The work to determine the feasibility of a pratical solution counter for alphas has been continued through this month. It has developed that the electronic technique using an Amperex mica-window GM tube (1.4 mg./cm{sup 2} mica) is reliable, and it now appears that the problem is mainly one of solution-mounting technique. Two four-input gamma mixers and three two-input B-wall mixers have been given a final check and assembled in cabinets with their associated scalers. These scalars, Instrument Development Laboratory Model 161-S, were modified for use with either type mixer and may also be used for straight GM counting. One Instrument Development Laboratory Model 161-S, scale of 256 gamma counter has been given performance tests before being placed in operation at Mound Laboratory. Tests are being run on the parallel plate alpha counter Model K-262.

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This report is a descriptive journey of the Activation Analysis of the Rare Earths.

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This document presents a summary of work and progress at the Hanford Engineer works for October, 1955. The report is divided into sections by department. A plant wide general summary is included at the beginning of the report, after which the departmental summaries begin. The Manufacturing Department reports plant statistics, and summaries for the Metal Preparation, Reactor and Separation sections. The Engineering Department`s section summarizes work for the Technical, Design, and Project Sections. Costs for the various departments are presented in the Financial Department`s summary. The Medical, Radiological Sciences, Utilities and General Services, Employee and Public Relations, and community Real Estate and Services departments have sections presenting their monthly statistics, work, progress, and summaries.

The reference letter documented the plutonium concentration flowsheet in use at that time and proposed that this flowsheet be used for the scope design basis of replacement equipment. A study of the application of the reference flowsheet to the proposed remote package design (1) indicated that attainment of a plant capacity factor of four would be doubtful. This letter outlines a flowsheet reconciling process and equipment limitations and capable of a capacity factor of four.

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The oxidation of Hg in HNC/sub 3/ was investigated. An empirical rate law, oxidation rates, and Hg species in solution were determined. A reaction mechanism based on the presence of nitrous acid can be interpreted in terms of the rate expression derived. (auth)

Numerous types of high temperature ceramic fuel elements for the Pebble Bed Reactor are being evaluated. Specimens are 1 1/2 in. diameter uranium graphite spheres with external coatings such as silicon carbide or pyrolytically deposited high density graphite and fuel particle coatings such as alumina. Low fission product leakage rates at high temperatures have been observed for some of these coatings. High-level irradiation has given no visible evidence of radiation damage to either the silicon carbide coating or the coating-graphite bond. (auth)

This document details the activities of the Fuels Preparation Department during the month of October 1959. (FI)

A quarter-scale flow model, using air as a working fluid, was used to obtain design data for the PEGCPR program. A design for the core-support cylinder, to provide optimum mixing and core-flow distribution, was developed, following which experimentul studies of core-flow distribution, mixing of flow from the two inlets, flow patterns in plenum spaces, flow patterns in the thermal- shield-coolant passage, and pressure drops throughout the model were carried out. Results obtained in the model were then converted to values applicable to helium flow in the prototype. (auth)

During the course of power run 14 on the Sodium Reactor Experiment (SRE) at low power, the temperature difference among various fuel channels was found to be undesirably high Normal operating practices did not succeed in reducing this temperature difference to acceptable values and on July 26, 1959, the run was terminated. A series of fuel element inspections was begun to ascertain the cause of these circumstances, and several fuel elements were discovered to have suffered substantial damage. On July 29, 1959, an Ad Hoc Committee was appointed by Atomics Intennational to assist in the analysis of the existing situation in the reactor and the determination of its origin. During the three-month period since the termination of power run 14, there has been a very active program of investigation. The data accumulated duning the operation of the SRE have been re- examined and evaluated. Metallurgical examination was made of a few samples of the fuel and other components of the reactor where possible. Some chemical analysis was made of the coolant and its contaminants. Radiochemical analyses have been made of the coolant and gaseous activity. Reactivity effects were investigated. Scme experimental programs were initiated to examine mechanisms of damage and …

