Over-all mass transfer coefficients were determined independently of the interfacial area by considering the operation of the pulse column to be a stage- wise process. The calculated coefficients describe only the operation of the pulse column for the free rise of organic droplets through a stage, since in the downstroke the aqueous phase passes through the organic phase in the form of rivulets preferentially wetting the plates. The driving force for transfer was based on the aqueous phase concentrations. The mass transfer coefficient was found to increase when the interfacial tension was reduced. Two factors appear to affect the mass transfer coefficient as the pulse frequency is changed and are: a) recycle or backmixing which tends to decrease the coefficient through a reduction of acetic acid in both phases and thereby increases the interfacial tension, and b) turbulence which enhances the coefficient by reducing the continuous film resistance. The first is important at or near the lower flooding limit, but the latter is the controlling factor at higher frequencies. (D.L.C.)

The mutual inductance or apparent susceptibility due to the presence of the sample in a coaxial inductance coil was observed for Gd, Tb, and Sm in fields of a few gauss over the respective temperature ranges of 78 to 310 deg K, 78 to 235 deg K, and 4.2 to 150 deg K. A paramagnetic Curie point of 284.2 deg K was found for Gd. For Tb, a transition was found which occurred to within 1 deg of the predicted temperature of 220 deg K, and a second peak occurred at 229.4 deg K which is within 2 deg of the heat capacity peak. The 220 deg K peak exhibited hysteresis and time dependence. Extrapolation of Tb data to DELTA M = 0 yielded a Curie point of 235.6 deg K. No hysteresis was observed in the 14.8 deg K peak for Sm. (D.L.C.)

A detailed study of the fundamental relations in the zirconium -- hydrogen system was made in order to clarify the many points of dispute and to evolve a complete picture describing all phases of this system. An engineering evaluation was made of means for utillzing the various high cross-section metal hydrides in shielding or control applications. These materials would combine the processes of thermalization and absorption. Consequently, they are of considerable interest for use in shielding or controlling epithermal reactors. (auth)

A study of the extraction characteristics of the three systems lanthanum nitrate--nitric acid--water--tributyl phosphate, praseodymium nitrate--nitric acid--water--tributyl phosphate, and lanthanum nitrate--praseodymium nitrate nitric acid -water--tributyl phosphate was conducted. The separation factors between praseodymium and lanthanum for the system lanthanum nitrate--praseodymium nitrate-nitric acid--water--tributyl phosphate were shown to be a function of the total nitrate concentration of an equilibrium phase and practically independent of solute composition. A method of predicting the concentrations of lanthanum nitrate, praseodymium nitrate, and nitric acid in an equilibrium phase of the system lanthanum nitrate--praseodymium nitrate--nitric acid--water--tributyl phosphate was presented. A comparison of the extraction characteristics of the system rare-earth nitrate--nitric acid--water--tributyl phosphate for the nitrates of lanthanum, praseodymium, neodymium, and samarium was made. (auth)

A project to provide information on the hazards associated wlth arc melting of Th is described. A general airsampling analysis was made to determine the separation, concentration, and distribution of Th daughter (decay) products throughout arc melting, machining, and forging processes found in a handling facility. The value of well coordinated health physics program is stressed in connection with potential health hazards and personnel protection. Building, equipment, and exhaust ventilation requirements for such a facility are discussed, along wlth special handling methods. (auth)

An investigation was made of methods for decontaminating the primary loop of the New Production Reactor (NPR). Decontamination factors ranging from fifty to several thousand were obtained with several different combinations of cleaning solutions. In general, fission product activity was removed to a lower residual level than was activated corrosion product activity, and carbon steel was easier to decontaminate than was stainless steel. It was necessary to completely remove the oxide film from the metal surfaces to obtain adequate activated corrosion product decontamination. Descaling procedures were also satisfactory for decontaminating fission product and uranium dioxide contamination. (M.C.G.)

An envelope-type microstructure produced by liquidphase sintering techniques was used to develop ductiIe molybdenum-base alloys for sodium-cooled fast reactor application. A total of 105 compositions was examined; 21 showed evidence of room-temperature ductility, and a number of other compositions appeared promising. The prerequisites for obtaining ductile liquid-phase sintered materials were demonstrated. The most ductile alloys produced are based on the Mo--Pd--Cu and Mo --W--Ni--Fe systems. Molybdenum-base compositions were developed which could be rolled to 90% reduction in thickness at room temperature. The compactibility of selected alloys with uranium at 750 deg C and liquid sodium at 650 deg C was studied. In both cases the presence of iron and nickel was found to be deleterious. Although alloys compatible with uranium were produced, the specimens exposed to liquid sodium all showed evidence of attack to varying degrees. Two compositions based on the Mo--W--Ni and W--Cr--Ni systems, respectively, appear to hold promise as potential canning materials for sodium- cooled fast reactor application. (auth)

