This study was initiated to determine the minimum safety requirements for safe anchorage of electric pole lines. The objectives were to investigate the suitability of manufactured concrete pole line anchors for use at the Hanford Atomic Products Operation, to design a concrete anchor slug of optimum shape and size for manufacture, to determine by field tests the resistance pullout of the concrete anchor when buried at various depths under the most favorable condition, and to develop a safe design load for the anchors at various depths of burial and angle of pull based on test results.

During an experiment involving alpha phase annealing (at 610 to 655 deg C for 5 and 10 min) of beta-quenched uranium, it was determined that the relatively coarse, irregular shaped grains (0.08 to 0.150 mm in diameter) in the beta-quenched structure, were replaced by fine equiaxed grains (0.04 to 0.09 mm in diameter).

A plutonium solution of known Pu-241 content was purified by solvent extraction techniques to remove uranium and americium impurities. Portions of this purified solution were then analyzed for U-237 as a function of time. Since the growth of U-237 occurs directly by the alpha decay of Pu-241, these measurements provided sufficient data for calculation of the Pu-241 alpha half-life, which was found to be 2.91 ± 0.50 x 10⁵ years.

Experimental studies on continuous denitration performed in 321 Building resulted in the development of an agitated-through-type calciner. The performance of a paddle-agitated 4-in.-diameter by 44-in.-long reactor was sufficiently promising to warrant construction of a 16-in.-diameter by 8 ft.-long semi-plant scale model. This progress report summarizes the results of studies to date employing the semi-plant scale calciner.

The desirability of operating many of the Purex columns at elevated temperatures, ca. 50-70 C, has made it necessary to obtain data on the stability of the solvent at these higher temperatures. Since the present diluent, Shell Spray Base, will not be entirely safe at the upper temperature limit, a number of diluents with higher flash point have been investigated concurrently with Shell Spray Base. This report presents data covering the first phases of this investigation, namely, the effect of nitric and nitrous acid on the solvent at 71 C. or higher temperatures.

The purpose of this investigation is to make a preliminary study and evaluation of the application of ultrasonic equipment and associated handling equipment to the problem of decontaminating reactor tube dummy pieces. Reactor Section is currently planning the construction of batch process nitric acid facilities for dummy decontamination. It was felt that use of ultrasonics together with a solution less corrosive than nitric acid might provide adequate decontamination at lower project cost than the presently proposed nitric acid facilities. This report gives the progress of the investigation to date.

A knowledge of the mechanical and physical properties of uranium is important in understanding and explaining its pile behavior. The effects of irradiation on the properties of uranium will be dealt with in another chapter. However, it is important to know and understand the pre-irradiation properties of uranium prior to investigating the effects of irradiation.

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.

Production Test 313-105-13-M authorized the irradiation of a uranium - 0.4 atomic per cent chromium alloy to determine the relative stability of this material during irradiation. Five tubes, each containing six four-inch alloy slugs, were irradiated to exposures ranging from 156 to 834 MWD/T. This report describes the results of the visual examination of the bare uranium alloy slugs and the results of the test to determine slug breaking strength. The breaking strength of both the U-Cr alloy slugs and the normal uranium control slugs was reduced by exposure in the pile. The breaking strength of those U-Cr alloy pieces not cracked or broken in the jackets before examination was consistently greater than the breaking strength of the normal uranium at all exposures.

As part of the recent intensified study of variables involved in the beta heat treatment of production uranium, the writer undertook a qualitative examination of the effects of heating and quenching media on the hydrogen content of pieces heat treated in slug form. Because of the limited amount of time available the work was necessarily restricted in scope. For the same reason no detailed study of any single heat treatment was attempted: rather, a number of beta treating schedules were selected so that estimates of the degree of hydrogen occlusion to be expected during a large variety of beta heat treating schedules would be possible.

The purpose of this report is to present a HAPO proposal for the design and construction of a high-temperature, high-pressure light water recirculation test facility for the Westinghouse Electric Corporation, Atomic Power Division This report does not attempt to justify the feasibility of site location nor does it attempt to elaborate on such technical considerations as permissible test specimen heat generation and enrichment requirements which have been covered in an earlier document. Instead, detail information is developed on the proposed design with major emphasis on detail descriptions of the proposed loop and auxiliaries, design and construction scheduling and costs, and operating costs.

Equations are developed for the computation of MPC's for radioisotopes assuming the excretion or retention to follow a power law rather than an exponential law. Computations are made for four isotopes on which some human data are available. Uncertainties in the initial data due to the use of intravenous injection to obtain the primary data, the definition of retention and the effects of age and diet are discussed.

It is the purpose of this document to outline a proposed addition to the previously considered clean-up systems as a possible solution for the maintenance of high purity water. A cyclone separator system has been devised in conjunction with the water clean-up of a high temperature recirculating reactor cooling system. By utilizing such a cyclone separator system, a high quality water can be obtained in a primary coolant system made up of carbon steel components without excessive clean-up costs.

Tests on twenty identically prepared 26 gauge Chromel-Alumel thermocouples have been completed. It appears that wires of this material without protection are unsatisfactory mechanically and thermoelectrically for service in atmospheres of CO₂ at 1000°C.

The Metallurgy Research Sub-Section at HAPO has conducted a research and development study of a variety of nuclear fuel materials. A summary of the status of the program currently being conducted on ceramic fuel material and a brief discussion of the future program is outlined below.

These initial operating procedures have been prepared for interim use during the first operation of Basins I, II, and III. They will be revised and included in the operating manual for the entire 105-C Metal Examination Facility.

In the analysis of problems involving transient flow and heat transfer in steam-water mixtures, it is necessary to derive an approximate integral equation of motion for a non-ideal compressible fluid in a duct having finite discontinuities in flow geometry. This equation is derived in this document.

Results of eight runs using in-line U photometers in organic and aqueous streams of the Purex Hot Semi-Works are presented. Their operation, both mechanically and electrically, was satisfactory, indicating changes in stream U concentrations over wide ranges.

A study was made to evaluate the possibility of an explosive reaction occurring in the Hanford Purex Plant concentrators as presently instrumented. Results revealed that the occurrence of an explosive reaction in the Purex Plant concentrators with present instrumentation is very unlikely.

periodically the Industrial Hygiene Unit has the occasion to perform efficiency tests on respiratory protective devices and on filter cartridges used with these units. A study of the relative merits of the flame photometer and conductivity meter was made to determine which is more applicable for the determination of NaCl in filter samples.

Flowsheet considerations and information developed for the large (size 2) Purex pulse generator⁽¹⁾ indicated the need for information on the operating characteristics of the small (size 1) Purex pulse generator.

An anion exchange precipitation procedure was developed for the determination of Cs¹³⁷ in various aged fission product waste solutions, and other solutions containing Cs¹³⁷. The procedure is based on the ability of an anion exchange column to function simultaneously at an anion exchanger and as a precipitating agent (1).

The purpose of this paper is to examine the theory and results of chemical exchange experiments for separation of heavy metal isotopes. The aim is to find a means of predicting those chemical exchange systems likely to produce large isotopic separation factors.

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.