Abstract: A study, consisting of field and laboratory work, was undertaken in an endeavor to establish possible structural mineralization controls associated with the copper-uranium occurrences in the Willaha area, Coconino County, Arizona. Uranium mineralization, apparent at present, is localized along small fissures and vugs and in certain beds and lenses of the middle member of the Kaibab formation(Permian). It is associated with copper and iron oxide staining. Though no definite ore controls were disclosed by this study, at least a limited program of shallow drilling is warranted on the property. This sub-surface exploration should determine possible extensions of known mineralized areas, explore surface radiometric anomalies, and provide data for the determination of possible guides to ore. Deeper exploration may encounter mineralization in other horizons of the Kaibab limestone.

"A design study is presented for a sodium cooked, graphite moderated power reactor utilizing low enrichment uranium fuel. The design is characterized by dependence on existing technology and the use of standard, or nearly standard, components. The reactor has a nominal rating of 167 thermal megawatts, and a plant comprising three such reactors for a total output of 500 thermal megawatts is described. Sodium in a secondary, non-radioactive, circulation system carries the heat to a steam generator at 910 degrees F and is returned at 420 degrees F. Steam conditions at the turbine throttle are 600 psig and 825 degrees F. Cost of the complete reactor power plant, consisting of the three reactors, and on 150-megawatt turbogenerator, is estimated to be approximately $43,165,000."

"The Atomic energy Commission has undertaken a development program to provide the technology needed for the evaluation and economic design of nuclear power plants. This program is to be carried out during the next five years at several national laboratories and industrial organizations. The Sodium Graphite Reactor (the SGR) is one of those to be investigated and experimentally tested as part of this 5-year effort. The program on the SGR is intended to expand our area of information covering sodium-graphite technology, experimentally demonstrate the feasibility of this reactor complex and extend its performance limits, and apply in information developed to designs suitable for the full-scale nuclear power plant. As a principal part of this program, a Sodium Reactor Experiment (the SRE) is to be constructed and operated; it will be the major experimental facility in which the performance of this reactor will be studied and new technological advances tested. This report continues an earlier series 2-7 in which previous work on the SGR and the SRE has been described. In this report, the progress on the program is described in two main sections. Section A is devoted to work relating to the general technology of Sodium Graphite Reactors, and to …

A membrane demineralizer has been tested in a pressurized loop. The loop water resistivity was maintained in the 1-2 megohm range by ionized solid transfer in the demineralizer. The size and power requirement of the unit tested per gpm flow through the unit, were 2.3 cu ft and 100 watts. In view of the fact that present designed could reduce the size and required maintenance, further studies may be warranted.

From Introduction: "The self-shielding characteristics of a number of absorbers and of U-235 and U-238 have been investigated in PPA-20 (66°). Most of the measurements have been made in Ring 3 at the core mid-plane. The foil doubling technique was employed. Particular attention was given to 'thin' foil regions for the purpose of extrapolating the data to zero thickness."

Purpose: "In order to determine the optimum procedure for sampling the coarse-feed tank (C-102) of the NP and MTR process, a systematic series of tests has been conducted to measure the extent of solution stratification and effect of air-sparging as a means of vessel homogenization."

This technical report includes (1) World-wide Network (Maps). (2) Fallout of Sr90 at selected sites during June and July. (3) Comparison of Sr90 collection by gummed paper and pot on the roof of the New York Operations Office March through July. (4) Sr90 contamination of cow's milk in Metropolitan New York. (5) Sr90 content of the upper air prior to Castle. (6) Sr90 contamination of pooled fetal bones collected during June from the Metropolitan Area. (7) Sr90 contamination of water supply in the Metropolitan Area June and July.

An addendum to report HW-30390 a additional power generation calculation for P, the rate of energy release, in kilowatts, was derived.

"A study has been made of the distribution of tracer fission products and plutonium between small samples of molten uranium and solid uranium oxide, carbine, and nitride. The distribution showed the same behavior i general for all three materials: 1. The rare earth elements, Cs, Ba, and Sr were extracted primarily into the solid scrub phase. 2. Zirconium and Nb partially concentrated in the scrub phase. 3. Plutonium, Mo, and Ru tended to remain completely in the metal phase. The distribution of activities agreed with trends predicted from the thermodynamic data. Uranium oxide appeared to be the most desirable scrub material for removing large amounts of fission products from the uranium while leaving beind the Pu. In addition the uranium metal was not severley contaminated by dissolved oxide."

"A method for solving multi-group reactor equations which arise in the diffusion approximation is outlined. Considerable work has been done on this problem at KAPL and ORNL. Their approach is to replace the differential equations by difference equations. Complications arise in this method where more than one slowing down medium is present since the fluxes are discontinuous at the interfaces. The primary purpose of this article is to develop an exact integral expression for the neutron flux which automatically satisfies the boundary conditions. An iterative method for obtaining the fluxes and critical neutron multiplication ratio based upon the above-mentioned integral expression is given. The only approximation used in obtaining the fluxes, in addition to the use of multi-group diffusion theory as the basic model, is the use of numerical integration to evaluate the analytic expression. The equations for a two region, two group spherical reactor are given in a form suitable for machine programing. The extension to more than two regions is also considered.