From introduction: "This report presents the apparent viscosities of supernate-sludge slurries prepared synthetically to simulate the conditions known to exist in the waste storage tanks, but with fission products and plutonium absent."

From statement of objectives: The purposes of the work described in this report are to provide data useful for designing and operating the dissolver solution clarification equipment to be installed in the Redox production plant.

The following document describes data collected from a 14-month period of analyzing crustaceans containing radioactivity. Provided is information based on the change of radioactivity through seasonal changes.

From summary: "The relative absorptions of X-rays by various substances are listed as a guide to the application of the X-ray photometric method in chemical analysis and to the estimation of interference by extraneous substances."

This report follows an investigation of a cargo truck found contaminated by various radioactive material on a routine trip from Arco, Idaho to the Hanford plant on May 8, 1952. The report analyzes the truck's route in an attempt to find the location of the contaminating region.

Report evaluating different methods of preparing nitric acid solution and different methods of separating plutonium from this nitric acid solution of slag and crucible. Methods are explained and experimental data is given.

This report discusses the settling rates and properties of 101-U sludge when slurried with increasing amounts of synthetic supernatant.

Report discussing a study regarding iodine removal from dissolver solutions. Removal was traced by gamma activity in an iodine isotope which was added to the dissolver.

Report presenting a study for determining a way to remove iodine from dissolver solutions through sparging. DIssolver solutions were spiked with an iodine isotope that was used to track iodine removal through gamma activity.

Introduction:"A corrosion test program in support of the Redox Project was carried out during the second quarter of 1950 to determine the corrosion rates of several austenitic stainless steels in simulated 202-B Waste Storage solution containing HF, HN03, and H2SO4 acids. The results of this program are contained herein and all original data are recorded in laboratory notebook HW-3282-T."

Abstract: An investigation was conducted to determine the absorption and translocation of fission products by Russian thistle from a localized spot of contaminated soil. The amount and identity of the radioactive elements absorbed and translocated by the Russian thistle is given along with the location of these elements in the plants. Beta radioactivity to the amounts of 10 microcuries per gram caused no visible effects on the growth habits of sectioned material are included.

The following report examines a rapid method used to eliminate the possibility of explosive perchlorate-organic mixtures in a control laboratory in order to determine radiocesium. These methods include the usage of chloroplatinic acid as a precipitate of radocesium.

Report discussing the progress made on research concerning sluge-supernate acidification. This report focuses mainly on "'reverse strike' acidification and the observations of the physical characteristics of the test material," which was a synthetic incubated metal waste.

Report discussing analytical methods that were developed for various procedures in the Radiolanthanum Laboratory.

Report discussing a study regarding the identification and property analysis of plutonium (IV) monobutyl phosphate, which formed as a white precipitate in another study of the effects of monobutyl phosphate on plutonium(IV) in tributyl phosphate extractions.

Report reviewing a study regarding the chemical properties, appearance, texture, homogeneity and slurrying ability of waste metal sludge. The settling rates of sludge-supernatant slurries is also discussed.

Report discussing a study of various types of mercuric salts in order to determine which salt prevents the evolution of iodine which results from sparging.

Report discussing changes in parameters regarding Sutton's diffusion hypothesis, which is used in calculations of plant stacks' atmospheric pollutant distribution. An explanation of the computation and results are given.

The following report covers a test with the objective to demonstrate that the plutonium and uranium losses associated with the aluminum jacket dissolution could be reduced by substituting a water wash for the 5 per cent nitric acid wash following the coating removal and that this change would not adversely effect the product yield of decontamination in subsequent process steps.

Introduction: "Evidence of masonite deterioration in test hole plugs was interpreted to mean that a previous estimate of the minimum lifetime of the masonite in a Hanford biological shield might be somewhat optimistic. The importance of understanding the effects of radiations on masonite for both present and future pile shields dictated that masonite be sampled from an operating pile shield and studied. The results and techniques are discussed in the following sections of this report."

It has been suggested by many workers in the power reactor field that a more efficient utilization of natural uranium may be obtained in thermal reactors if the Pu produced in a fuel cycle were used to enrich a subsequent cycle in which the irradiation-depleted-uranium would be reirradiated. The work described here was done for the purpose of evaluating the reactivity value of the Pu enrichment under the assumption that all plutonium produced is, after chemical separation, fabricated into separate fuel elements and not alloyed with recycled uranium. It has been suggested that the reactivity value of the Pu decreases with exposure to such an extent that highly burned plutonium should be discarded. We, therefore, wish to look at two of the variables affecting the limiting exposure: the reactivity value of the fuel as a function of exposure and temperature and the fraction of potential fissions which would be discarded as a function of exposure. Though the residence time of the Pu fuel before reprocessing, refabrication, and recycling depends upon the relative decrease in specific power, decrease in reactivity due to fission product build-up, the lifetime of the fuel elements before failure, and the cost of reprocessing, only the first …

The Hanford Test Pile is a heterogeneous, low power, graphite moderated natural uranium pile. The pile consists of an eighteen foot cube of graphite containing 292 charged channels in a square array with an 8-1/2-inch lattice spacing. This report describes the reactor and the operating procedures used, and presents the latest results of calibrations performed. These calibrations supersede other work which has been done on the Test Pile and contain refinements of most of the early calibrations.

Past experience has shown that the demand for increase separations plant capacity comes up very regularly. One of the variables which greatly affects plant capacity is cross-sectional area of the individual vessels. Larger areas permit greater flow rates as well as more space for the installation of heat transfer piping (shell and tube concentrators). Design considerations of the separations plants vessels have been based on both circular cylinder and slab geometries. A study has been made to determine other vessel geometries that will result in safe vessels from a nuclear safety standpoint and at the same time offer larger cross-sectional areas than right circular cylinders. Vessels of elliptic as well as annular cross sections have been considered. It is neither the intent of this study to discuss the effects of intersection, vessel piping, etc., nor the pros and cons of fabricating feasibility and structural strength of these different shaped vessels. The main purpose is to make comparisons of cross-sectional areas (capacity parameter) of safe vessels so that vessel shape may be evaluated as one of the parameters in any design study for separation plants.

Approximately 15,000 "P" Process Aluminum caps clad with aluminum-silicon on the face and sides have been successfully canned. These caps were fabricated by the Aluminum Company of America from clad plate stock. An increase in canned slug yield of about two percent was realized from using these caps and no significant process problems indicated that the additional cost of clad plate might be off-set by the elimination of the centerless grinding now required on the side of the cap base. Therefore, processing of at least 100,000 additional aluminum-silicon clad caps is recommended, and consideration should be given at this time to completely converting from the present cap design to the aluminum-silicon clad cap.