The recoil-proton polarization in {pi}{sup +}-p elastic scattering at 310-Mev incident-pion laboratory kinetic energy has been experimentally measured at four scattering angles with scintillation counters. Polarization values obtained, related rms experimental errors, and mean center-of-mass recoil angles are: +0.044 {+-} 0.062 at 114.2 deg. =0.164 {+-} 0.057 at 124.5 deg, -0.155 {+-} 0.044 at 133.8 deg, and -0.162 {+-} 0.037 at 145.2 deg. The sign of the polarization is defined to be positive when a preponderance of the recoil protons had their spin vectors pointing in the direction of {rvec P}{sub i} x {rvec p}{sub f}, where this quantity is the cross product of the initial and final momentum vectors of the conjugate pions. A beam of 1 x 10{sup 6} pions per sec incident upon a 1.0-g/cm{sup 2}-thick liquid-hydrogen target produced the recoil protons, which were then scattered by a carbon target at a mean energy varying with recoil angle from 113 to 141 Mev. The polarization of the recoil protons was analyzed by measuring the asymmetry produced in the carbon scattering. A proton beam of known polarization was used to determine the analyzing ability (measured asymmetry divided by the polarization of the incident protons) of the system at …

A comprehensive phase-shift analysis of {pi}{sup +}-p elastic-scattering data at 310-Mev incident-pion laboratory kinetic energy has been performed. The experimental data utilized include measurements of the differential and total cross sections and of the recoil-proton polarization. The D-wave phase shifts were found to be definitely needed in order to attain an adequate fit to the data. A general search for phase-shift solutions was carried out, using S-, P-, and D-wave phase shifts. One solution--of the Fermi type--was found that fits the data significantly better than any of the other solutions obtained. The calculated errors in the phase shifts of this set vary from 0.4 to 0.6 deg. Because it was felt that these errors might be deceivingly restrictive, the effects of small nuclear F-wave phase shifts on the results of the analysis were investigated and were found to be large; not only are the uncertainties in the original Fermi-type solution increased, but additional sets of phase shifts arise that fit the data well. One of these new solutions is similar to the original Fermi set except that the magnitudes of the phase shifts in this new fit are in general larger than those in the initial solution, and the signs …

The General Electric Company operates eight large filter plants for the Atomic Energy Commission at the Hanford works in the state of Washington. Because of the importance of water to the process, research and development on water treatment has been an important part of the overall Hanford research and development program. The research and development on water treatment has resulted in important capital and operating savings and in the production of better quality water. It is the purpose of this paper to present some of the information developed by the programs. 3 tabs.

This report, from the Chemical Processing Department at HAPO for October 1960, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

This document discusses the test program of evaluating the temperature effect of crud film build-up on fuel element heat generating surfaces in a carbon steel system. This program has three phases: Measurement of the effect of film build-up during normal high pH equilibrium operation; measurement of the temperature effect of film build-up subsequent to a loop decontamination; and measurement of the effect of film build-up in the event of loss of pH control.

In response to specific requests of the AEC and as part of a Combined Operations over-all review, a number of engineering studies have been made of alternative methods for increasing availabiltiy of neptunium at Hanford. The report updates an earlier study in which recycling of both natural and enriched uranium was considered for Hanford. The earlier study showed that recycled natural uranium would provide appreciable gains in neptunium availability but at an excessive cost. Recycle of the slightly enriched uranium streams proved a more economical means of realizing smaller but still significant gains in neptunium production. Subsequent to the earlier report, a feasible and immediately applicable scheme for UFI blending has been conceived demonstrating further advantages for recycling the enriched uranium. Approximately 34 kilograins of supplementary neptunium could be produced at Hanford during the next seven to eight years by upgrading irradiated E-metal and NPR uranium through a blending operation at Oak Ridge (rather than in the diffusion cascades) and then recycling the material through the Hanford reactors. (cf Table 1) Such a scheme would conserve uranium-236 for use as a source of neptunium-237 in the reactors without incurring major capital costs. Oak Ridge Operations personnel have estimated that capital …

This document is the third of a series of question and answer lists issued as a review of the material discussed in the Reactor Processing Fundamentals Course. Each document represents the material covered by the Reactor Specialists during a typical three-month training program. Each question is discussed individually and the entire list completed during the three-month session. Each three-month-training period is devoted to the complete discussion of a single chapter of the IPD Physics Primer series. The questions are compiled in a logical sequence with the material as presented in Chapter V of the Primer series control. As the course progresses, a certain amount of recall of earlier chapters is essential to a thorough knowledge of the specifics of reactivity and distribution control. Therefore the first half of this document consists of material previously presented in a slightly different manner. Basic points are stressed; the intent is, of course, that informative discussions lead to a better understanding of the material presented in the Primer.

