The calculation of shielding the thickness for plutonium is complicated by the many different energies represented in the gamma radiation emitted during decay of the plutonium isotopes. Dose rate predictions are also frequently confused by gamma from varying content of fission product impurities in the plutonium, as well as other gamma radiation induced through alpha and neutron particle absorption within the source material or its environment. After assumptions are made for these many factors the radiation data for shielding determination is still frequently inadequate because of wide variations in dose rates resulting from self-absorption. The degree if self-shielding is in turn dependent on nature of the plutonium compound, degree of compactness, weight, and over-all geometrical distribution of the source material. By preparing a variety of plutonium samples representing combinations of these varying factors, actual dose rates and gamma spectra, as obtained from them, can then be extrapolated for application to specific situations.

The Non-Production Fuel Reprocessing Program involves the chemical processing of valuable reactor fuels received from privately owned power reactors. It is necessary therefore, to accurately measure the fuel material received in order to insure proper payment to reactor operator and to provide the Atomic Energy Commission with appropriate accountability data. The volume measurement study described herein was conducted in order to determine the limits of accuracy that could be obtained in measuring relatively large volumes of solution under plant processing conditions.

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 …

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.)

The following document was done under the auspices of the U.S. Atomic Energy Commission with the intention of analyzing air core cryogenic magnet coils and its usage in fusion research and high energy nuclear physics applications.

Routine monitoring procedures are summarized. Based on samples collected by continuous air monitors, the wcekly average value for air contamiration in the laboratory area was 1.98 x 10/sup -12/ mu c/cc. Radioactive fall-out was determined by measuring the total activity and total number of particles collected on gum paper fall-out trays. Two peak periods of activity in July and September were concluded to be the result of weapons tests. Meteorological data are tabulated. Data are included from a survey of natural radioactive elements in TVA lake waters, activity levels in samples of river and lake waters in the laboratory drainage area, rain water, and laundry decontamination measuremcnts. The combination film dosimeter-personnel identification badge was redesigned. A drawing of the badge and its component parts is included. Design modifications are described which wcre made in an automatic air sampler rized from radiation surveys. (C.H.)

From summary: This report summarizes the work performed in a brief study made for the AEC of boiling water reactor power plants in the 5 to 40 eMW output range.

"In this paper the single particle motion in the magnetic field created by a circular array of circular current loops is investigated." (p. 3)

From foreword: This report describes the work accomplished including the Conceptual Design task and the initial results of the tasks on Facility Design and Fabrication, Fuel Fabrication and Evaluation, and Fuel Reprocessing.

Surface active materials in aluminum nitrate-nitric acid fuel reprocessing solutions were characterized. Polymerized silica, zirconium- modified silica and soluble dibutyl phosphate species were found to contribute to stable emulsion formation. These surfactants were reduced in effectiveness by added acid. (auth)

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The so-called Collisionless Boltzmann Equation'' or Vlasov Equation'' has been derived previously by Harris by use of a complete statistical treatment of both the plasma particles and the electromagnetic fields. It is shown that a consequence of this approximation is that the entropy of the electromagnetic field as well as that of the plasma particles is a constant. This result is used to demonstrate that only completely the Vlasov equation. (auth)

Methods for analyzing chemical impurities in sodium metal samples are presented. Chemical analysis was used to determine impurities in calcium, carbon, chromium, iron, lithium, nickel, oxygen, potassium, and zirconium. Spectrographic analysis was used to determine other impurities. Sodium samples obtained from experimental apparatus were analyzed by these methods. (auth)

Methods for analyzing chemical impurities in sodium metal samples are presented. Chemical analysis was used to determine the following impurities: Calcium, Carbon, Chromium, Iron, Lithium, Nickel, Oxygen, Potassium, and Zirconium. Spectrographic analysis was used to determine many other impurities. Sodium samples obtained from experimental apparatus operated as part of the work being conducted for Atomics International were analyzed by these methods.

