Activities during the period were directed primarily toward determination of the engineering properties of hydrided materials. The tensile strength and elongation, thermal expansion, thermal conductivity, and heat capacity of ZrH were determined as functions of temperature and H content. Preliminary results of Ti-base material studies indicate that these materials can be used for shielding and structural materials with the incorporation of rare earth hydrides. A summary of the work from Oct. l958 to Sept. 1959 is included. (J.R.D.)

The work reported deals with the preparation and physical properties, especially thermal dissociation pressures, and densities of hydrides, hydrogen- metal systems, and mixtures of hydrides with other substances. Possible applicatlons as moderators, high-temperature neutron shields, and low-temperature shields are cited and design problems discussed. Most of the data on dissociation pressures cover ranges and compounds not hltherto explored because of experimental difficulties and the basic knowledge of the thermal behavior of hydrides was substantially increased. New hydrldes were prepared and several reported in the literature were shown not to exist. The following compounds, mixtures, and systems were studled: Tl-H, U-H, Ll-H, Na-H, Ca-H, Ba-H, Th-H, Sr- H; NaH-NaF, NaH-NaOH, NaH-CaH/, LlH-LiF, CaH/sub 2/-CaF/sub 2/, CaH/sub 2/-CaC/ sub 2/,CaH/sub 2/-Ca/sub 3/N/sub 2/; FeH/sub 3/ (alleged), NiH/sub 2/ (alleged), Ti(BH/sub 4/)/sub 3/, Th(BH/sub 4/)/sub 4/, WH/sub 4/ (attempted), W(BH/sub 4/)/ sub 4/ (attempted), /sub 4/NBH/sub 4/, (CH , and ydrides are ing an N/sub H/ comparable to water yet stable at red heat, compounds giving a neutron shield weight less than half that of water, and compounds suitable for use as hightemperature moderators containing large amounts of hydrogen. (auth)

As part of the program for obtaining data to extend the scope of the systematics of alpha-radioactivity and to obtain more data pertaining to the energy surface in the heavy region of elements, they have made some pertinent new measurements on protactinium isotopes. The heaviest isotope of protactinium hitherto reported is that of mass 234 and hence it would be of interest to know the beta-emission properties of heavier isotopes in order to tie them in with the known radioactive data in this general region. Since low-energy deuteron bombardment of U{sup 238} might be expected to lead to Pa{sup 235} and Pa{sup 236} by (d,{alpha}n) and (d,{alpha}) reactions and proton bombardment to Pa{sup 235} by the (p,{alpha}) reaction, these irradiations were made in the 60-inch cyclotron at energies of 19 and 9.5 Mev, respectively. The protactinium was chemically separated following the bombardment of natural uranium by a procedure which involved a number of manganese dioxide cycles coupled with extractions of protactinium from aqueous into organic solvents. The manganese dioxide cycles consisted of precipitating this compound from the solution of uranium in nitric acid, followed by centrifugation, dissolution of the solid with hydroxylamine solution, dilution, and reprecipitation. The dissolved precipitate from …

Management of first-cycle wastes from the Bismuth Phosphate Plant using Nickel Ferrocyanide scavenging is described.

This document provides tables which give the computed acid, silicate, and silica requirements which will be necessary after completion of the water treatment plant expansion programs.

Project C-431-A includes the design of the C Water Plant outfall structure. This report documents design assumptions and the design of the spillway, outfall sewers, spillway chute, trash racks, walkways, stop log and trash rack slots, main sewer, and retention sewers. The spillway capacity is to be 268,000 gpm with a crest elevation at 429.0. This flow capacity is composed of flow from the 105-C Retention Basin (80,000 gpm), 105-B Retention Basin (45,000 gpm), and the 190-C Area main sewer (143,000 gpm).

Present thinking and planning appears to be developing from the following factors as concern the 13.8 KV temporary construction power limit. 1. It is understood that the present intent is to supply 100-N operating requirements from a single stub source in the 230 KV loop. 2. The original thoughts were to obtain construction power over a 13.8 KV line from 151-D substation. 3. Construction load requirements are now less than originally planned since steam has been substituted for electrical drive of primary loop pumps and 5500 hp motor tests are no longer necessary. 4. An extreme emergency backup source for the K plants has always been of concern, although minimized in recent planning. It is desirable to review the temporary construction line requirements from a future operating viewpoint to determine if the line could be useful to the operating plants after completion of construction. It is highly desirable to provide T.C. power source from K plants rather than 151-D and then leave the line and breakers in place for future maintenance assistance and as extreme emergency backup to K plants.

This document is Revision No. 2 of the ``Process Specifications for Critical Mass Control -- Purex Plant``. It encompasses the changes in flowsheet, equipment and plant operating philosophy that have developed since the plant startup. It supersedes Document HW-44064 (Revision No. 1). Plutonium-239 and uranium-235, either as the metals or as dissolved or solid compounds are capable of spontaneous and violent nuclear reactions. Following initial dissolution of irradiated natural uranium fuel elements the Purex Plant is normally involved in the processing of liquid solutions containing these isotopes. However, in Purex Plant solutions, uranium-235 is invariably associated with a large fraction of neutron-absorbing uranium-238, a condition which precludes the possibility of a nuclear reaction. (These specifications are based on the premise that no enriched uranium is processed in the Purex Plant.) Plutonium-239, on the other hand, is separated from uranium by means of the Purex process; and even prior (process-wise) to its separation, the ratio of plutonium to uranium must be controlled during the off-standard conditions of plutonium rework or recycle. Critical mass control in the Purex Plant is concerned solely with prevention of nuclear reactions of plutonim-239, which has been produced by irradiation of natural uranium fuel elements.