The assumption of tube replacement losses can have a significant effect in the determination of optimum power levels and goal exposures. The tube replacement loss term in the reactor process optimization model is based on calculating the average projected tube replacement losses that will result from operation at given power and temperature conditions. Tube replacement losses associated with external corrosion, Van Stone flange failure, etc. (i.e., basically non-power level-temperature dependent) are assumed to be miscellaneous losses and are not included in the tube replacement term. Over a period of several years the experienced tube replacement losses (due to internal corrosion) should compare reasonably well with the losses predicted by the optimization model. Planned tube replacement project action which would require, in some cases, premature tube removal is a special situation which requires a modified approach to the prediction of tube replacement losses and to the optimization of reactor power levels prior to project action. A method has been developed for optimizing reactor power levels and goal exposures based on any assumed calendar date for major tube replacement project action. It is the purpose of this document to discuss the general application of this method in the optimization and illustrate the …

Test metal was irradiated to simulate a `blanket` loading and striped E-N load under PT-IP-255-A-9-FP. The product analysis has now been obtained on the test material. The interpretation of this data and plant direction that might follow is considered in this document.

In irradiations of rod cluster and tubular fuel elements and fuels rods in NaK-filled capsules, the Zircaloy-2 cladding has failed by necking in localized region. Purpose of the proposed tests to to help establish cladding thickness tolerances for NPR fuel elements, and to study the failure mechanism and influence of several variables.

Three fuel element assemblies, each containing U-Mo fuel plates, will be irradiated in a side test hole at KE-Reactor. This document presents efforts made to determine the heat generation rate of the fuel plates from data from a capsule mockup irradiation in HTR.

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Discussions presented in this report result from product development activity conducted for the Hanford Atomic Products Operation of General Electric Company, Richland, Washington, by the Research Division, Sciaky Bros., Inc. Program objective, discussion of development effort, and test results are presented with a summary of final weld schedule, conclusions and recommendations. The objective of the development effort was to define the optimum projection weld area configuration and the weld schedule to consistently resistance projection weld join a machined Zircaloy {minus}2 end closure (ring with projection/s) to the exposed (depressed) surface of a natural uranium tube clad (or contained within a double wall) Zircaloy {minus}2 tube assembly and to further process said assemblies by series seam welding the outer Zircaloy {minus}2 clad wall to the Zircaloy {minus}2 end closure simultaneously with the weld joining of the Zircaloy {minus}2 end clad to the inner Zircaloy {minus}2 clad wall.

The fixation of radioactive waste in a stable solid media by means of calcination of these aqueous solutions has been the subject of considerable-effort throughout the U. S. Atomic Energy Commission and by atomic energy organizations in other countries. Several methods of doing this on a continuous or semi-continuous basis have been devised, and a fev have been demonstrated to be feasible for the handling of non-radioactive, or low-activity, simulated wastes. Notable among methods currently under development are: (a) batch-operated pot calcination of waste generated from reprocessing stainless steel clad fuel elements (Darex process) and Purex waste, (b) combination rotary kiln and ball mill calcination of aluminum nitrate (TBP-25 and Redox process), and (c) fluidized bed calcination of TBP-25 and Purex wastes. Although a considerable amount of engineering experience has been obtained on the calcination of dissolved salts in a fluidized bed, and the other methods have been the subjects of a great deal of study, none of them have been developed to-the extent which would rule out the desirability of further investigation of other possible methods of calcination.

The objective of the production test detailed in this report is to evaluate the effect of a 36 mil Zircaloy-2 outer jacket on the behavior of heavy walled tubular elements during high temperature irradiation. Zircaloy-2 jacketed unalloyed natural uranium fuel elements nominally 1.43 inch OD, 0.52 inch ID, with 36 mil jackets on the outer surfaces will be irradiated in the KER Loops to an exposure not greater than 3500 WD/T. Several of the fuel element failures that have occurred during testing in the KER Loops have apparently resulted from localized jacket thinning on the external surfaces of high exposure elements. One of the methods proposed to prevent this failure mechanism was to increase the Zircaloy-2 jacket thickness on the fuel element outer surface. The fuel elements authorized by this production test have been prepared with nominally 36 mil external jackets. The behavior of these elements will be compared to the behavior of similar elements with 20 mil jackets currently being irradiated in KER-3.

This document details activities of the Hanford Operations Office during the month of November 1960. (FI)

The objective of this production test is to compare the dimensional and structural behavior of fuel elements with various heat treatments during irradiation at conditions similar to those anticipated for the NPR. Zircaloy-2 jacketed unalloyed natural uranium fuel elements of the same diametral dimensions an those anticipated for the N Reactor inner fuel tube, incorporating various heat treatments, will be irradiated to about 2000 MWD/T in the KER Loops. The elements will be enclosed in a Zircaloy-2 sleeve to provide a more balanced flow distribution through and around the pieces. An optimum heat treatment for N Reactor fuel elements providing minimum fuel distortion through control of uranium grain size and orientation has not been determined. This production test provides for the irradiation of fuel elements nearly equivalent to N Reactor inner tubes in size, geometry, and temperature conditions during irradiation. These elements have been subjected to varying heat treatments providing a range of grain sizes and degrees of orientation. Post-irradiation examination of the elements will show the fuel element distortion associated with each heat treatment and provide information on which to base further irradiation tenting in the development of a heat treating procedure for N Reactor fuel elements.