Report containing hazard and safety information regarding the Experimental Breeder Reactor-II in Idaho.

The ACIR Library is composed of publications that study the interactions between different levels of government. This document addresses alternative approaches to governmental reorganization in metropolitan areas.

"An Automated Spectrograph has been developed for the analyses of beryllium contained on filter paper discs. With this instrument, unattended analyses can be carried out on a number of samples following simple sample preparation. Filter paper discs representing beryllium in the atmosphere as well as those used for evaluating surface contamination are accommodated. Included in the uniting are an automatic electrode changer, analytical gap servo control, spectrograph, signal detector, and computer and readout system. The performance of the instrument in the range of 0.5 to 40 mg of beryllium is discussed."

Introduction: The primary objective of the work of this report was research and development leading to large-scale applications of low level isotope tracer technology.

The summary of this report is that a substance isolated from Chlorella Pyrenoidosa metabolizing {sup 14}CO{sub 2} in the light, previously believed to be a diphosphate ester of a 2-carboxy-4-pentulose, has now been shown to be a disphosphate of 2-keto-L-gulonic acid. The phosphate groups appear to be attached to two of the carbon atoms 3-6. Evidence is presented suggesting that this compound arises from glucose, or a glucose phosphate, which is not in rapid equilibrium with photosynthetically produced glucose derivatives.

Additional experiments conducted on nitridation of irradiated U-fissium fuel pins revealed that irradiation does not greatly affect the nitridation rate at 300 deg C. In skullreclamation development, a phase separation common to both the blanket- and skull-processes was investigated in which a 50% Mg-- Zn supernatant solution was removed from precipitated U metal. In most runs the supernatant phase was removed with negligible U entrainment. The reduction rate of ZrO by Zn--Mg solution under skull-recovery process conditions was found to be lower than that of U oxides. It may be possible in this process to effect some Zr separation by limiting the reduction time to that necessary for U. Methods of procesging EBR-II fuels are being investigated to establish methods of separating rare earths from Pu. Development work on preparation of UC by addition of C to U dissolved in liquid metal media showed that the limited addition of alkali metals improves C wetting and the tests showod high C-to-U ratios and high O/sub 2/ contamination; procedure and equipment improvements are being made. Studies are in progress to evaluate the compatibility of various materials with the liquid metal-salt systems contemplated for reactor fuel reprocessing. Tungsten appears to have high corrosion …

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This program of the GEORGE Assembly Routine (GAR) will accept Fortran- like statements from paper tape and create the GAR language program on tape. This includes the needed calls for common subroutines and the reservations for the named variables and temporaries. The original statements in Fortran are carried along as remarks. The GAR language program may then be processed in the usual way by the GEORGE Assembly Routine, giving machine-language code. The level of sophistication of the source language is roughly equal to that of Fortransit or SALT. (auth)

Monthly newsletter of the High Plains Underground Water Conservation District No. 1, discussing the field of underground water. Topics include profiles of water conservation research, annual pre-plant soil moisture survey data, annual Winter Water Level measurement data, and information about the latest water conservation tips.

The Darex process developed for ihe recovery of U from stainless steel- containing reactor fuels consists of dissolution of the fuel material in dilute aqua regia, removal of chloride from the solution to prevent corrosion of downstream stainless steel process equipment, and adjustment of the nitrate solution to solvent extraction feed conditions. Each step can be either continuous, semi-continuous, or batch with continuous operation showing much higher throughput for comparable equipment. The preferred dissolvent is 5 M HNO/ sub 3/-2 M HCl, since dissolution rates and metal loadings are near maximum. Nitric acid from 60 to 95 wt% can be used in decreasing ihe chloride concentration to <350 ppm; ihe higher strength acids have process advantages. Excess nitric acid is recovered and recycled during produciion of a concentrated metal-salt solution, which is diluted io Purex solvent extraction feed acidity, 2- 3 M HNO/sub 3/. Titanium is a satisfactory material of construction, wiih corrosion rates <l mil/mo in all process environments and over-all heat transfer coefficients comparable to those of stainless steel. (auth)

Concentric tubular reactor fuel element geometries to give equal coolant outlet temperatures are presented. Oscillations from tube to tube in thickness and temperatures generally occur but it is possible to eliminate them by choice of the centre element. This may be a fuel rod or a non-heat—producing rod with or without a surrounding annulus of fuel. The geometries and temperatures are dependent on the voidage and on a non-dimensional parameter equivalent to a Biot number based on the channel equivalent diameter.

From introduction: "The work described in this report represents the Joint United States - European Atomic Energy Community effort which is in keeping with the spirit of cooperation in contributing to the common good by the sharing of scientific and technical information and minimizing the duplication of effort by the limited pool of technical talent available in Western Europe and the United States."

Summary: "This report maps the distribution of soils which border the Mississippi between river mile 189 and Head of Passes in southeast Louisiana with special regard to their engineering significance. The subsurface disposition of depositional environments and their associated soil types are shown on 34 subsurface profiles. The text describes the physiographic and geologic development of the area studied, summarizes physical and engineering characteristics of the engineering soil types mapped, and discusses some of the effects of geologic factors on river migration" (p. vii).

