A healthy 39-year-old male weighing 65 kg was exposed for a period of 8 months to varying levels of HTO. The average tritium activity in body fluids over the entire period was 23 {mu}c/liter. A few weeks after exposure, when the HTO activity in body fluids had declined to about 0.2 {mu}c/liter, a biopsy was pe formed on skin and fat taken from the region of the lower abdomen, and the material was analyzed for tritium activity. The skin showed an average activity equivalent to 0.4 {mu}c/kg of dry tissue and the fat about 0.3 {mu}c/kg of dry tissue. The radiation dose per unit time from these activities was only 1 to 2 percent of the radiation dose per unit time during the 8-month exposure period. It was concluded that the radiation hazard due to retention of tritium in the organic components of these tissues of man after chronic exposure was negligible compared to the radiation hazard from HTO activity in the body fluids which was necessary to induce the activity into the organic components. Comparable experiments on mice previously reported indicate that this conclusion may hold for all tissues in the body. The water content of the skin and …

The fifteenth quarterly report being submitted for Contract No. AT04-1-GEN-12 is issued in accordance with Service Request Number 1 except for the report of the Alamogordo Section, Code 91810, which is submitted in accordance with the provisions of Service Request Number 2. Work is in progress on continuing existing projects. In addition, new projects have been initiated including the Kinetics and Mechanism of Protein Denaturation (10018); The Effect of Irradiation on the Constituents of Embryonic Serum (30033); and The Use of Controlled Atmospheres for Spectrographic Excitation Sources (40053). Many of the Project units are either wholly or partially completed and the following initial reports are available: Identification of Ferritin in Blood of Dogs Subjected to Radiation from an Atomic Detonation (UCLA-180); The Nutritional Value of Intravenous Tapioca Dextrin in Normal and Irradiated Rabbits (UCLA-181); The-Decarboxylation and Reconstitution of Linoleic Acid (UCLA-183); Preparation and Properties of Thymus Nucleic Acid (UCLA-184); The Radiation Chemistry of Cysteine Solutions Part II. (a) The Action of Sulfite on the Irradiated Solutions; (b) The Effect on Cystine (UCLA-185); A Revolving Specimen Stage for the Electron Microscope (UCLA-178); An Automatic Geiger-Mueller Tube Tester (UCLA-186); The Value of Gamma Radiation Dosimetry in Atomic Warfare Including a Discussion of …

Physical data are presented for transverse CSF samples with capsule exposures of 568, 1049, and 1617 MD/CT. The higher exposures indicate a sharper damage gradient toward the front of the pile. Additional casings of various types of graphite were loaded into test holes during this month. Average values of the thermal conductivity and electrical resistivity for several types of virgin graphites are presented. Data of this nature will be a regular portion of this report henceforth. Process tube channel 2677-H was mined and traversed for bore diameter. Although several of the tube block junctions were obscured, the channel was quite uniform. Examination of all previously mined graphite powder samples for aluminum oxide corrosion product has been completed and the results are reported.

This report discusses seven items related to the design of the 100-K water plant facilities. These items include: the number of process water connections to the reactor building, valving at the control building, and number of main pumping units; effect on heating system of the Hanford power outage during the winter period; reactor supply conditions; power requirements and steam flow at the local power plant; the service water system; general arrangement of the heat exchanger building; and pumping station control.

This study was undertaken at the request of the Hanford Works to investigate the possibility of detecting 3/8 inch diameter boron-steel control-balls which become lodged within cracks between the graphite blocks of an atomic pile. The cracks concerned occur radially from 4 3/16 inch diameter holes which pass vertically through the pile. The problem is complicated by the following facts: The graphite blocks are conducting and will therefore give rise to spurious signals primarily due to the cracks between blocks. Numerous aluminum tubes containing water and bars of uranium pass horizontally through the pile at distances closer to the hole than the ball at its extreme position. The vertical holes themselves are warped in an arbitrary manner. Calculations were made to determine theoretically whether or not the ball could be detected. Best operating frequency and coil design were also determined. Tests were made utilizing a specially designed search coil and a test section of graphite pile. Measurements of particle voltage vs. position relative to the coil were made and compared with that resulting from the graphite.

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The Redox Plant operated at a 76.7% time efficiency (IAF basis) and averaged 3.16 tons of uranium per operating day during the month. During the period from July 4 through July 13, 38,000 gallons of 72% ANN from Tygon-lined storage tank SS-112 were used in aqueous salt solution make-ups with some emulsion difficulty and fission-product carry-over resulting. New ANN, from stainless-steel tanks, was used for the balance of the month. An extensive (thermally) hot 60% HNO{sub 3} cleanout was given to the IA, IB, 2D, 3D, 2A, and 3A columns and associated salt solution head tanks to prepare the solvent-extraction battery for operation using 90-day `cooled` uranium. Continuous cross-over oxidation to improve 2A Column waste losses, and employment of KOH (vice NaOH) in 2DS, ICU, and 2EU butt additions to minimize Na in 3EU, were begun concurrently on July 20. Aged (90-day) uranium in head-end treated feed batches, was started to the IA Column on July 21, with early data indicating adequate two-cycle decontamination for both uranium and plutonium.

