The filtrate stream from the precipitation--filtration step of the Button Line process must be recycled to recovery operations to reclaim residual plutonium. Present practice is to transfer the filtrates to Recuplex, concentrate by a factor of 6 to 10, and blend into the feed stream for solvent extraction purification. Limited tankage is available for solution storage. Interruption of Recuplex operations necessitates termination of operations in the precipitation-filtration steps and other Button Line areas. Filtrates can be recycled to one of the parent solvent extraction plants for recovery. However, excessive handling operations are encountered by this processing method. This document describes the anion exchange flow sheets and equipment for processing the filtrate stream without dependence upon other recovery operations.

This report, for September 1962 from the Chemical Processing Department at HAPO, discusses the following; Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; and employee relations.

In the operation of 6-in.-dia. foam-liquid columns, increase in either the liquid flow or foam flow caused increased channeling with increased HTU. Flooding values for the Immi mixer-settler were obtalned using the amine extraction final cycle Pu flowsheet. Two tests of the addltlon of uranyl nitrate to a thoria sol were made, one of which gave a good product and the other a product which disintegrated during calcination. Material balances for 22 waste calcination runs are summarized. (auth)

An interfacial viseometer was built for use in an interfacial phenomena study. Installation of a 6-in.-ID foam separation column system was completed. The dispersiondrying-sintering characteristics of six low-nitrate batches of thoria sol material were studied. The average effective porosity of the CuO pellets used for reactor helium purification was determined to be 0.0545 for H/ sub 2/ transport and 0.0526 for CO transport. In continuous Zirflex dissolution studies, no H/sub 2/O/sub 2/ decomposition was observed when 10% H/sub 2/O/sup 2 was fed into boiling dissoivent through a water-cooled nozzle and the oxygen concentration in the scrubbed off-gas could be used to control the H/sub 2/O/sub 2/ concentration in the dissolver. The free fluoride in Zirflex solutions must be maintained above 1 molar in order to prevent uranium precipitation at low concentrations of uranium even though the F/sup -//U ratio exceeds 100. Chopped stainless steel-clad UO/sub 2/ sections were leached in a 4 stage pyrex leacher model using 6, 7, and 8 M nitric acid as the dissolvent. The temperature distribution expected within fuel elements consisting of square arrays of tubes was calculated for shipping conditions assuming heat to be transferred only by radiation. HETS values were calculated for uranium stripping …

The economics of mechanical shear-leach and Sulfex decladding-core dissolution head end treatments for processing typical tubular bundles of stainless steel-jacketed UO/sub 2/ nuclear fuels were compared. A 2.66 metric ton U/day head end portion of a plant was designed for each process and capital and operating costs were developed. For plants of this size and larger, mechanical shear-leach processing has the advantage of ~20% lower capital cost and 50% lower operating cost. Processing costs of stainless steel-jacketed UO/ sub 2/ by the Sulfex and shear-leach methods, including amortization and waste disposal but excluding solvent extraction, were .78 and 7l/kg U, respectively. Storage of stainless steel waste produced by the shear-leach method is less costly by a factor of 20 than for Sulfex. (auth)

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Materials with improved corrosion resistance are needed for use in processing equipment handling HNO/sub 3/ -- HF and other corrosive solutions. Investigations were conducted to evaluate special modifications of Types 309 and 310 and higher Cr and Ni alloys. The 0.02% carbon Type 310 Cb, 0.02% carbon Type 310, and all of the higher Cr-Ni alloys were found to have a much improved general corrosion resistance over Type 309. The 0.02% carbon Type 310 Cb is recommended since it is not attacked in weld heat affected zones. (D.L.C.)

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A Monte Carlo IBM-7090 program (OGRE-P1) was written for calculation of the dose rate on one side of a slab owing to an isotropic, cosine, or collimated monoenergetic gamma radiation source on the other side of the slab. A maximum of 50 homogeneous regions are permitted. (auth)

The over-all design and operation of ORR is reviewed in the light of four years of operating experience. Items discussed consist of the reactor components and instrumentation, reactor and pool cooling systems (including system cleanup), the emergency systems for electric power and reactor cooling, the waste-disposal systems (liquid, gaseous, and solid), and the building, itself. The ORR was the first of a class of reactors which combined the features of both the pool-reactor and tank-reactor types. Four years of operation indicated areas where problems of various degrees have developed. These problems are discussed. (auth)

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The effect of polarizing mngnetic field intensity on the nuclear resonant absorption was studied by varying the field strength at a 1-mc Co/sup 57/ source from 0 to 1000 gauss while keeping the absorber between the poles of a magnet having a fixed field of 800 gauss. The rates of resonance absorption change with field intensity were greatest in the region of 300 to 1000 gauss, and the% nuclear resonant absorption for 1000-gauss fields was 8.5 and 26% for perpendicular and parallel fields, respectively, as compared with 15% for no fields. Other absorption measurements for Co/sup 57/ sources are also reported. Calculations on the use of nuclear resonant absorption to measure gravitational fields and altitudes were made which indicates that this application is not promising. (D.L.C.)

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Detailed measurements were made of the fast-neutron and gamma dose rates and the thermal-neutron fluxes existing at the surfaces of the biological shield of the University of Illinois TRIGA Mark II Research Reactor. Dose rates and fluxes were found to be extremely low. A comparison was made by means of threshold foil techniques between the fast-neutron flux in a beam hole of the reactor during steady-state operation at 1 kw and during a nominal 250 Mw, 30 msec pulse produced by rapid insertion of approximates two dollars of excess reactivity. The flux over the duration of the pulse was approximately four times that for steady-state operation for 1 hr at 1 kw. Fast-neutron dose rates and thermal-neutron fluxes were measured at each of the four beam ports of the reactor at operating powers of 0, 1, and 5 watts. Some gamma dose rate data was also obtained at 0 and 1 watt. The threshold foil technique used is described. (auth)

Overbore fuel for charging under the proposed Plant Improvement Program is fabricated according to the standard ``F`` process with a few exceptions. Due to the size and weight of individual components and cores, it was necessary to modify the process and change several of the process times to produce a fuel of equal or better quality than standard production. The additions and modifications to the standard lead-dip fuel element manufacturing process are required for production of Model CVIN fuel elements and are discussed in this report.

A transistorized instrument prototype was designed and constructed to replace a vacuum-tube instrument in an alpha gauge, which measures the thickness density of gases. The instrument amplifies, shapes, discriminates, and counts alpha pulses from a Au-Si surface-barrier detector exposed to an alpha source in a gas-filled chamber. The circuit consists of a charge-sensitive preamplifier, a main amplifier with pulse clipping, a Schmitt trigger, a diode pump, and a count rate meter. Preliminary tests gave results comparable to the vacuum-tube instrument. Accuracy of counting was within 10% for 0.5- to 10-Mev alpha particles emitted at a maximum rate of 10/sup 6 per sec. The instrument was stable at 25 to 55 deg C, is small and portable, and costs less than 0. An infinitely thick, alpha source that will give a high count rate is being constructed for final tests. (auth)

Containment of the ORR is attained by means of a ventilation system which is capable of exhausting air from the building at a rate sufficient to ensure that all leakage of air at ground level is into the building. The exhaust air is treated by scrubbing and filtration and is discharged from a large stack at a height and velocity sufficient to guarantee that meteorological dispersion will reduce the resulting concentration of radioactive material to an acceptable level. Factors involved in selection of this type of controlled containment are discussed. Original design and modifications are described. Performance is also discussed. (M.C.G.)