This test will determine approximately the relationship between the power rate of rise indicated by the power rate meter and the actual pile power, rising period, and power rate. The actual pile power and power rate cannot be measured accurately during rapidly changing conditions; it is the intent of this test primarily to demonstrate that rate of rise protection offered by the power rate meter an be calculated to the same order of accuracy as the measurements of actual conditions.

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

Typical transient temperature curves for the reactor outlet coolant have been obtained for the cooldown period. Additional curves showing the relationship of several variables on the required heat exchanger area were obtained.

This document provides a discussion of activities at the Hanford Works during 1951.

The Technical Division was asked recently to consider designs for new reactors that would add 8000 MW capacity to the Savannah River Plant. One modification of the existing SRP design that would enable a higher power rating, and therefore require fewer new reactors, is an increase in the maximum pressure in the D{sub 2}O system. The existing reactors at SRP are designed for a maximum pressure in the gas plenum of only 5 psig. Higher pressures enable higher D{sub 2} temperatures and higher sheath temperatures without local boiling or burnout. The requirements in reactor cooling facilities at any given power level would therefore be reduced by pressurizing.

At the request of Hanford Operations Office, Atomic Energy Commission, a study was made of the effects on the Hanford Facilities of a navigation and power dam at Columbia River mile 348, about five miles upstream of the 300 Area. The original study was based on a nominal slack-water pool elevation of 395 to 400 feet at the dam location. A supplemental study evaluated the effects on plant facilities of a dam at the same location but with slack-water pool elevation of 385 feet. In addition to effects of the dam on Hanford, a study was performed to evaluate the effects the dam would have on the environment.

This document presents a summary of work and progress at the Hanford Engineer Works for December 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.

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.

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This report describes the second part of an IBM 704 Computer Code to determine transient thermal stress in pressure vessels. This part computes stresses from temperature distribution determined by the first part. A brief account is given of the solution of axially symmetric thermal stresses, using harmonic stress functions. This is followed by a description of the Code and complete input and operating instructions for a number of problems: a. Transient temperature distribution due to sudden change of temperature. b. Transient temperature distribution due to surface temperature changing as a function of time. c. Thermal stresses for temperature problems ( a ) or (b). d. Thermal stresses due to previously known temperature distribution. The code is known as TURF6. It was found that the computer results for thermal stresses in a solid sphere subjected to a sudden change of temperature at the surface corresponded within 5% of the theoretical values. The IBM 704 Computer required for this code has a 16 K memory, three drums and two tape units.

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The schedule for the installation of the PAR slurry loop experiment in the South Facility of the ORR has been reviewed and revised. The design, fabrications and Installation is approximately two weeks behind schedule at this time due to many factors; however, indications are that this time can be made up. Design is estimated to be 75% complete, fabrication 32% complete and installation 12% complete.

Most methods of separating isotopes depend on the use of a large magnetic field. The isotron is an electromagnetic device for separating isotopes, but it effects the separation by the use of radiofrequency voltages instead of magnetic fields. It has the advantage that plane sources of large area can be used instead of the slit sources to which most magnetic methods are limited. Before entering into any discussion of the details, a simplified description of the principles of the method is given.

A .0005 inch thick chemically nickel plated, C-64 aluminum clad, natural uranium, internally and externally cooled, Hanford production fuel element, which had incurred a ``side hot spot`` during irradiation, was shipped to the Radiometallurgy Laboratory in June 1959 for post irradiation examination. The slug vas irradiated to approximately 400 MWD/T in C Reactor as part of PT-IP-207-A. The examination was requested by personnel from Process Engineering, Fuels Preparation Department and Process and Reactor Development, Irradiation Preparation Department to determine the irradiation behavior cf nickel plated fuel elements and to aid in evaluating the nickel plated fuel element program. The presence of the nickel plate probably averted a ``side hot spot`` failure. Al-Si spheroidization and Ni-Al diffusion indicated that the maximum surface temperature vas 325--350{degrees}C for at least one hundred hours, however no sloughing of the nickel vas seen in the ``hot spot.`` Sloughing of the nickel plate, associated with poor nickel bonding, vas observed near the cap end of the slug. The condition my have been aggravated by the formation of hydrogen gas originating from the diffusion of atomic hydrogen through and/or from the nickel plate into the voids between the nickel and aluminum in poorly banded areas. Observations prompted …

To date, no data is available on the swelling rates of unrestrained unalloyed uranium rods operating at 450--630C core temperature and 250--350C surface temperature. These temperature ranges are of extreme interest in the Hanford fuel element development program. To obtain information on the effect of cladding restraint, exposure, and the above temperatures, on uranium swelling rates, a series of experiments is being designed for irradiation in Hanford reactors. The experimental assembly consists of a uranium fuel rod supported concentrically in an aluminum capsule. Surrounding the fuel rod is an annulus of liquid NaK to serve as a heat transfer agent. To provide a check on the operation and temperature calculations of the capsules, it is proposed initially to make a single irradiation in a normal process tube. The uranium core temperature will be continuously monitored by a thermocouple, and flux measurements will be made in the capsule to determine the specific power of the fuel.

This document presents a summary of work and progress at the Hanford Engineer Works for November 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.