Introduction: "This is the ninth letter report on ARF Project C 127, entitled "Preliminary Studies of Scavenging Systems Related to Radioactive Fallout." This report covers the period from August 1 to October 1, 1959."

The following report discusses direct recording of nuclear pulse height data on magnetic tape that represents an inexpensive method of data storage where some degradation of the original energy resolution can be tolerated.

This report discusses the concept of an improved mass gauging technique with a scintillation counter as the radiation detector in order to meet industrial requirements.

The following document is an updated note on the process of building an improved nuclear density gauge within the time interval of September 2 to October 1, 1959.

Abstract: "A simple solution to Saha's equation has been obtained. The method of solution involves iteration with respect to the electron pressure or concentration and can be applied to the simultaneous calculations of any number of ions. Sample results are given for lithium and aluminum."

Abstract: A study has been made of the source, acceleration, and neutralization of ions for application to space propulsion. A summary is made in this paper of some of the most important problems in this area and their solutions.

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

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.

Abstract: "Preliminary design and experimental work aimed toward fabrication of large magnet coils with encapsulated sodium conductors is outlined. Aspects of cooling cycles are discussed along with heat transfer, thermal stability, and the mechanical reinforcement of such coils."

Abstract: "The effects of a number of ternary additions on the constitution, casting, and fabricating characteristics and the physical properties of aluminum-35 w/o uranium were investigated. Initial investigations were concerned with the effects of 3 w/o ternary additions on the microstructure and press-forging characteristics of the alloy. It was found that additions of this magnitude often introduced extrinsic phases in the alloy. At the 3 w/o level, additions of germanium, silicon, tin or zirconium inhibited the formation of UAl4 and thereby increased the content of the aluminum matrix in the alloy. It was also noted that these additions decreased the pressures required for extruding, and the tin addition also improved the homogeneity of cast shapes. Lead and palladium also improved the homogeneity of the cast material; however, neither of these was an effective inhibitor of UAl4 and free lead was detected in the alloy to which lead had been added at the ternary. From these studies it appears that tin and zirconium are as effective as silicon in enhancing the fabricating characteristics of aluminum-35 w/o uranium alloys, and may prove superior when evaluated on the bases of casting qualities and recycling characteristics."

From Abstract: "A literature survey was made of both welding and brazing methods that might be used to minimize the amount of exposed core material and provide the longest path for corrosion in the event of a defective weld in enextruded Zircaloy-2-clad uranium and uranium-2 w/o zirconium fuel elements."

Review of a study regarding a lid-tank shielding facility's properties, including fast-neutron and gamma dose rates and thermal-neutron fluxes, which were mapped in a symmetry plane. The removal cross section of gadolinium was determined through comparison with various other materials.

A Lucite model of a multi-stage countercurrent hydroclone solvent extraction apparatus has been constructed and tested with Amsco-water. The diffusivity of Cs 134 tracer in aqueous chloride solution was measured to check the performance of the capillary diffusivity measuring system. The experimental data from four Druhm runs showed that 1/8in. thick graphite liners are usable for reactor temperatures above the boiling point of sodium.

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.

Design Test Request PR-20 Calandria Characteristics, outlined the need for experimental data concerning the performance of the calandria under transient conditions. Test data was required to confirm that the moderator dump system would drop the level the required 24 inches in less than one second. The original calandria dump chamber design was modified until the criteria was met. This information is recorded in HW-58333, Interim Report, Design Test PR-20, Calandria Characteristics, which lists the drop for the first 24 inches only.

The Fermi age of thermal neutrons in a lattice containing both water and graphite in some sort of weighted average of the age in water and in graphite independently, with correction for volumes of non-moderating elements or voids. The correct weighting function has been in question during design calculations for the NPR. This paper presents a very simple and direct approach to the problem, resulting in a weighting equation which seems to be theoretically sound. Unfortunately, simple theories cannot be guaranteed to give good results in complicated systems; nevertheless, it appears that the weighting dunction derived here is to be preferred over methods involving empirical relationships which are of doubtful validity in the NPR geometry.

For purposes of an earlier consideration of the effects of in-reactor test loops on PRTR operation and program, the design of the high pressure, H20 cooled fuel test loop was assumed to be similar to that for the high pressure, gas cooled loop. Detailed design criteria for the H20 loop have recently been completed so that this assumption is no longer necessary. Therefore, to insure that the effects of all proposed in-reactor loops are fully evaluated with respect to Plutonium Recycle Program objectives, the present study has been carried out. Much of the qualitive discussion in the original analysis is still considered valid. Thus, rather than repeating the entire analysis as it pertains to the H20 loop, only those departures resulting from differences between the assumed design and the scope design are presented.

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.

This report is being submitted to provide information and data on the planning and installation of a new mechanized drawing system which utilizes a new drawing index system and a 35mm microfilm image in a standard IBM card. The communication of engineering information to date has been primarily the full size contact print produced by exposing and developing light sensitive material. This method produces legible copies; but, it has the following disadvantages : (a) the original drawings are fragile, and become dirty and worn through excessive use and handling; (b) the sensitized paper has a limited shelf life; (c) the reproduction cycle is slow because it requires hand feeding and adjustment of the reproduction machine to the condition of the original; (d) the prints are large and unwieldy to handle as working documents; and (e) the filling of originals and copies is slow and expensive. In addition, providing reference prints which are accessible to engineering personnel in the widely separated areas from a central file and reproduction facility is slow, expensive and time consuming.

Presently rotary drum filters are being used for plutonium oxalate slurry filtration in the 234-5 Building. Changing of the filter cloth used on a rotary drum is a time consuming operation which involves ever increasing radiation exposure to maintenance personnel. Consequently, studies were conducted in the 321 Building on a disk type filter which could be adapted for simple and quick ("one-nut") replacement of the filter medium.

"The factors of atomic size, ion core potential, and charge, which contribute to the increment in electrical resistivity produced by solutes in metallic solutions, were separated and experimentally evaluated in liquid sodium. The results are expressed in a semi-empirical relationship which defines the increment in resistivity of a monovalent solution as the sum of the fractional difference in atomic volume of solute and solvent and the difference in atomic number of Na and the solutes Li, K, Cs, Rb, Ag, and Au."

At the request of Process Control Development, Chemical Development Operation, statistical assistance was provided by the Operations Research and Synthesis operation in the design and analysis of an experiment to determine how accurately volumes of solution in a large process vessel could be measured. A detailed discussion of the vessel calibration tests, the associated liquid level measuring equipment and the conclusions thereof are presented in HW-62049. The main purpose of this report is to present the precision and accuracy statements obtained, the method by which they were calculated, and to point out the advantages of this method over other "least square" procedures.

Conceptual design studies conducted at HAPO indicate that the improved thermal efficiencies which can be obtained in a high pressure, high temperature, nuclear-electric power plant might reduce unit power costs substantially. The Plutonium Recycle Test Reactor (PRTR) is designed primarily to investigate various facets of plutonium recycle operation. In order to increases the usefulness of the PRTR as an experimental tool, it is desirable to include facilities for testing materials and fuel concepts under a variety of conditions, including the high temperatures and pressures which may be encountered in future water-cooled power reactors.

This report is an outgrowth of the MURA Users' Conference of June 1959. At that conference the group on beam separators discussed the problem of whether particle separation could be achieved at the machine energies under consideration. A preliminary version of the scheme outlined here was given at the conference. Later, after flaws were discovered, it was modified. The attempt is not to show that this is the way it should be done, but to show with reasonable certainty that there is at least one way it can be done.