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The effect of fuel programming, i.e., the scheme used for changing fuel in a core, on the reactivity and specific power of a sodium graphite reactor is discussed Fuel programs considered Include replacing fuel a core-load at a time or a radial zone at a time, replacing fuel to manutain the same average exposure of fuel elements throughout the core, and replacing and transferring fuel elements to maintain more highly exposed fuel in the center or at the periphery of the core. Flux and criticality calculations show the degree of power flattening and the concurrent decrease in effective multiplication which results from maintaining more exposed fuel toward the core center. Corverse effects are shown for the case of maintaining more exposed fuel near the core periphery. The excess reactivity which must be controlled in the various programs is considered. Illustrative schedules for implementing each of these programs in an SGR are presented. (auth)

The general prospects of several radioisotopes are reviewed; the special properties of U/sup 232/ and Th/sup 228/ are poi nted out; and ionium (Th/sup 230/ ) and protactinium target materials are discussed from the sthndpoint of availability and chemical separations processes required for the preparation of U/ sup 232/ and Th/sup 228/. Outlines are given for potential schem es for the separation of U/sup 232/ and Th/sup 228/ from uranium milling pr ocess waste streams and from the irradiation products of Th/sup 230/--Th/sup 232/ mixtures. The high heat generating rates of these potent alpha emitters make them especially suitable for primary consideration as heat sources for small thermoelectric generators. The exceptionally high alpha activity suggests their use in special neutron sources as Ra-Be sources, and they may have sufficiently high neutron generating rates to be in contention with some of the smaller research reactors and experimental neutron producers. (B.O.G.)

A set of four strong focusing magnetic quadrupole lenses has been constructed and operated. Each lens consists of four air cooled electromagnets with pole tips having a hyperbolic cross section. Each lens is 4 in. long and has an aperture 2 in. in diameter. Measurements of the magnetic field demonstrate that the hyperbolic cross section satisfies the requirements of a constant magnetic field gradient very well. The technique of deflecting a current carrying flexible wire has been used to measure the trajectory of charged particles through the system of lenses. It has been observed that the strong focusing requirements are satisfied. The system of lenses was then used to focus 0.5 Mev protons, 20 Mev deuterons, and 40 Mev alpha particles. The parallel beam of 0.5 Mev protons was detected by observing the incandescence of a quartz plate while the protons were bombarding it. The focused beam was less than 1 mm in diameter. The astigmatic 20 Mev deuteron beam from the 60 in. cyclotron was increased in current density by a factor greater than 30.

The distribution of the heated effluents discharged by Hanford reactors to the Columbia River has been a matter of interest since the early design stage of the first reactors. The pattern of this distribution is a major factor in determining the extent to which a downstream reactor is affected by those upstream, as well as the localized effects on the ecology of the river. Pollutional characteristics of the effluents are three - heat load (or temperature increase), chemical contents and radioactivity. The latter has received the greatest attention in connection with potential personnel exposure and effects on river biota; it has been assumed however, and generally confirmed by sampling that the measure of distribution of any one of these characteristics in the saw an for the others. Observed distributions of radioactivity for various river and reactor flow rates are documented. Unfortunately, any extrapolation of those observed distributions to altered flow conditions of river regimes is of questionable validity. Mathematical models of the problem have been formulated but have been of little value due to the necessity of measuring certain parameters under the conditions for which a solution is desired. Even so, calculated distributions provide only general patterns and would not …

The intent of this manual is to present a description of the main process building, equipment, and auxiliary facilities as well as a process summary. Material is of a scope nature with more detail devoted to features unlike those of existing separations systems. An attempt is made to convey some of the basic design philosophy and the problems encountered in the development of design criteria. This information is written primarily for Separations Section supervision who have not had an opportunity to become conversant with the Purex Project. The manual may also be of assistance in orientation and training of personnel. In order to avoid repetition and duplication of effort, one line service diagrams, equipment sketches, tables, and detailed data are not a part of this manual.

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Stainless steel-base fuel components of thin plate-typs construction and containing a dispersion of enriched UO/sub 2/ have been successfully employed in powering the Army package Power Reactor. The stationary fuel compcnent proposed for operation in the second core loading of the reactor is discussed. The component is designed for radioactive service in pressurized water at 4504DEF and consists of eighteen composite fuel plates joined into an Integral unit or assembly by brazing. Design specifications covering the material and dimensional requirements as well as the operating conditions are discussed. Step-by-step procedures developed and utilized in manufacturing the component are presented in detail. (auth)

Two designs of sodium cold traps for the HNPF have been subjected to full scale tests, Performance features that were investigated include oxide removal efficiency, oxide capacity, pressure drop characteristics, economizer effectiveness, and temperature profiles, Results indicate that both designs should perform satisfactorily in the Hallam plant, (auth)

