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 following document analyzes radioactive spot contamination in and around the T and B Plant Areas.

An experiment now in progress should give some accurate information about the thermal neutron absorption cross section of carbon 12. This report outlines and summarizes this experiment and analyzes it to determine the main sources of error and the probably error in the final result.

From introduction: "It is the purpose of this report to give two further methods of determining end point of pontentiometric titration, neither of which require the plotting of a graph."

Objective: The objective of this study is to develop a simple method of either removing the C1- ion from the RAW or converting the ion into a less corrosive valence state.

The following report was made to determine the effective clarity improvement, flow rate characteristics, and operational problems encountered on passing Redox uranium feed solution through sintered, modified Type 304 stainless steel filter elements.

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.

From introduction: This report summarizes a study of the composution and physical properties of a number of hydrocarbon diluents together with some generalizations regarding relations between [vapor pressure, density and viscosity] properties.

Report from Hanford Laboratories concerning "the design and construction of the 100-N Reactor and heat dissipation plant complete with the necessary auxiliaries" (p. 2). Details of its construction and the plant's systems and instrumentation are described as well as economic considerations.

A continuous gamma-monitoring instrument capable of measuring the activity of the product of the Uranium Recovery plant has been developed and applied under process conditions. It has a detection limit of 15% and a normal range of up 1000% of the activity of natural uranium. Activity measurements made with this instrument have shown good correlation with laboratory determinations and are contributing to improved control of the solvent extraction process.

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.

Entrained solids have in the past been one of the major factors limiting solvent extraction fission product decontamination. These solids are known to be surface active, to absorb fission products, and to form interfacial films or "cruds". A major step in eliminating these cruds was taken when the Purex columns were modified to operate with the interfaces at the waste ends of the columns rather than the product ends. This type of operation tended to flush the solids out with the waste rather than the product streams.

This document presents a set of machine-computed curves involving a wide range of Kb and Kc, from which both thermal and resonance values of (E-1) may be obtained.

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.

The following report describes a successful procedure used for the determination of americium in the presence of plutonium. Provided are articles listed that describe the research done in developing the method.

The following report describes the developmental process of the determination of free nitric acid in solutions containing uranyl nitrate, aluminum nitrate and sodium dichromate.

Introduction: "When uranium slugs are irradiated in the Hanford piles, one of the many substances formed is U-237. It arises primarily from the reaction: U-238 (n,2n) U-237. This uranium isotope is a beta and gamma emitter with a half-life of 6.7 days. Because it is isotopic with U-238, it accompanies the uranium recovered by the Redox Process, and its concentration is, of course, not diminished by the fission-product decontamination cycles. As a consequence, the maximum total decontamination of irradiated uranium which can be achieved depends upon the quantity of U-237 present in the processed uranium."

Introduction: "During the early planning of the multicurie cells at Hanford it was decided that a slave type manipulator would be necessary to perform the experiments for which the cells were being designed. The design required steel roofs for installation through a port hole in the face. Technical Services Equipment Development began developing the proposed manipulator in January 1951 by making a study of those existing at that tune. None met the requirements which had been established; therefore, the design of a special manipulator was started immediately."

The processing of irradiated fuels from power and propulsion reactors is planned by Hanford Atomic Products Operation as part of the Atomic Energy Commission's interim reprocessing scheme. The several chemical processes proposed for the reprocessing of these fuels may be divided into two categories: first, total dissolution processes which dissolve both fuel element cladding and core material and, second, selective dissolution processes which dissolve either cladding or core material. The Niflex process uses a nitric and hydrofluoric acid solution for the total dissolution of stainless steel clad uranium fuel elements.

Report of data for the distributions of elements 90 through 96 between molten salt solutions of their chlorides and molten aluminum (or alloy) at 725 Celsius.

Transfer operations of highly radioactive solutions by the normal transfer pipette leads in most instances to a spread of contamination in the shielded facility. In the dripless pipette, the solution is held in the reservoir by the use of the siphon principle.

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.