Figures have and can be generated that indicate a higher charge-discharge rate if required before 105-N will be comparable with existing reactors. Also, these figures show an apparent operating cost incentive to increase the charge-discharge rates proposed for 105-N. Although these figures may be true by themselves, other figures developed from the same information and stated on a basis that affords a true comparison, show that the proposed rates for 105-N are compatible with those in existing reactors. However, the accomplishments of existing reactors should be considered as a guide only and not as Criteria since the design basis has already been established for Project CAI-816. An average charge-discharge rate has been proposed for 105-N that is compatible with the two main ground rules of the Project. Namely, the capital cost limitation and the plant factor. This rate of 8 tubes/hr. is one that appears to be reasonable from the charge-discharge design aspects and there is a good possibility that it can be increased with operational experience.

The purpose of this document is to update 190-C pump capacity information previous released in HW-52449{sup 1} and HW-58580{sup 2}. Improvements in motor cooling has resulted in raising the previous 3500 HP limit to 3660 HP{sup 3} thus increasing total pumping capacity.

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Activities at the Michigan Reactor and pneumatic tube system are outlined. Various modifications including installation and operation of a "bunny" rabbit pneumatic tube system for transferring samples from laboratory to counter are described. Prelimirary investigation for design of a neutron generator for activation work is described. Modifications to the 100-channel pulse-height analyzer including addition of auxillary circuits and additioral detectors are sumnnarized. In nuclear chemistry a cyclotron bombardment to produce long-lived vanadium tracer for yield determinations in activation analysis is described. Summaries of project work which have been published on absolute (d, alpha) reaction cross sections and excitation functions, use of computers in nuclear data analysis, and research on Ir/sup 196/ are presented. Study of the gamma spectra of 33-second Kr/sup 90/, 41-second Xe/sup 139/, and some longer-lived fission product rare gases is reported and a summary of preliminary results is given. Detection of a long-lived isomer of Ag/sup 108/ in old Ag/sup 110m/ samples is reported and new values for the branching ratios of 2.4minute Ag/sup 108/ are given. Activities of the Subcommittee on Radiochemistry are summarized. Development of a radiochemical procedure by the use of amalgam exchange is reported and exchange of the elements Cd, Tl, Zn, Pb, …

The disassembly of the `Adams` primary was scheduled for April 28, 29, and 30, 1959. The method of disassembly is provided as a procedure to be accomplished in order and the time and initials of the person accomplishing each step recorded.

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Performance characteristics of two activated charcoal columns at room temperature in separating fission-product xenon from an air stream were investigated by installing each column in the exhaust from an enclosure in which irradiated slugs were dissolved. Breakthrough curves are presented and the variation in xenon concentration within the columns is examined. Theoretical treatments of adsorption columns in the literature are found to agree well with the experimental data. Performance of the colunms is evaluated in terms of concentration factor'' and number of effective theoretical plates. (auth)

The magnetic jack is a device for positioning the control rods In a nuclear reactor, especially in a reactor containing water under pressure. Magnetic actuation precludes the need for shaft seals and eliminates the problems associated with mechanisms operating in water. It consists of a pressure shell, four sets of external stationary magnet coils (hold, grip, lift, pull down), and one Internal moving part (ammature) that impants linear motion to a cluster of rods. (W.L.H.)

It has been proposed that Pu-238 be produced by irradiating neptunium solution in one or more loops in a reactor and then recovering the Pu-238 in a close-coupled separations plant. Such a scheme could replace the more conventional scheme of solid element fabrication, irradiation, and reprocessing for plutonium and neptunium recovery. This document presents the advantages of such a scheme from the standpoint of product purity and Pu-238 production.

