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Tungsten and graphite are unsuitable materials of construction for a UF/ sub 6/ inlet nozzle in a continuous DRUHM reactor. Preparation of feed was completed for an extended Fluorox test. Difficulties were experienced in the operation of a fluidized bed TbNT denitrator. Flame denitration of UNH and TbNO produced mixed oxides of 1 to 14 micron mean particle size. The chloride capacity of Dowex 21K was measured, and equilibria measurements of uranium sorption from sulfate solutions were continued. Siliceous deposits in the stripping column caused the termination of a Darex run with a prototype APPR fuel element. Hot runs were begun on the chemical dejacketing of irradiated PWR blanket pins. The addition of formaldehyde to neutralize "25" waste reduced the recovery of nitrate from the calciner off-gas from 76% to 23%. (For preceding report see CF-58-9-62.) (auth)

BS>Neutron nonelastic cross sections measurcd at different detector biases have been used to determine the cross sections for inelastically scattering 14-Mev neutrons into 9- to 14-Mev energy range. The cross section for producing these high-energy neutrons, which may be attributed to direct interaction processes, is roughly 10% of the nonelastic cross section, for all elements. A comparison is made with data of Coon and co-workers, who measured angular distributions for the samc high-energy inelastically scattered neutron group. (auth)

The flowsheet proposed by KAPL for the treatment of alkaline nitrate radiochemical processing waste has been modified to include an ion-exchange step for the decontamination of cesium, strontium, and other cationic fission products. In laboratory studies of the electrolysis steps in a nitrate reduction cell, synthetic alkaline waste, 0.55 to 4.83 M total electrolyte, was decontaminated from ruthenium by factors of > 210. The nitrate of 3.0 M waste was reduced to ammonia with current efficiencies of 100%. Power consumption was 4.7 kwh/lb of nitrate reduced. Significant factors affecting the rate of ruthenium decontamination were temperature, cathode area, cathode current density, and electrolyte concentration. Those affecting nitrate reduction current efficiencies were the cathode current density, electrolyte concentration, and stirring rate. In an acid-base membrane cell, reusable nitric acid as well as sodium hydroxide was regenerated. However, such a cell is less economical to construct and operate than the nitrate reduction cell. At least 5 kwh of power is required to transfer 1 lb of nitrate from ORNL type waste. In addition, the greater complexity of the acid-base cell makes it less adaptable for remote control.

Cold-worked specimens of iodide zirconium, Zircaloy-2, iodide titanium, and A-55 titanium were tested in a high-pressure recombiner loop in an attempt to duplicate anomalous results obtained in a prior recombiner loop. Hydrogen analyses and metallographic examinations were made on all specimens. The titanium materials and Zircaloy-2 picked up major amounts of hydrogen in the cell section. None of the materials tested showed appreciable hydrogen absorption in the recombiner section. Complete recrystallization occurred in all cell specimens while only Zircaloy-2, of the recombiner specimens, showed any degree of recrystallization. No explanation for this behavior can be given. A survnnary of the data obtained in previous recombiner loops is compared with the results of this loop. Conclusions were based on the results of three recombiner loops. Primarlly because of the hydrogen absorption data obtained in all three recombiner loops it is recommended that the zirconium and titunium materials tested not be used in environments similar to those encountered in high pressure recombiner loops. (auth)

Because of wide variations in the venturi throat pressure, well beyond the panellit gage trip range, that occur during the sequence of operational charge-discharge, the panellit gage cannot be included in the scram safety circuit during the period of time that charge- discharge operations are being performed. In its stead, the function of the panellit gage is replaced in an overlapping manner by a tube inlet pressure monitor that is equipped with high and low pressure trip mechanisms that may be included in the scram safety circuit during the time that the panellit gage must be by-passed. The tube inlet pressure monitor is then used to provide the protection from unstable flow that is normally obtained with the panellit gage. This memorandum describes the manner in which the tube inlet pressure monitor trip points are to be determined and used.

This report describes the operation of the fuels preparation department for the month of February, 1958. Manufacturing employee relations, process development, plant improvements and financial operations are described.

This report describes the operation of the fuels preparation department for the month of July 1958. Manufacturing, employee relations, process development, plant improvements, and financial operations are described.

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The General Electric Company has prepared in past years numerous documents for the Atomic Energy Commission outlining in some detail its current and proposed programs at Hanford. Last year, at the request of the commission, program types of information were consolidated into one document, the HAPO Five Year Program. That document was well received and useful to an extent that the General Electric Company was requested by the Commission to prepare another covering the fiscal years 1959 through 1963. In outlining the five year program for the Hanford Atomic Products Operation, the General Electric Company has made assumptions regarding the US Atomic Energy Commission`s current and future programs and Hanford`s relationship to these programs. Two major objectives of the Commission were considered basic for future planning: first, to provide adequate atomic weapons for national defense; and second, to encourage and assist in the development of atomic energy for peaceful uses. The Hanford contributions to these objectives are conceived to include the acquisition and potential application of new technology in support of: (1) Increased production of plutonium, as nitrate solution, buttons, and as current and future shapes. (2) A program that will provide the technological capability to implement broad changes in …

