Extraction systems involving chelating agents are widely used in radiochemical separations on a control basis due to the economy in time and equipment. In the case of waste streams, the element of interest is, by definition, at a very low concentration in the presence of gross fission product activities. The resultant large sample size necessary for a reliable analytical results creates a marked radiation hazard to laboratory personnel.

Disposal on a specific retention basic of certain radioactive liquid waste solutions emanating from separations from separations plants has been practiced at Hanford since 1944. As used at Hanford, the term "specific retention" is defined as that volume of waste liquids that may be disposed to the soil* and be held against the force of gravity by the molecular attraction between sand grains and the surface tension of the water, when expressed as percent of packed soil volume. In practice it represents the volume of liquid that may be discharged to a disposal pit of known dimensions without leakage to the ground, water, expressed as percent of the total volume of a column of soil with the same cross section as the pit, and extending from the bottom of the pit to the water table. It is recognized that some degree of lateral spreading will occur which has the effect of enlarging the volume of soil contacted by the liquid.

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.

In recent months, relatively large quantities of neptunium-237 have been recovered from Hanford irradiated uranium. The over-all recovery process consisted of 1) isolation from Purex plant solutions by solvent extraction, to yield an impure neptunium nitrate solution contaminated principally with uranium, plutonium, and fission products (2) purification by anion exchange 3) precipitation of neptunium (IV) oxalate; and 4) calcination of the oxalate to yield neptunium (IV) oxide, NpO2. This paper describes the oxalate precipitation process investigations and results.

Since the calculation of the PRTR energy distribution was first carried out by J. R. Triplett, the design has become sufficiently fixed to allow a refinement of his values. The present analysis, also, includes a calculation of the fraction of energy which is released in the shroud and process tubers that flows to the primary coolant to the top and bottom shield coolant is taken into consideration. Nuclear data used in the original calculations still appears satisfactory and is, therefore, utilized in the present analysis.

The region in which Hanford's chemical processing plant areas (separation areas) are located in the central portion of the Hanford Works area on a plateau, shaped like a giant bar, of glaciofluviatile materials. The general locations of these areas is shown in Figure 1. The region studied rises from 200 to 300 feet above the surrounding area and encompasses approximately 40 square miles or somewhat less than 8% of the Hanford reservation (Figure 2). Within the region are located the chemical separation plants and most of the major waste disposal sites.

Information was desired on applicability of seismic methods of geophysical exploration in helping to better define and refine hydrological and geological factors affecting ground disposal of liquid radioactive wastes. To date most of information on the area geology has been obtained from surface mapping and well drilling, although it has long been recognized that geophysical techniques may greatly assist in defining the geology of the region. Although direct measurements are generally preferred to indirect measurements have proven extremely valuable in many applications. Careful analysis of field data from geophysical measurements, and correlation with known geologic conditions, should provide significant knowledge of earth formations. Geophysical methods do not directly determine depth, type and composition of underground earth materials, but rely rather on detecting differences in physical properties (elastic wave velocity, magnetic susceptibility, density or electrical resistivity) of such earth materials.

This work was initiated as an evaluation of the application pf rubber finger ring dosimeters to the control of extremity radiation exposure. The use of film dosimeters is a well known general practice throughout the radiation protection profession, however, film radiation exposure measurements associated with plutonium metal handling present a unique problem. The practical measurement of extremity dose depends upon the accuracy and reproducibility of results under the most severe conditions using a dosimeter of minimum size of the radiation exposure to the hands in the rubber finger ring.

The calculation of shielding the thickness for plutonium is complicated by the many different energies represented in the gamma radiation emitted during decay of the plutonium isotopes. Dose rate predictions are also frequently confused by gamma from varying content of fission product impurities in the plutonium, as well as other gamma radiation induced through alpha and neutron particle absorption within the source material or its environment. After assumptions are made for these many factors the radiation data for shielding determination is still frequently inadequate because of wide variations in dose rates resulting from self-absorption. The degree if self-shielding is in turn dependent on nature of the plutonium compound, degree of compactness, weight, and over-all geometrical distribution of the source material. By preparing a variety of plutonium samples representing combinations of these varying factors, actual dose rates and gamma spectra, as obtained from them, can then be extrapolated for application to specific situations.

aluminum has been considered as a material for cladding fuel elements to be used in reactors cooled by high-temperature recirculating water. Aluminum is abundant, chap, has a relatively low cross section for thermal neutrons, and is readily fabricated into complex shapes. For these reasons, it appears that the use of aluminum rather than steel, Zircaloy or other candidate materials for fuel element cladding, would result in considerable savings in operating expense. The actual economics will of course depend on specific factors such as core materials and design of the reactor/. Aluminum appears particularly attractive for cladding the cores of aluminum-plutonium alloy proposed for the Plutonium Recycle Test Reactor since, in this case, the physical characteristics of the core and cladding would be practically identical.

The relative neutron capture gamma production in several prospective iron thermal shielding materials for the New Production Reactor was measured to determine the merit of adding boron to the metal. It was found that for the beam geometry the used addition of 1 1/2 percent boron to the iron before casting reduced the soft gamma production by a factor of 6.5 and the hard gamma production by a factor of 10. No attempt was made to measure gamma or neutron transmissions.

Since the start of E-metal processing in December, 1958, critical mass control in the Redox dissolvers has been provided by a combination of administrative controls, revised dissolver charging procedures, and new dissolving processes and operation control procedures. These were all based on a maximum allowable charge of E-Metal to the dissolvers which was equivalent to 45 percent of the minimum critical mass.

