A report which deals with the Redox IA Column studies made in small-scale columns packed with 3/16-in. Fenske helices and using NH4NO3 salting agent.

Introduction: Static, total immersion corrosion tests were carried out at room temperatures with Carboloy grades 44-A, 78, 907, X-3119-A and X-3119-B in Redox process solutions (ANL, June 1, 1949 Flowsheet) 1AX, 1AF, and 1AS.

Abstract: "The critical mass both of an untamped and water-tamped sphere with parabolic radial variation of the intrinsic buckling is calculated by a perturbation method. The result is applied to finding the minimum critical mass of plutonium in water suspension with infinite water temper; the calculations show that the minimum critical mass at constant concentration may be reduced by an amount of the order of 7.6% when the concentration of plutonium is permitted to vary throughout the suspension."

Summary: "Criticality was considered impossible in any contingency encountered in practice under either of the following conditions which include no safety factor: (a) In any isolated cylinder not more than 5 in. in disruptor, if the total amount of U235 present does not exceed 7.5 Eg. (b) In an isolated vessel of any shape and size, if the total amount of U235 present does not exceed 880 g. An "isolated" vessel is one which does not "see" more than 0.1 [...] other vessels containing fissionable material."

The following report examines methods to determine uranium nitrate hexahydrate on pretreated hexone in the range of 0 to 40 g./l. by infrared absorption measurements.

Introduction: "A laboratory investigation of the relative corrosion resistance of several austenitic stainless steels to the RAW stream of the Tributyl Phosphate Waste Metal Recovery Process was made during the first quarter of 1950. The results of this investigation are reported herein; all original data are recorded in laboratory notebook HW-3282-T."

Objective: "The observations summarized in the present report were made to determine the stability of ferrous and sulfamate ions and sulfamic acid in RA Column scrub solution (RAS) and in aqueous solution comparable to the RA Column feed tee composition (RAFS). Knowledge of the stability of the reducing agent (ferrous iron) is necessary to properly plan the method of preparation and introduction of this component into the RA system and to assure dependable and reproducible performance."

From introduction: "Analytical procedures for the determination of plutonium samples from the extraction and concentration plants utilize the carrying of the hydroxylamine-reduced trivulent plutonium by coprecipitation on lanthanum fluoride. Most other plutonium procedures also take advantage of the efficient carrying on lanthanum fluoride."

Introduction: "A corrosion test program, to determine the suitability of LaBour B-55 and Worthite for use in the construction of circulating pumps for th UO3 process has been completed and the data are appended."

Introduction: A limited test progress on Alloghany-Ludlum type 347 stainless steel plate which failed the Huey test (ASTN A-262-44T) has been completed.

This report follows an investigation on how hydrolysis products of tributyl phosphate (such as dibutyl and monobutyl phosphoric acids, orthophosphoric acid and butanol) effect the tributyl phosphate process for uranium recovery.

Recent developments in the field of photoconductive devices made possible the simplification and attendant improvement of the recording anemometer system used with the portable meteorological mast. The purpose of the work was to produce a simple, reliable recording anemometer system compatible with the project's existing instrumentation. Of primary importance was reduction of maintenance without loss of accuracy. This report discusses some of the factors considered in the design of the system, describes in detail the electronic circuitry and mechanical components, and presents some performance data. The data obtained from a twenty-four hour trial run indicates that the system has fulfilled the requirements.

This report has been prepared to review the status of design for buried waste line and presents the major factors involved in support of the conclusions. This report is the second in a series reporting on the general subject of underground corrosion. The first on cathodic protection of waste lines noted the technology of stainless steel and its effect on existing waste line and recommended continued use of the cathodic protection. Experience with buried stainless pipe had shown that, unless protected, such installations are subject to severe pit corrosion. As designs developed the use of encasements has become a standardized approach. This discussion will evaluate this trend.

