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."

Report presenting a spectrochemical method for analyzing the concentration of iron and silicon in uranium, which requires approximately 1/3 of the time required for the previously used "wet 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.

The heat generated by reactor fuel elements create thermal gradiants which cause thermal stresses of considerable magnitude. An elastic solution for these thermal stresses can be calculated with the equations given in Timoshenko's and Goodier's book (1) on elasticity if the fuel element is assumed infinite in length. An exact elastic solution for a finite cylindrical fuel element is given in this report in order to evaluate the effects of fuel element length on the thermal stresses and the effects of having the free end material distorting the end closure.

The Redox solvent extraction process for the separation of uranium and plutonium from their fission products and from each other has operated on an over-all basis very successfully for more than four and one-half years. During this time, there have been occasional periods of high product losses and/or poor separation of fission products from uranium or plutonium as well as periods of poor operability due, among other things, to emulsification and entrainment. The search for the cause of these various process troubles has resulted in almost continuous testing of the process solvent in attempting to determine its contribution, if any, to these difficulties. This report documents the history of the Redox Plant solvent.

A two and one-half year period of extreme caution where titanium and nitric acid are involved has passed since the occurrence of the last and most serious uncontrolled incident involving titanium and nitric acid. During this period controlled explosions involving titanium and nitric acid have been routinely reproduced, additional laboratory and pilot plant experience has been gained and successful industrial applications have been appraised. Based on laboratory data and the known industrial experiences, it is concluded that titanium can be safely employed in systems handling nitric acid that contains more than two (2) percent water. It is recommended that the titanium heat exchanger recently fabricated by the Technical Shops be installed in the Purex Plant plutonium concentrator when replacement of the stainless steel unit now in service becomes necessary.

The object of this study was to determine the effects of several annealing treatments and aging treatments on the tensile test behavior of Zircaloy-3b.

Hydrogen peroxide has been used to reduce plutonium (VI) in nitric acid solution to the (IV) valence state. However, reduction of plutonium (VI) solutions containing iron has often been incomplete when hydrogen peroxide was used as the reducing agent. Since reduction of plutonium (VI) has been applied in plant solutions containing iron, a study of the plutonium (VI) reduction with hydrogen peroxide was undertaken. Variables tested included plutonium, nitric acid, and iron concentrations.