A series of experiments has been conducted during the past eighteen months with the objective of degerming the accuracy with which precision parts can be duplicated using a castable tooling plastic as a test shape to be reproduced. A precision ground ground 1/4 inch-thick, 90° steel sector of 4'' radius was chosen. The plastics used in the experiments were the die surfacer mixes of the Devon Cooperation and Ren Plastics, Inc. These were selected for use in the test because of their dimensional stability and wear resistance. Ordinary paste type floor wax was used as a mold release. The first series of tests were primarily intended to gain facility in handling this material. After achieving some familiarity with the use of these castable plastics a technique was devised which seemed to hold promise of precision duplication.

In the field of nondestructive testing the ultrasonic test method is used extensively. In this method ultrasonic energy generated by a transducer is caused to enter the test specimen. The effect on the transducer is cause to enter the test specimen. The effect on the transmission or reflection of the ultrasonic energy by differing test specimen structure is used as an indirect measure of test specimen characteristics.

Maximum safe cylinder diameters, slab thicknesses, and minimum critical masses in spherical geometries for fuel of < 5.0& U-235 enrichment have been evaluated. It is of interest to know the minimum critical masses as a function of cylinder diameter for vessels that are not safe by geometry. Curves will be presented of critical masses as a function of cylinder diameter and fuel rod size for 1.6%, 3.0%, and 5.0% U-235 enriched metal-water systems.

Basic Studies. Sinterability studies on the isomorphous system UO2-PuO2 have continued. One-half inch diameter X 3/8'' long compacts containing physical mixtures of the two components have been heated in hydrogen. Sintered density as a function of time and temperature has been determined for times of one and eight hours up to 1600 C. In general, there seems to be minima on isothermal plots of density versus composition in the intermediate range 20-40 w/o PuO2. At concentrations greater than 40 w/o PuO2, density increases rapidly with increasing PuO2. In every case pure PuO2, (produced by decomposing plutonium oxalate at at 300 C) sintered to a greater density that ball ball milled PWR grade UO2. The data, for a one hour soak time, and tabulated in percent of the theoretical crystallographic density, can be seen below:

Neutron radiation can be greatly reduced by substituting plutonium trichloride for plutonium tetrafluoride in the dry chemistry step of metal preparation. Chloride processes are now common in the metal preparation industry, and the 234-5 Developmental Laboratory has shown plutonium trichloride can be satisfactorily prepared and reduced to the metal.

In many cases of analysis of two-phase flow in systems, considerable computation or program time could be saved if the critical pressures ratio were known. If a reservoir or plenum pressure is fixed, the usual computational procedure involves the assumption of several critical pressures and the generation of several momentum terms to find the applicable critical pressure ratio and thereby the critical flow. The formulation of an equation of state make it possible to compute critical pressure ratios directly.

Late in 1958 it became apparent that some fuel elements were failing in the Hanford reactors as a result of water entering through the weld. The mode of entry appeared to be first through a void in the weld, then through a non-wet area or a train of voids in the braze, and finally to the uranium core. Defective closures of a similar nature were also typical of many fuel elements which have failed in the autoclaving operation as shown in Figure 1.

The amount of dry air or inert gas supplied to a "one pass" glove box ventilation system is proportional to the glove box inleakage. Most glove boxes in the 234r5 Building are equipped with sundry attachments, each contributing to the inleakage. No individual leak rate date is available for these auxiliary components in the "as installed" condition. Nor is the effect of time upon the leak rates known. Knowledge of these values, or at least an indication of the order of magnitude of the leakage attributable to each item, would provide a basis for analyzing glove box ventilation problems and for establishing criteria for new glove box designs.

This is the second of two reports describing the phases of a study of the wear and galling of autoclaved Zircaloy-2 by various fuel element support materials. The first report (1) showed that increasing contact area between the autoclaved Zircaloy-2 support and the autoclaved Zircaloy-2 process tube will not significantly improve the fuel support performance, and that a large flat contact area between the fuel supports and the process tube would be impractical to endeavor to achieve and maintain. It also showed that metals softer than Zircaloy-2 do not scratch the process tube but that loss of these metals by abrasion into the reactor coolant might be a problem.

The decontamination of the present Hanford reactors involves a once-through cleaning operation. Considerable interest has been shown in determining the feasibility of this once-through technique for cleanup of certain portions of the NPR system. This is the second interim report of a series that covers tests performed in the 242-B Single Pass Flow Facility. The first was distributed in January, 1960.

Theoretically aqueous hydrazine will react with dissolved oxygen for form only nitrogen and water. Most of the experimental evidence, however, indicates that the reaction proceeds rather slowly in low-temperature systems. Tests are currently being conducted to determine whether or not it is feasible to use activated carbon to catalyze the reaction in filtered water. A system of this type would be useful in high temperature reactors when it becomes necessary to convert to single-pass operation.

In order to prevent dangerous concentrations of combustible gases such as hydrogen and ammonia in Redox dissolvers, a continuous analysis of the off-gases is needed. Hydrogen is evolved from most fuel element dissolution processes now used or planned for use at Redox. Ammonia is also released from ammonium fluoride dissolution of zirconium and caustic dissolution of aluminum cladding. Oxides of nitrogen and hydrogen are formed during nitric acid dissolution of uranium or plutonium-aluminum alloys. Since no fixed ratio exists between the concentrations of any two of these gases, the determination of combustible gases in complex mixtures is not possible with a single bulk property analyzer.