual report of Brookhaven National Laboratory describes its program and activities for the fiscal year 1958. The progress and trends of the research program are presented along with a description of the operational, service, and administrative activities of the Laboratory. The scientific and technical details of the many research and development activities are covered more fully in scientific and technical periodicals and in the quarterly scientific progress reports and other scientiflc reports of the Laboratory. A list of all publications for July 1, 1957 to June 30, 1958, is given. Status and progress are given in fields of physics, accelerator development, instrumentation, applied mathematics, chemistry, nuclear engineering, biology, and medical research. (For preceding period see BNL-462.) (W.D.M.)