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

This report discusses methods of producing aluminum nitrate for the Redox process. Large amounts of aluminum nitrate are consumed during the Redox process, so this report explores preparation of this material through the dissolution of hydrated aluminum oxide in nitric acid.

Report discussing experiments and developments in purification of graphite starting from March, 1949.

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.

This report follows studies of the accumulation of beta omitting radioactive elements in Columbia River fish.

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."

"Recent surveys by the Health Instrument Section have disclosed the presence of many small radioactive spots on ground surfaces in the T and B plant areas. Investigation has shown that representative samples of the spots when mechanically separated invariably end in a single radioactive particle."

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

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: "The carrier (pyroelectric) concentration method is commonly used in the spectrographic analysis of plutonium metal. This report describes work which lead to substantial improvements in accuracy, precision and ease with which the analysis is made. Included are data from analyses of plutonium metal by the original and the improved carrier methods, and the cupferron extraction method."

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.

"An apparatus was designed and fabricated to obtain film samples from irradiated process slugs. This apparatus has been used to obtain film samples from slugs of the sodium dichromate elimination test."

This report discusses preliminary results acquired from investigations made on the ultrasonic properties of uranium, such as sonic vibrations in the megacycle frequency range.

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."

This report discusses the the surface characteristics of virgin and irradiated artificial graphites and the changes which occur upon irradiation.

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.

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.

Report presenting results of a study regarding the critical masses and nuclear properties of plutonium as well as the use and effects of various tampers.

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.

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.