This report addresses the AEC symposium on particle-fluid mechanics

The Advanced Engineering Test Reactor (AETR) nuclear mockup is designed to be a flexible, inexpensive developmental facility which duplicates the reactor portion of the AETR and which would be used to verify the operation of reactor controls. The mockup would also furnish measurements of temperature and void coefficients, flux shapes, and critical mass, and facilitate a reliable AETR design in a minimum time, and with minimum development cost. For even greater usefulness, the mockup is also designed for use in conjunction with an operating AETR to check the reactivity of experiments and fuel assemblies, and for other annular core reactor development purposes. In these respects, the mockup design embodies the concepts for a very low power nuclear auxiliary outlined in an earlier report. Detail design was limited to the reactor assembly and control systems and it is assumed that a simple building with supporting facilities may be readily supplied by private industry or the AEC. Results of negotiations with potential vendors and fabricators, a description of the facility components, and design drawings suitable for contractor use are presented. Preliminary quotations from fabricators and suppliers indicate that the facility can be in operation within an eight month period at a total …

The variations of size and density of aerosol particles were examined. Experiments were made with a wide variety of substances using a modification of Whytlaw-Gray and Patterson's methods in which freefall velocity was determined. The mass was estimated both by balancing the gravitational and electrostatic forces and from rising velocities under constant field strength. The results were compared with the shape and structure of the aerosols shown by electron microscopy. The variation of the Stokes' (or drag) diameters was found by combining the results with cascadeimpactor and microscope measurements. A modification of the Millikan oil drop apparatus was used. It was concluded that the free-fall velocity of an agglomerate is proportional to the two-thirds power of its mass, and when primary particles are not uniform in size and shape, wide variation in agglomerate densities occur. It was also noted that when primary particles have irregular shapes there is a difference in drag diameter of the agglomerate determined from the rising and falling velocities. This is a factor in electrical precipitation of such aerosols. (J.R.D.)

It is shown that cryogenic techniques offer the possibility of substantially improving the efficiency and practicality of generating high magnetic fields in air-core coils of large size. Over-all reductions in power requirements of as high as 25, by comparison with conventional coils, are predicted, provided high purity conductors and efficient refrigeration cycles are used. (W.D.M.)

This report addresses air-core strong focusing synchrotron.

This report supersedes K-1-3425 and LA-2327(Prelim). The first experiment (LAMPRE I) in a program to develop molten plutonium fuels for fast reactors is described and the hazards associated with reactor operation are discussed and evaluated. The reactor desc=iption includes fuel element design, core configuration, sodium coolant system control, safety systems, fuel capsule charger, cover gas system, and shielding. Information of the site comprises population in surrounding areas, meteorological data, geology, and details of the reactor building. The hazmalfunction of the several elements comprising the reactor system. A calculation on the effect of fuel element bowiing appears in an appendix. (auth)

Using a special reactive form of high-purity alumina, claddings were applied to UO/sub 2/ particles by a tumbling technique. The clad pellets were isostatically pressed at 100,000 psi and then sintered at 2800 deg F in hydrogen. crack-free spheroidal pellets ranging from 1000 to 2000 mu in diameter were produced. The dense Al/sub 2/O/sub 3/ envelopes surrounding the UO/sub 2/ particles were estimated to be 300 to 500 mu thick. The Al/sub 2/O/sub 3/ claddings protected the UO/sub 2/ from oxidation when the pellets were heated in air for 100 hr at 1200 or 1800 deg . There was no measurable release of fission products from irradiated clad particles heat treated at 1700 deg F in vacuum for 7 days after exposure to 6.0 x 10/sup 12/ nv for 1 hr at room temperature. Claddings of other oxides, such as Beo or MgO, probably could be applied by the same techniques used in applying the Al/sub 2/O/sub 3/ claddings. (auth)

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The demand plot has a 5-set, modified pump decay curve; it shows that 20,000 gpm emergency flow would be required within 80 seconds of complete pump power failure. Bases for the demand curve are constant bulk inlet temperature of 2 C, constant bulk outlet temperature of 95 C, K-3 I&E fuel elements, and initial reactor flow of 188,000 gpm.

