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.

A program has been prepared for the 709 computer which will determine extrapolation length from horizontal traverses measurements taken in an exponential pile. The flux across the pile is assured to fit a cosine function of the form y = A cos B (X-C)

Abstract: This report gives the results of a radiochemical analysis of particles collected by electrostatic precipitation from ventilation air of the "B" plant at Hanford Works.

One of the problems associated with evaluating the hazards resulting from the discharge of radioactive pile coolant water into the Columbia River results from the subsequent consumption of the river water by cities downstream of Hanford. Determination of the travel time of the river water from the reactor cooling water discharge points to the first downstream cities to consume the river water (Pasco and Kennevick) is a necessary part of the hazard evaluation. Knowledge of the travel time is required to establish decay correction factors for river samples collected in the Regional Monitoring Program and to determine the correct sampling time at downstream locations for samples that may represent specific Hanford operating conditions.

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.

This document describes required laboratory space and lists major equipment items necessary for a routine water quality laboratory in the P. R. T. R. Building. During discussions with R. D. Widrig and V. L. Rooney about the analytical sample program for the Plutonium Recycle Test Reactor, the author was asked to summarize equipment and space needs for a water control laboratory to provide routine analytical coverage on some of the water systems. Based upon 1706-KE-KER experience, some operating personnel may be used to provide analytical coverage on those routine analyses that are needed on around-the-clock basis with a savings of both time and money.

This paper is a preliminary report of experiments being carried out on the compaction of UO₂ powders by swaging within a restraining tube. Various types of UO₂ have been tried with resulting apparent densities as great as 9.89 grams per cubic centimeter.

Unanticipated radioactive contamination of the 200 Areas environs was discovered in the latter part of 1947, about 2.5 years after the startup of the Bismuth Phosphate Separation Plants. It was subsequently established that this contamination was due to radioactive particles emanating from the Separation Plants' stacks, and the need for exhaust ventilation air filtration equipment was recognized. Sand filters were installed in the fall of 1948 in the ventilation systems of the B and T Bismuth Phosphate Plants. These filter units essentially eliminated this contamination problem and have performed satisfactorily to the present time.

Observations on sheep representing flocks in Utah adjacent to the Nevada Proving Ground were compared with findings made on experimental sheep exposed to various amounts of radioiodine. The Utah sheep showed no evidence of the radiation damage observed in experimentally treated sheep. Estimations of amounts of radioiodine on vegetation required to cause serious thyroid damage following a contamination event are included.

A laboratory test was performed to determine if a 10% oxalic acid solution could be safely be employed as a substitute to 2% chromic acid plus 5% phosphoric acid for removal of corrosion products from aluminum corrosion samples or slugs.

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.

periodically the Industrial Hygiene Unit has the occasion to perform efficiency tests on respiratory protective devices and on filter cartridges used with these units. A study of the relative merits of the flame photometer and conductivity meter was made to determine which is more applicable for the determination of NaCl in filter samples.

Report describing the methods, discrepancies, accuracy of methods, and conclusions of two different analytical methods of spectrochemical analysis of plutionium.

This document will be concerned with the design, operation and maintenance of an automatic fuel element stamping machine. In order to take advantage of automatic unloading and length checking at the Acme-Gridley lathes in the 313 Building, automatic stamping of fuel elements is desirable. This, along with the need for a permanent canning line identification, resulted in a request for a machine to automatically stamp HAPO fuel elements on a production basis.

The solids in Purex lww were first observed during flowsheet tests for recovery of fission products from plant wastes. Since the nature of this solid was not apparent from the flowsheet composition of lww, some work was performed to characterize this material. Although this work has been conducted over a period of about one year, it has been subordinate to the main one of testing flowsheets for fission product recovery. The solids have been observed in each of about six samples of plant lww that have been studied, and the centrifuged volume of solids has been about four percent in each case.

The computation of liquid-boiling heat exchanger performance under a wide variety of operating conditions is a cumbersome, time-consuming procedure. Rapid evaluation of a design requires the use of some form of computational aid. Exchanger equations covering the case of liquid-boiling heat transfer have been derived and evaluated for a specific correlation of local boiling condition. Examination of various data as presented by Bonilla and McAdams indicates that the boiling coefficients are not fully reproducible and appear to be sensitive to surface conditions, gases, geometry, etc. It therefore is believed advantageous to extend the reference analysis to cover a wide range of surface and pool boiling conditions. In this manner, the effects of potential errors in both boiling and liquid film coefficients may be investigated with minimum effort.

It is anticipated that neptunium will be recovered in the Purex process by solvent extraction or ion exchange methods as a nitric acid solution of greater than 0.1 g. Np/1 and containing varying amounts of fission products, plutonium, uranium, and thorium, including Th234 (UX1). At the present time this solution is thermally concentrated in the Purex L-cell package to several grams of neptunium per liter. In this operation the solution is contaminated rather badly with plutonium and stainless steel corrosion products. The present specifications are for the neptunium final product to contain less than 0.1 weight percent plutonium, to be relatively free of gross metallic contaminates, and to be low enough in fission product game activity and Th234-Pa234 (UX1-UX2) beta activity to be handled without resorting to remote techniques.

This study was initiated to determine the minimum safety requirements for safe anchorage of electric pole lines. The objectives were to investigate the suitability of manufactured concrete pole line anchors for use at the Hanford Atomic Products Operation, to design a concrete anchor slug of optimum shape and size for manufacture, to determine by field tests the resistance pullout of the concrete anchor when buried at various depths under the most favorable condition, and to develop a safe design load for the anchors at various depths of burial and angle of pull based on test results.

Instrumentation shall be provided as shown on the attached drawing, SK-1-26835 for Confinement System Control. The Primary objective of the confinement system shall be to provide maximum possible confinement of contamination, consistent with building limitations, in the event of any contamination, consistent with building limitations, in the event of any credible reactor accident resulting from equipment failure or reactor mis operation.

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.

From introduction: "The test was performed to gain knowledge which would be useful for evaluating and improving currently planned methods for the design and construction of high density concrete shielding."

Experimental studies on continuous denitration performed in 321 Building resulted in the development of an agitated-through-type calciner. The performance of a paddle-agitated 4-in.-diameter by 44-in.-long reactor was sufficiently promising to warrant construction of a 16-in.-diameter by 8 ft.-long semi-plant scale model. This progress report summarizes the results of studies to date employing the semi-plant scale calciner.

To forestall the oxidation of graphite in HAPO reactors, a process specification provides that the oxygen concentration in pile gas should be kept to a minimum. Close control in maintaining this limit is difficult when only periodic samples are taken because laboratory results are always obtained long after any required corrections should have been made. A continuous-type analyzer that provides immediate knowledge of the oxygen concentration eliminates these time delays.

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.