"Mixed Fission Products (MFP) for use as a heat source for thermoelectric generators will become increasingly available in the coming years. The Atomic Energy Commission sponsored program on solidification of nuclear wastes is now entering the hot-bench scale test phase. During this phase approximately 5000 thermal watts of two year old MFP could be produced monthly. Two different types of hot calcination pilot plants are planned for installation at the Hanford National Laboratories in the 1964 to 1966 time period. Each of these plants should be able to produce 160,000 thermal watts of two year MFP and 16,000 thermal watts of ten year MFP on a monthly basis. A full scale plant for a 15,000 Mw(e) nuclear economy is estimated to produce four to five times as much MFP as either of the pilot plants. Costs will be dependent upon AEC policy in effect at the time the plant is operating."

Mixed Fission Products (MFP) for use as a heat source for thermoelectric generators will become increasingly available in the coming years. The Atomic Energy Conamission sponsored program on solidification of nuclear wastes is now entering the hot-bench scale test phase. During this phase approximately 5000 thermal watts of two year old MFP could be produced monthly. Two different types of hot calcination pilot plants are planned for installation at the Hanford National Laboratories in the 1964 to 1966 time period. Each of these plants should be able to produce 160,000 thermal watts of two year MFP and 16,000 thermal watts of ten year MFP on a monthly basis. During this phase, MFP costs should be less than 15 per ihermal watt for two year MFP and 50 for ten year MFP. This cost includes operation of the plant solely to obtain heat sources and sealing the MFP into fuel containers. A full scale plant for a 15,000 Mw(e) nuclear economy is estimated to produce four to five times as much MFP as either of the pilot plants. Costs will be dependent upon AEC policy in effect at the time the plant is operating. lf the policy indicates that the full …

The potential-current density relationships for 304 stainless steel dissolution in a nitrate system were studied as a function of solution composition and temperature in order to optimize the conditions for electrolytic dissolution of ihis material. In the nitrate system, the anodic dissolution of steel takes place in the transpassive region. Under some conditions, deviations from Tafel behavior are observed which depend greatly on the nitrate and hydrogen ion concentration, and on temperature. A discussion of passivity, transpassivity, secondary passivation, the limiting current density, and the effect of alloy composition on the dissolution behavior is given. It was found that at temperatures above 60 deg C efficient dissolver operation should be possible over a wide range of solution compositions and at current densities up to 2 amp/cm/sup 2/. (auth)

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In an effort to determine the environment to which the equipment designed by Sandia Corporation will be exposed, a &quot;Data Bank&quot; of environmental information was compiled. Measured quantities resulting from actual uses were continually being summarized.

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Report regarding experiments made to determine the fallout-radiation protection offered by four types of buildings in Los Angeles, California. The first structure was the Laboratory of Nuclear Medicine and Radiation Biology and the University of California at Los Angeles, the second was a fallout shelter, the third was the communications area of the Los Angeles Police Department building, and the fourth was a classroom.

A total of 11,117 enriched uranium-5 wt.% fissium alloy fuel pins was manufactured for EBR-II, Core I. These were made from a synthetic fission product alloy of nonradioactive elements, natural uranium, and enriched uranium. The material was supplied as precast billets. The manufacturing methods were developed for the EBR-III Fuel Cycle Facility. Experimental refabrication equipment was used to production test both the methods and the equipment. The billets were induction melted and pressure cast into precision-bore, high-silica glass molds in batches of 90 to 160. The number of molds used was adjusted according to batch weight. After casting, the molds were broken away and the castings were fed into a pin-process and inspection machine. Both ends were sheared from the castings to produce finished pins measuring 0.144 in. in diameter by 14.22 in. long. The pins were inspected for diameter, porosity, weight, and length. Rejected pins and sheared ends were broken into short lengths and returned for consolidation. Acceptable fuel pins were sealed and sodium bonded in stainless steel jackets, and assembled into Core-I fuel elements. (auth)

Use of Pu/sup 239/ -- U/sup 235/-fueled fast critical assemblies to estimate properties of Pu/sup 239/-fueled assemblies is of interest because of safety considerations and limited plutonium availability. Bare and reflected homogeneous cores and reflected two-region cores are considered. The fuel, 5% by volume, is assumed to be Pu/sup 239/ and U/sup 235/ of various fuel composition ratios for the homogeneous cores. For the tworegion cores the 5% fuel volume is Pu/sup 235/ in the central region and U/sup 235/ in the outer core region. Core diluents, simulating fertile, structural, and coolant materials, are assumed identical in all cases. it is estimated that construction of the reflected two- region core with ratio of central core region volume to total core volume of 0.1 will theoretically decrease the calculated error in prediction of the critical size of a corresponding solely Pu/sup 239/-fueled assembly by a factor of about 10 to 20. (auth)

The purpose of these tests is to provide necessary laboratory data in support of an R,PEO program in determining the energy loss associated with various hardware size combinations on the rear face of the B-D-F reactors. The original method used to check for critical flow was determined to be faulty. A revised method demonstrated critical flow did occur in the 5/8-inch inconel connector and combination 1 fittings. The remaining fitting combinations with the 5/8-inch inconel and 3/4-inch aluminum connector were not rechecked because of the reaming of the I.D. to permit the continuation of the original tests. During test number 6, audible cavitation was heard with the highest severity at a point midway between pressure points 3 and 4 on the connector. This condition appeared again in tests 6A, 7, and 7A, with incipient cavitation at approximately 40 gpm in each test, regardless of the rear header pressure and/or temperature.

Flow tests which were requested have been completed and the results of these tests are reported. Data which allow the determination of the normal operating flow rates for process tubes equipped with the various combinations of fittings tested is presented. Losses through these fittings, both with and without vaporization allowed, are shown and the relative contribution of the different fittings to the total pressure drop across the outlet assembly is also shown.