This production test was designed to evaluate the suitability of low hydrogen dingot uranium as routine process material. Nine tubes of I and E fuel elements (6 dingot, 3 ingot) with 32 fuel elements in each tube, have recently been discharged at the C Reactor and this document contains the results of analyses made on the dimensional stability properties of this material.

A technique based on a determinaiion of the differential counting rate exhibited by the 184-kev gamma radiation associated with the decay of U/sup 235/ was developed for the determination of the U/sup 235/ content in Argonne Low Power Reactor fuel element core blanks. The Argonne Low Power Reactor core blanks were an aluminum-highly enriched uranium alloy containing 17.5 weight per cent uranium (approximately 4 g U/sup 235/) having the following dimensions: length, 6.875 inches, width, 3.31 inches, and thickness, 0.200 inch. The gamma- ray spectrum emitied by uranium is rather complex. Using a scintillation spectrometer and scanning the spectrum, the energy is found to be concentrated primarily in two regions, at 184 and 90 kev. The 184-kev gamma rays result primarily from the decay of U/sup 235/ The gammas in the 90-kev region result from the U/sup 235/ decay and daughter products of U/sup 238/ and U/sup 235/. Using a pulse-height analyzer, it is possibie to select the desired radiation emitted from the source and determine the counting rate for a given source. In this work the 184-kev gamma radiation was counted to determine the amount of U/ sup 235/ present in the individual core blanks. (auth)

Some calculations have been made to estimate the characteristics of a neutron-burst type fast reactor similar to Godiva but made up of relatively small component parts--the so-called "layered assembly." One spherical and three cylindrical assemblies have been considered. Critical masses, assuming 5% voids, range from 58 to 65 kg of 93.4% enriched U/sup 235/. For a reactivity addition of 0.33 dollars above prompt criticals bursts between 2 x 10/sup 17/ and 6.7 x 10/ sup 17/ fissions were computed with accompanying temperature rises varying from 514 to 1600 deg C. The burst width at half-maximum was about 12 microseconds. To obtain an idea of the possibilities of stress reduction which might be achieved by layerings an assembly made of small rings was considered. While the critical masses obtained here are believed to be fairly accurates the predictions concerning mechanical energy generated, total fissions, and burst width may be subject to sizeable error due to the many simplitications required to allow hand computations. Neverthelesss considerable improvement in safety and burst-size is indicated by the use of a "layered assembly" instead of an assembly composed of relatively thick parts. (auth)

The expected maximum contamination of the HNPF cold traps and primary sodium pumps was determined along with the maximum dose rates from these components during removal and shipping. Suitable shielding for casks to be used in the removal operation and for shipping these components away from the reactor site is specified. Access to an unshielded cold trap is limited by high dose rates, i.e., 100 mr/hr at 120 ft, after 180 days decay time. A handling cask providing a radial shield of 3 in. of lead will provide adequate personnel protection for the removal operation, if 180 days decay time is allowed before the trap is removed. An additional 2.4 in. of lead is required for offsite shipment of the cask. This additional shielding can be added after the trap is removed from the reactor building. Dose rates from the cold trap after the shield plug is removed from the access hole are shown. If direct line-ofsight exposure is avoided, dose rates to personnel will be below 100 mr/hr at any position, and below 10 mr/hr at distances greater than 20 ft from the access hole. Dose rates from the cask during its travel away from the hole, will be …

Methods are presented for the analysis of 41 refractory materials. An evaluation of the accuracy and the precision of these techniques are also given The materials studied are the borides of hafnium, molybdenum, niobiumL rhenium, tantalum, thorium, titanium, tungsten, uranium vanadium, and zirconium; the carbides of hafnium molybdenum, miobium, silicon, tantalum, thorium, titanium, tungsten, uranium, vanadium, and zircomium; the nitrides of boron, hafnium, niobium, silicon, tantalum, titanium, uranium, and zirconium; the silicides of molybdenum, rhenium, tantalum, titanium, tungsten, vanadium, and zirconium; and mixed carbides of uranium with hafnium, niobium, tantalum, or zirconium. (auth)