On a project of the magnitude and complexity of the Plutonium Recycle Test Reactor is was considered prudent engineering practice to obtain an independant design review by a competent off-site group that had not participated in any way in the development of the design. Accordingly, the Atomic Power Equipment Department of the General Electric Company was requested to make such an engineering review, and an Assistance to Hanford contract authorizing this work was approved by the Atomic Energy Commission in November, 1958.

Property changes to common graphites resulting from reactor irradiation depends strongly on the temperature of exposure. Radiation affects at 30 C and 400 to 500 C are quite well known; however, at intermediate temperatures, results are not well known although a large volume of graphite in most reactors operate in this range. Production test, IP-22A, was reactivated, after a dormant period of many months, to study radiation damage between 30 and 400 C. This report concerns mainly the experimental details and evaluation of the equipment necessary for the controlled temperature irradiation.

A core parameter study of the operating costs was performed for a 300- Mwe sodium graphite reactor, a scale-up of the Hallam Power Reactor. The results of the study indicate that the core design is nsar optimum and that core modifications would reduce the power costs by less than 5%. The lattice spacing, fuel rod diameter, and sodium flow can be varied within a rather broad range without significant changes in power generation costs. The effect of the fuel cladning thickness is more significant; fuel cycle costs can be reduced if stainless steel canning is replaced with zirconium canning. Use of UC in place of uraniummolybdenum fuel would also permit cost reductions. (D.L.C.)

Submitted as a thesis by William Berdell Barlage, Jr. Heat transfer studies were refined and extended to the point where the results could be generalized into a mathematical correlation valid over a reasonably wide range of variables. An extended study was made of the rate of heat transfer between benzene and water in direct contact in a pulse column, and of the operating variables which controlled the behavior. In this mode of heat transfer the benzene was dispersed as a family of droplets into the water, which formed the continuous phase. The pulse column was 1.92-in. in inside diam., and contained 8 stainless steel perforated plates spaced 2.00 inches apart. The plates were perforated with 1/8-in. holes in a triangular 60 deg pattern, providing 24.6 per cent free area. The equipment was carefully insulated and modified in many details in order to permit precise heat transfer measurements. The operating conditions were varied so as to provide wide ranges of variables and heat transfer responses. (W.L.H.)

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A complete algebraic analysts was obtained for the variation of the steady state ion density n+ with injected current I in an OGRA-tane fusion device (i.e., a device based on trapping of ions by breakup of energetic molecular ions on collision with either the background gas or trapped ions). The most general variation of n+ with I is shown to be an s-curve with at most three rots of n+ for a given input I. A physical interpretation of these three roots is given. In addition algebraic expressions are obtained for the two currents at which the bends in the s-curve occur. It will be necessary to attain the larger current in order to build up a high density plasma when the density ia being increased from below. On the other hand, once the high density has been achieved it may be maintained by steady injection of a current larger than the lower value. Parameters corresponding roughly io the specifications of OGRA are used io obtaam some numerical results. The previously published formulas for burnout in DCX are extended to include effects of neutral backstreaming from the input beam and "ion- pumping." (aath)

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.

From Introduction: "In this monograph, the background of the exponential reference atmosphere will be outlined and a set of radio ray tracings presented. These ray may be used for the solution of many problems involving refraction by means of the exponential reference atmosphere."

The Atomic Energy Commission has assigned the Hanford Atomic Products Operation the responsibility of providing the technology and facilities for the interim reprocessing of slightly enriched uranium fuels discharged from power and propulsion reactors. The fuel from the various reactors vary greatly in physical dimensions and in the composition of the core and cladding. For the purposes of chemical reprocessing, the fuels may be divided into the three main categories of aluminum, Zircaloy or stainless steel clad elements. The fuels may be sub-divided by the nature of the core material; e.g. metallic uranium, uranium dioxide, uranium-molybdenum alloys uranium-aluminum alloys, etc.

The PRTR pressure loop is an experimental facility to be installed in the Plutonium Recycle Test Reactor, Building 309, for use in testing the feasibility of reactor cooling systems utilizing high pressure water or steam as the coolant. It will furnish operating experience with such systems and serve to test fuel elements and components designed to operate in such environments.