The mean linear and cubical coefficients of thermal expansion of eight commercial samples of graphite were determined for temperature intervals between 80 and 2000 deg F. The linear thermal expansion was measured with an automatic recording dilatometer using a rod-shaped specimen 2 in. long and 1/4 in. across. The specimen was heated in an atmosphere of helium. The results were in good agreement with those of Currie, Hamister, and MacPherson. The mean linear coefficient was found to increase with temperature. For the samples studied, the mean linear coefficients from 80 to 2000 deg F were 1.50 to 2.34 x 10/sup -6// deg F parallel and 2.26 to 3.45 x 10/sup -6// deg F perpendicular to the grain and were found to vary linearly with the electrical resistivity measured at 32 deg F. (auth)

The hydrodynamics code BOCH was used to obtain the relationship between a large number of variables such as voids, flux, pin spacing, power density, and pressure for VMR lattices. Based upon these relationships, a reference core design was selected for examination, using the integrated physics and hydrodynamics analytical modeis. The analog representations of the VMR core kinetics describing the individual blocks of a model block diagram were completed. A simplified core physics analysis was completed, and design parameters were established. An analysis of these parameters using PUREE' was initiated. The first three energy groups and the thermal group were compared with experimental lattices. Portions of the developed code are being used to analyze the critical experiment design and to investigate the reference core design. The critical experiment facility design is about 60% complete. The preliminary program which has been planned includes consideration of the number, kinds, and schedule of experiments to be performed. Portions of the PUREE' physics analysis are being used to generate information for the hazards repert on the critical experiment. (For preceding period see ATL-A-100.) (J.R.D.)

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Development studies were made of methods for producing adhesives and plastics having characteristics superior to those currently in use. Processes are being developed for 230 sources of supply are being evaluated. Research was initiated to determine the density, viscosity, thermal capacity, thermal conductivity, and phase equilibria of plutonium and plutonium alloys proposed as fast breeder reactor fuels. Work is being conducted on the fabrication of plutonium-bearing glass fibers. (B.O.G.)

The design arrangement of the Experimental Gas Cooled Reactor contains six fuel assemblies stacked vertically, above one another, in each of 232 vertical fuel channels in the graphite core. Each stack of six fuel assemblies is preceded and followed by a dummy assembly. The upper dummy assembly contains a reflector plug and a latch mechanism. The lower dummy assembly contains an orifice with an adjustable plunger capable of varying reactor coolant flow through the orifice. The pressure drop characteristics of the dummy assemblies for eight different throttling positions were obtained in the test program at near atmospheric pressure using an air flow corresponding to Reynolds Numbers of 12,000 to 60,000 in the fuel element assemblies. The data of the tests are presented in the form of velocity head coefficients expressed in terms of the highest velocity head of the two test locations involved for each coefficient. (auth)

To permit either the irradiation of two defected UO{sub 2} four-rod-cluster fuel element assemblies during the same operating period, or the irradiation of the two assemblies singly during two different operating periods in KE Reactor front-to-rear test hole 3865. The density of UO{sub 2} in one fuel element assembly is 90% of theoretical, in the other 87%. In both assemblies the fuel element cladding is Zircaloy and the fuel material natural uranium. An artificial defect will be made in each assembly before irradiation by drilling a .005 in. diameter hole through the cladding near the mid-point of two of the rods in each assembly, thus a total of eight UO{sub 2} rods, four defected and four nondefected, will be irradiated.

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A pump was developed for pumping carbon dioxide in a closed loop without introducing impurities. This pump will give flow rates of up to 3 liters/min and will develop a working pressure of over 70 mm Hg. No wear was observed after 2000 hr of testing. It is felt that this pump is more desirable for this application than those developed by other experimenters for two reasons: relatively inexpensive construction of the pump and the associated electronic circuit and the low coefficient of friction between the piston and the cylinder wall. Various modifications are suggested which will make this pump satisfactory for other applications. (auth)