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Procedures for the handling and recovery of nitrogen oxides and rare-gas fission products from the nitric acid dissolution of irradiated uranium are described. (C.J.G.)

Final approval of a flow-pressure characteristic curve and orifice pattern for C Pile is desired as soon as possible to allow the completion of orifice assemblies and parallit instrumentation. The presentation of a recommended flow-pressure curve and an orifice pattern has been delayed by a lack of experimental flow data on the C Pile flow system. Discussion with members of the C-431-B working committee and Pile Technology Unit personnel indicated that the use of flows higher than those recommended by the Reactor Unit, in HDC-2327, was desirable on the basis that the costs of supplying more water to C Pile than required by present outlet temperature limits will be small compared to the probable saving that can be made in reduction of corrosion and slug rupture frequency. Using this agreement as well as previous data and the assumption of low enrichment as a basis, a preliminary recommendation of a higher flow curve is presented.

Attention is now being drawn to the refractory properties of the pure dioxide of uranium, which make it suitable for special applications, such as the melting of pure uranium. Properties of uranium dioxide are described as they relate to fabrication. Forming of UO{sub 2} refractories by pressing and slip casting is discussed, as well as the firing of urania refractories and the properties of fired ware. Health hazards pertaining to UO{sub 2} are briefly described.

This test project is being carried on for the purpose of testing certain machines which are designed to charge and discharge reactor units with process metal while the unit is in operation. Satisfactory charging of process pieces while the reactor now required for charging operations. These presently required shutdowns are costly in terms of production time, and may cause undesirable thermal stresses in the unit as it changes power level. If charging can be accomplished during operation, reactor shutdowns would only be necessary for maintenance or emergency purposes. During the testing of these metal handling machines, problems have arisen involving the behavior of the process pieces as they are charged. Because these problems directly affect the charging operations, they are being closely investigated also. the machines covered by this report are designated as pressurized machines since they contain the same water pressure as the end of the process tube on ;which they are located. In most of the test work covered by this report, the full scale pressurized charging machine was replaced with a small charging cylinder. The charging operation was simplified by this change.

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The Reactor Safeguard Committee formula for the allowable minimum distance between a reactor and the public domain is expressed as: r = 0.01{radical}P, where r = radius from reactor in miles, and P = reactor power level in kilowatts. The six plates in this report illustrates the application of this formula for Hanford Works, both for individual reactors and all reactors in an area.

This document is a compilation of obsolete security codes, convenience codes and local jargon in use, or previously used, at Hanford Works. It is compiled to aid readers in understanding the Hanford report literature, particularly that originated in the early years of the Project.

Eleven thorium-carbon alloys in the range 0.02 to 2.0% carbon were prepared by vacuum-arc melting. The hardness varied from 19 R/sub A/ for the 0.02% alloy to 108 R/sub B/ for the 2.0% alloy. The preparation of thorium- chromium alloy was hampered by a reaction with the zirconia molds, which results in porosity in the casting. The failure of aluminum-silicon bonded X-10 slugs after heat treatment for seven days at 400 deg C was traced to the failure to remove tin after the tin dip and the presence of a gas during heat treatment. Tensile tests were made on extruded thorium, and no correlation was found with extrusion variables. A composition effect was found, however, which showed a linear relationship between carbon content and tensile strength. Creep tests on thorium, uranium, and Inconel continue. The program of evaluating the variables in the fabrication of clad fuel elements, using MTR techniques was concerned mainly with the distribution of UO/sub 2/ in the core compact, the temperature of rolling, and the particle size of the metallic core powder. Progress is being made in the preparation of three-layer laminates without the use of a protective can during the hot-rolling operation but with the …

This is a progress report of the production reactors on the Hanford Reservation for the month of April 1952. 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.

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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.

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 …

A high-density concrete shield is proposed for the biological shield. Adequate shielding of personnel and equipment can be obtained for about two-fifths the cost of a laminated shield of steel and masonite. The proposed shield is expected to behave satisfactorily during the life of the Reactor Unit and, in addition, the shield will possess considerable resistance to earthquake and blast loads. Details of design specification are given.

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.

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This report addresses the biological effects of ionized radiation

Estimated water flow rates at H Pile for various postulated conditions are given. The conditions under which the data are valid are as follows: (1) the estimated accuracy of the tube flow rates is {+-}5%, and of pile flow rates {+-}7%; (2) the pile flow rates are based upon the present tube geometry, orifice zone pattern and total flow at H Pile; (3) the flow rates should be increased about 2% if film-free slugs and tubes are considered; and (4) flow rates for enlarged fittings are based upon fittings similar in shape to the present H Pile fittings. The maximum tube flow rates for the DR Pile would be about 2% greater than those given in the table for present tube fittings. For enlarged fittings, the rates for DR Pile tube would be the same as those given for H Pile tubes. The ultimate permissible flow will depend, in part, upon limitations on maximum cross-header pressure and maximum tube inlet pressure, and upon the instrumentation necessary to detect abnormal tube flow conditions.

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This study soughs to determine the x-ray activity of two samples of stainless steel prepared from tubing used in the ANL-140, in-pile experiment (P-13). This information is necessary in order to predict the activity of the P-13 rig at the time of removal. The primary contribution to the over-all activity will be from Ta{sup 181} and Co{sup 59} impurities.