This progress report from the Argonne National Laboratory covers the work in Biological and Medical Research, Radiological Physics, and Health services for the quarterly period ending March 31, 1952. Numerous experiments were conducted in an attempt to answer some of the questions arising from exposure to ionizing radiation, especially X radiation. Some of the research involved the radiosensitivity of cells and some involved animals. The effects of radium in humans was also evaluated. Other studies were performed in biology, such as the effect of photoperiodism on plant growth and the biological of beryllium.

Since 1945 the routine production of plutonium (``A`` buttons) at Atomic Energy plants has been accomplished by heating mixtures of plutonium tetrafluoride, calcium and iodine in crucible-bomb assemblies. Yields above 97% and metal of adequate purity are consistently obtained by this process. Plutonium (``X`` button) production has also been carried out routinely by including plutonium turnings with the powder mixture given above. A majority (several thousand) of buttons produced in the 234-5 Building have been ``X`` buttons made by recycling up to 1.1 units of turnings/unit of plutonium in the plutonium tetrafluoride. Operating hazards exist in either process, however, and these are reviewed below.

This document provides the process specifications for the Triple Dip Process for Alpha Fabricated Uranium for production canning of uranium cores.

After almost four months elapsed time since product introduction into the RMA Line, it was pertinent to evaluate both process and equipment performance and arrive at a program for future improvement. The suggestions and recommendations provided are those of the Technical Section. In those passages made to acknowledge the views of the appropriate groups.

The object of the work described in this report was to adapt the laboratory procedures for precipitating and hydrofluorinating plutonium (IV) oxalate to plant equipment on 160 and 350 gram scale. This adaptation was done by making minor process changes in concentration, digestion time, rate of chemical addition, etc.

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Redox effluent for production was first processed in the 231 Building on February 4, 1952. Samples of process streams were taken after each purification step to evaluate the adequacy of purification processes in,the 231 and 234-5 Buildings and to provide close control of this material. Data regarding pile residence histories, radiation surveys of ``Sample Cans``, reduction yields, C/Q values, and product purity after purification steps were accumulated and are presented in this report to establish preliminary data which may be used to evaluate future process changes.

Several operations that are important to the process as carried out in Task III (Reduction) are performed by the mixer. In order to specify the process certain tests were made to study these operations and are discussed in this report. They include: Mixing Time, Mixer Discharge Rate, Mixer Holdup, and Mixer Capacity. A description of the test, conclusions and recommendations is provided.

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This document details activities of the Chemical Development group of the Separations Technology Division during the month of December 1952.

The present RMA Line has now progressed through the run-in period and is engaged in actual production. Considerable new data and experiences are available for use in an evaluation of the basis used for design of the equipment in regard to contamination control and the safety of operating and maintenance personnel. Research and development funds have been allocated to the improvement of the 234-5 process equipment. Thus it is desirable that an up-to-date basis be established for use as criteria for the design development of these improvements. It is concluded that remote control operation of all plutonium purification and fabrication equipment is ultimately desirable. However, until reliable, tested, remote-control equipment is available, the remote control requirement may have to be compromised on some steps of the process to allow non-remote or rubber glove operation in order to obtain equipment with sufficient operating reliability.

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This report presents information pertaining to the life of the reduction bombs in both the RG and RMM Lines of the Plutonium Purification and Fabrication operations in the 234-5 Building. The expected average life to rupture of the RG Line Bomb of carbon steel subjected to an internal pressure of 450 psig is ten hours at 750 C. The expected average life of the RM Line bomb of V-36 alloy is more than 100,000 hours at 750 C under the same operating conditions. These expected average life values do not include any calculated factor of safety. Since they are average figures, some bombs could be expected to have shorter life to rupture. The BG Line bomb of carbon steel is definitely hazardous under the extreme service conditions of 450 psig and 750 C. If the use of carbon steel is to be continued it is recommended that the bombs be redesigned to reduce unit stresses.

The General Engineering Laboratory has submitted a cost estimate and a proposal covering the design, fabrication, construction, and testing of a uranium slug canning machine. This machine will cover requirements for mechanization of only the operations occurring within the aluminum-silicon canning bath. These operations are: Can and cap preheating, Can and cap wetting, Canning assembly, and Quenching. Stated briefly, the machine is required to insert a prepared uranium slug into an aluminum can and close the can opening with an aluminum cap. All assembly operations will be carried out beneath a molten bath of standard Hanford aluminum-silicon bonding alloy. The uranium slug is preheated and prewetted before being manually transferred to the slug assembly and quench machine. The process performed by the machine will be integrated with the manual slug preheating and wetting equipment so that proper preheating and wetting is attained before the slug is inserted into the machine. After assembly the completed canned slug will be transferred to the water quenching station, where the components are held firmly together until the molten aluminum-silicon has frozen, forming a homogeneously bonded assembly. This report provides a detailed description of the machine.

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An estimated minimum critical mass for each of the process vessels in the 231 Building has been calculated on the basis of critical mass data given in the P-11 Project Document HW-24514. The calculations are made assuming the plutonium to be a homogeneous mixture of precipitate and water with some slight neutron poisoning due to other elements. The precipitate is further assumed to have partially settled making an effectively infinite water reflector above the plutonium and hence reducing the critical mass.