The KPL No. 6, first-cycle Purex lA flow sheet is presented as an examaple of waste volume reduction through optimum solvent-extraction flow sheet design. Calculated on the basis of previous flow sheets, a 54 to 69% reduction of first-cycle waste storage volume is achieved by taming advantage of the extractability and reflux of nitric acid in TBP systems to provide adequate salting while minimizing the amount of acid going to waste. Although this report is concerned only with the first Purex cycle, the calculations are equally valid for the second U cycle, and the principles presented are applicable to similar nitnic acid-salted TBP systems. (auth)

A basic study of the mechanism of deposition of small particles on cylindrical collectors in the presence of electrostatic forces has been made. Theoretical equations predicting the efficiency of collection have been derived and solved with a digital computer. The solution of these equations is shown. The deposition of liquid aerosol particles on cylindrical collectors was measured experimentally for a range of electrostatic conditions. The experimental results agree with the theory. An experimental study was made of the effect of electrical charges on aerosol particles on the filtration efficiency of glass fiber mats. Experiments were also made on the effect of placing a potential on a mat of fine wire and on filtration by a mat of a tangled wire dipole. (auth)

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Uranium can be rolled from cast metal or forged ingot to sheet satisfactory for cupping, deep drawing, and similar fabrication procedures by a combination of hot breakdown in the neighborhood of 600 deg C and warm finishing at 225 to 325 deg C. Sheet may also be obtained by hot rolling alone and by warm rolling alone, but the combination of hot and warm rolling afforded the best and most practical method to secure good quality sheet in the quantity required. The percent reduction by hot working does not appear to be critical, but at least 60% warm reduction is desirable to obtain complete and controlled grain size by recrystallization with high ductility and strength properties. Except for research investigation, rolling of uranium below 225 deg C is not recommended. Hot rolling of unplated uranium from the as-cast or as-forged surface is recommended, using a bath of 35% Li/sub 2/CO/sub 3/ plus 65% K/sub 2/CO/sub 3/ for a heating medium. No further preparation other than washing the salt from the hot rolled surface is required before warm rolling, and a bath of Meltemp No. 7 oil is recommended for warm rolling. Starting with an as-cast tensile strength averaging 60,000 psi, …

Prior to the installation of improved charging machines and charging techniques serious gouging of the soft aluminum fuel element jackets was not infrequent. A certain amount of mechanical abrasion also occurs as the element is pushed over the tube ribs in the course of the loading operation. To eliminate or at least minimize mechanical damage there has been some experimentation with hard anodic coatings applied to the aluminum fuel element jackets. This report is an evaluation of thinner anodic coatings which should develop lower mechanical stresses within the oxide and should be less subject to cracking.

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Tensile properties amd impact strengths were determined for iodide Zr, sponge Zr, and alloys based on both grades of metal, containing nominally 1.5% M. Sheet specimens, welded and unwelded, were tested. Tensile properties were established at room temperature and 300 deg C. Impact strength values were measured at --100 deg , R.T., 100 deg , 200 deg , and 300 deg C. These (alpha) materials generally exhibited a lack of heat treatability, litile or no deterioration of properties due to welding, and essentially no indication of impact transition temperature. Tensile strength and impact behavior were established for alloys based on iodide zirconium containing 15% Nb, 15% Nb + 2% Pd, 15% Nb + 2% Pt, 15% Nb + 1% Fe, and a binary sponge zirconium + 15% Hb alloy. High strength levels could be established by proper heat treatment. The presence of Pd, Ft, or Fe seemed to delay the formation of the embrittiing agent as determined by hardness and resistivity vs time of anneal curves for these alloys. uth)

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Temperature distributions, irradiation effects, stacking arrangements, voidage, and economics for the borated-graphite shield of the PRDC reactor were investigated. Of the shield systems considered, four are reported here. System 1 contalns 30 in. of 1% borated graphite, with either ordinary graphite or a cement as a filler for the remaindcr of the volume. The maximum temperature at the flex plates in this system was calculated to be 5OO deg F. Systems 2 and 3 consist of 2 in. of 5% borated graphite near the core vessel and 1/2 in. of Boral at the primary-shield tank. A filler material of carbon blocks is used in System 2 and graphite in System 3. The calculated maximum temperatures were 700 deg F and 35O deg F, respectively. System 4 consists of a laminated structure of Boral and graphite near the primary-shield tank and carbon-block filler. It was calculated to have a maximum temperature of 600 deg F at the flex plates. The maximum temperature at the flex plates recommended by APDA is 500 deg F. Energy storage and radiation damage were found to be within permissible limits in all four systems. However, these conclusions are based on experimental data from the Hanford reactor …