Loop L-2-19 operated in HB-2 of the WTR for 1148.9 hours of circulation and 2705 Mwhr of LITR operation. The solution contained 0.17 m enriched UO/sub 2/ SO/sub 4/, 0.02 m CuSO/sub 4/, 0.12 m excess H/sub 2/SO/sub 4/, and 0.20 m Li/sub 2/SO/sub 4/ in H /.sub 2/O, The main stream was 280 deg C and the pressurizer 295 deg C. There was no unusual difficulty during loop operation. A revised loop sample procedure gave better flushing of the sample lines. Four valves failed due to severe corrosion and fracturing of the 420 SS stems. 1-131 was not detected in a sample of gas taken from the gas hold-up tank. The run was terminated when the circulating pump failed. (auth)

Using the output of programs 653 and 654, each card of which contains the dose information for one dose point at one energy, program 655 produces the sum of the dose values of all energies for one dose point. Program details are given including 650 input and output formats, operating instructions, and guides. (auth)

Material balance in atmospheric-pressure water and steam corrosion of uranium have been studied by examination of the phase composition and valence state of the corrosion product and by hydrogen-evolution measurements. The corrosion rates in atmospheric-pressure steam above 100 deg C are lower than those obtained in tests carried out in water with a hydrogen overpressure. The atmospheric-pressure-water corrosion product was found to be two phase: an oxygen- rich oxide, UO/sub 2.2/, and uncorroded metal particles. No hydride phase was detected, in contrast to previously reported evidence for hydride in uranium corrosion. The differences are explained on the basis of hydrogen pressure in the reaction vessel. (auth)

This document details activities of the irradiation processing department during the month of August, 1958. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

This document details activities of the irradiation processing department during the month of November 1958. A general summary is included at the start of the report, after which the report is divided into the following sections: research and engineering operation; production and reactor operations; facilities engineering operation; employee relations operations; and financial operation.

Liquid-solid fluidization studies at room temperature were initiated to evaluate the severity of flow instabilities associated with fluidized beds. In these exploratory tests, bubbling a moderate amount of gas through the fluidized bed completely disrupted the flow pattern. Inclining the cylindrical container at angles up to 20 deg from the vertical did not appear to affect the stability of the upper interface between the fluidized bed and disengaged water; however, at the larger angles pronounced channeling of the water was observed. Agglomeration was indicated in hindered settling tests even though the mean primary particle diameter was about 15 microns. Since the fluidization and hindered-settling velocities were essentially equal, the agglomeration may have existed in fluidization tests also. None of the above instabilities are believed to be sufficiently serious to eliminate the use of the fluidization technique in the design of nuclear reactor cores. Based on heat transfer considerations, calculations indicate that ThO/sub 2/ particles 20 to 30 microns in diameter would be of interest for large, single-region fluidized-bed rcactors. Thorium oxide particles of 150 to 1500 microns diameter would be of interest for the blanket region of two-region breeder reactors. (auth)

In December 1957, the proposal for Project CG-772, ``E-Metal Dissolvers -- Redox`` was returned for reconsideration in order that the units might be designed to accommodate higher enrichment fuels, if practical. Alternate design approaches have been under consideration since that time. It is the purpose of this document to summarize the parameters involved and to present the recommendation developed from the joint thinking of Hanford Laboratories, Research and Engineering, Facilities Engineering, and Redox Operation.

During the period from May 1957 to July 1958, four nuclear batiery powered timers were fabricated and tested from two basic designs in the time ranges of onesecond, three-second, annd half-hour intervals. The timers were temperature-tested over a range of -65 to +165 F with accuracics over this temperature range from plus or minus 10 perceat to plus or minus 15 percent. Each unit has a volume of 10 cubic inches, and the timer can be initiated either by an explosive squib or a pull-out wire. At the end of the timing interval, the timer has ann output of 30,000 ergs. The cost of the program was ,000. From the results of this development program, it appears quite feasible to build operable nuclear battery powered timers on a production basis. (auth)

A commercial light-weight portable bandsaw was fitted with a grip to permit it to be maneuvered remotely in a hot cell by means of a General Mills manipulator The bandsaw was supported in various positions to make cuts on typical pieces. Photographs show the saw in operation. (auth)

A proposed development program is made for the treatment and disposal of radioactive wastes from Merchant Ship Reactors. Both ion exchange resin and filters are to be used for coolant system cleanup. The development program is concerned with methods of handling and packaging of the spent resin, filter cake and rare gases for safe disposal. (auth)

Techniques were devised for quantitatively determining the accuracy of potentiometric uranium analyses in uranium-aluminum alloys containing up to 55 wt. % U and for evaluatin; the segregation existing in uraniumaluminum alloys containing as low as 7 wt. % U and as high as 50 wt. % U. A theory for predicting the mode of uranium segrcgation in these alloys was postulated. On the basis of the observed uranium segregation, the uranium content of a hypothetical fuel element was predicted by means of several sampling schemes. Dip sampling of the melt was demonstrated to be satisfactory for alloys containing 7 to 19 wt. % U. However, this technique was not considered suitable for alloys containiog 40 to 50 wt. % U, because a significant number of samples is required from the casting or the wrought alloy to adequately represent the fuel content. (auth) An accurate and rapid method for the volumetric determination of uranium has been developed for the ranges 0.3% to 90% uranium. The existing mercury cathode deposition equipment has been modified for the rapid removal of metallic impurities and for the electrolytic reduction of the uranium. (auth)