Acenaphtheno is a low-melting organic compound, C{sub 10}H{sub 6}(CH{sub 2}){sub 2}, used as a temperature indicating coating for slugs tested in the Frost Test induction coil. In preparation for this test, slugs are coated by spraying a 12 to 24 per cent by weight solution of acenaphthene in a high vapor pressure solvent on the rotating slug. The solvent evaporate leaving a thin, adherent coating of acenaphthene. This process is completely described in reference 1. To be entirely satisfactory for this use, a solvent must satisfy four requirements. High vapor pressure -- rapid evaporation. Non-flammability. Low toxicity. Carbon tetrachloride, the solvent used at percent, is operationally satisfactory but undersirable because of its toxicity. The purpose of this test is to investigate the suitability of various solvents as substituted for carbon tetrachloride.

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It was agreed at the February, 1951 meeting of the Atomic Energy Commission Technical Information Panel that, wherever practical, progress reports should periodically include a list of previous reports issued in the same series. Such a list would greatly facilitate servicing by installation document rooms since requests are frequently received for reports in a specific series. This report lists the Technical Progress Letters from 1944 to date, and is distributed to recipients of the Letters in accordance with the recommendations of the Panel. Because of its length it was not deemed practical to include it as a part of the report itself. lt is anticipated that it will serve as a reference list and permit reduction of personal files.

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This is a memorandum from the Hanford Operations Office about the possibility of tritium wastes escaping into the atmosphere. The waste which is disposed of in burial grounds eventually migrates to the Columbia River where it evaporates into the atmosphere. Also included in this memo is a statement of an error in the units of some data compiled for the monitoring of the tritium wastes. (MB)

This is a progress report of the production reactors on the Hanford Reservation for the month of January 1951. This report takes each division (e.g. manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

This is a progress report of the production on the Hanford Reservation for the month of June 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

This memorandum discusses investigations concerning the coupling step required to adapt a Redox IIBP to a 234-5 feed which have indicated that a single peroxide strike from a solution 50--60 g./1. in product and about 2 M in nitric acid is adequate. Additional work is planned to demonstrate both peroxide and Pu(IV) oxalate strikes under these conditions. The present Redox PR cage product contains 10 g./1. in product and over 500 g./1. in nitric acid, and probably requires tow peroxide strikes; in addition, the operating capacity of the PR cage may be inadequate to process the throughputs desired in the Redox plant. A two stage continuous concentration process for Redox IIBP is jointly suggested by Design and Development and Chemical Development personnel, and is described in this document. This process is believed to be simple and easily operable. Acid is removed from the product solution adequately for the single peroxide strike coupling procedure. Batch size appears to be easily controllable.

This report discusses particle size measurements with a modified cascade impactor of the particulate matter in the dissolver off-gases during the period of maximum gas evolution, and during the air sparge of the dissolver solution after the completion of the dissolving cycle which indicate that the mass-size median of the particles was less than 0.1 micron in both cases. The measurements were made, unfortunately, at a time when the efficiency of the silver reactor upstream to the sampling point was low enough to permit appreciable amounts of iodine through to the cascade impactor. The samples were therefore permitted to decay for thirty-five days before the relative amounts of radioactivity on the stages of the impactor and the follow- up filter paper were evaluated. The size of the particles obtained, 0.1 micron, should be considered primarily as an order of magnitude rather than an absolute value, because the cascade impactor (as well as any other impaction device) can not give an accurate value for the particle size in this size range.

This is a progress report of the production reactors on the Hanford Reservation for the month of November 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

This is a progress report of the production reactors on the Hanford Reservation for the month of September 1951. This report takes each division (e.g., manufacturing, medical, accounting, occupational safety, security, reactor operations, etc.) of the site and summarizes its accomplishments and employee relations for that month.

The study of radiation damage to the graphite moderator is largely dependent upon the availability of graphite samples from the moderator itself. Because of the varied temperature-flux history and complicated structure of the piles, it is impossible to study the problem from the laboratory standpoint alone. The samples of greatest interest are those taken from the sections of the moderator which have received the greatest damage. However, a comprehensive study requires samples from all sections of the graphite moderator. It is desirable that these samples be solid in order that physical and dimensional measurements of all types may be made. Solid samples enable thermal conductivity measurements and the gradients of damage across tube blocks to be measured. To provide such samples, it was necessary to design some device which would cut solid samples from any location in the pile through the process Channels. The graphite core borer was desigred specifically for this purpose. The objective of this Production Test was to test the operation of the core barer and evaluate the results of its performance.

It was the purpose of this talk to present up-to-date information with regards to dimensional stabilityof metallic uranium. A brief discusssion concerning the history, present status, proposed theories, and recent results are given. The factors affecting and methods of minimizing dimension changes are outlined.

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This volume is one set of a series which has been prepared as a record of the research work done under the Manhattan Project and the Atomic Energy Commission.

The fused fluoride-Inconel reactor system was studied under irradiation with nominal 19 Mev deuterons as supplied by the Berkeley 60-inch cyclotron. Chemical, metallographic, magnetic susceptibility, electron diffraction and X-ray diffraction studies were made on the as-received materials, one control run and two irradiated runs. No changes in the fused fluoride fuel were noted. Accelerated intergranular corrosion and increased grain size were observed in the irradiated Inconel specimens. This report is based upon studies conducted for the Atomic Energy Commission under Contract AT-40-1-GEN-1064.

A. M. Gaudin and collaborators have published a curve of refining costs versus ore grade in terms of dollars per pound of U308 recovered on the basis of 90% recovery.

The purpose of the research work reported in this volume is the development of a process for the separation and subsequent purification of radium from the K-65 (pitchblende) residue. Except for the accumulation of additional experimental data the process is essentially complete. After a preliminary extraction of about 85% of the lead and 40% of the silica with a 40% sodium hydroxide solution, the residue is treated with a solution containing both sodium hydroxide and sodium carbonate to obtain additional lead removal and partial conversion of the radium and barium sulfates to carbonates. After leaching out the soluble carbonates, the residue containing unconverted sulfates is treated with a sodium carbonate solution at 170{sup o} under pressure. A study of the amount and concentration of sodium carbonate required to obtain a satisfactory conversion of the sulfates has shown that the total sodium carbonate may be reduced by 40% and that smaller volumes of solution may be employed without serious decrease in the efficiency of the conversion. In order to make calculations on the number of steps and tank sizes for the radium-barium separation by fractional precipitation, equations have been developed to calculate this information from the theoretical separation factors determined experimentally. …

A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.

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Siimplified shielding calculations indicating the proposed design for the water boiler assembly will reduce the radiation at normal operaton to values well below those which are considered tolerable.

Two important problems in the investigation of reactor hazards are the chemical reactivity of various materials employed in reactor construction and the chracteristics of heat transfer under transient conditions, specifically heat transfer when driven by an exponentially increasing heat source (exp t/T). Although these problems are independent of each other, when studied in relation to reactor hazards they may occur in a closely coupled sequence. For example the onset of a dangerous chemical reactor may be due to structural failure of various reactor components under an exponentially rising heat source originating with a runaway nuclear reactor. For this reason, these two problems should eventually be studied together after an exploratory experimental survey has been made in which they are considered separately.