Buckling measurements were taken with hollow natural uranium slugs in graphite-moderated lattices, in which the lattice spacing was varied from 6 3/16 to 15 inches in five steps. A 1.66" O.D. slug was used with two different core sizes, 1.11" I.D. and 0.81" I.D. The results of wet (water in the core) and dry measurements are reported. Buckling curves are given showing the effect on the buckling of coring the slug and placing water in the core. Simple three-region diffusion theory is applied to the measured lattices in an effort to correlate measurement with theory, but with only limited success.

Startup and testing of the PRTR has been divided into three parts, Design Tests, Critical Tests and Power Tests. This document is the third of the series and describes the performance tests which will be made under nuclear power and the general schedule of operation. Responsibility for the startup of the reactor has been assigned to the PRTR Startup Council. The Power Test Sub-Council, which has done the planning of the initial power operation, was established by the Startup Council. Operation of the reactor will be the responsibility of the PRTR Operation (PRTRO).

An electrical conductivity technique was developed to determine the dilution-displacement rate of an electrolyte in a well from which data the velocity of the water through the well can be calculated. The electrical current flow between two of the electrodes in a well drops proportionately with the dilution and the displacement of the electrolyte by fresh water entering the well. The drop of electrical current flow over a period of time was applied to a derived equation to obtain a value for ground water velocity. The technique gives results comparable to other methods and some problems of the standard methods of measuring ground water velocity are eliminated. However, a correction factor for porosity of the aquifier must be used in the derived equation. Porosity values are not adequately known in most cases, therefore the ground water velocity figures are generally relative rather than absolute.

This twenty-second quarterly report of the research and development activities of the Radiological Sciences Department, Hanford Atomic Products Operations, as before, includes some items charged to control but included for general interest. Such are identified as "not charged to research." The previous report of this series was numbered: HW-34408 January 10, 1955

It is recommended that self-concentration of Purex wastes should not be started until much more complete knowledge is obtained of the conditions and reactions in these wastes. The presently planned tests in the Semi-Works will be a practical demonstration of the problems but will be of little value on the initial Purex wastes as these tests are being run almost concurrent with the filling of the first Purex storage tank.

The purpose of this work was to produce a reliable, inexpensive instrument for the detection of hard beta contamination on the clothing of personnel. Also, the monitoring was required to be accomplished without interrupting the normal activities of the employees.

The build up of contaminated film on the internal surfaces of high temperature in-reactor recirculating water loops has created serious radiation exposure problems to operational and maintenance personnel. A considerable amount of work has been applied to develop an effective decontamination process for the decontamination of these loops and their components.

The Atomic Energy Commission has recently developed an enlarged program for the study of graphite moderated, gas cooled power reactor systems. It has been recognized, however, that understanding of radiation damage and radiation induced chemical reactions of graphite at the proposed high moderator temperatures is inadequate and that improved understanding is essential if the design of such reactors is to be optimized. Accordingly, the Atomic Energy Commission requested Hanford to organize a modest research and development program directed toward the study of these graphite problems.

The temperature characteristics associated with the handling of self-boiling radioactive wastes from the separations extraction processes in the Chemical Processing Department have necessitated several heat transfer studies. Earlier studies 1,2,3 defined the feasibility of self-concentration in existing waste storage facilities by determining the rate of heat generation from the decay of stored fission products and by defining the rate of heat loss from existing storage tanks to the surrounding soil.

Laboratory studies involving fluid flow through porous media require use of fluids having low dissolved gas content. Water is the major fluid used in various and box model and soil permeability studies carried out by the Geochemical and Geophysical Research group. Tap water supplied to the 222-U Bldg. contains a large amount of dissolved air. Under the reduced pressure encountered during model studies, the air is released from solution and gradually clogs the pores of the sand or other porous material. This, of course. leads to anomalous results and cannot be tolerated in precious studies. A system was required to effectively remove the air and make available a continuous supply of desired water for the model studies.

"The PRTR Pressured Gas-Cooled Loop Facility, or Gas Loop, is an experimental facility to be installed in the Plutonium Recycle Test Reactor for use in studies contributing to advancement of the technology of gas-cooled reactors. The facility will provide an in-reactor loop for studying phenomena occurring under conditions likely to exist in gas-cooled reactors.