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This report provides a compilation of memorandums and studies related to a Hanford production cut-back.

This report provides operational data for the KPR Reactor utilizing Alternate Tube Neptunium 237 targets.

This report itemizes the irradiations performed by Testing and Irradiation Services for Battelle-Northwest. It lists the material being irradiated, awaiting disposition and material shipped during the report period.

Using a Mechrolab Vapor Phase Osmometer and a Hallikainen Automatic Membrane Osmometer the number-average molecular weight of two samples of dimethylpolysiloxane - 2300 and 8000 cstk - as well as samples made by mixing the two previously mentioned materials were determined.

The Building Electrical Power System, transmits, distributes and controls the electrical power utilized by the equipment of all of the FFTF system. Included are system functions and design requirements, the physical description of the system, safety considerations, principles of operations, and maintenance principles.

A primary factor in assessing the impact on the biosphere of the discharge of heated reactor effluent to the Columbia River is the degree of initial dilution. Each of the reactor effluent lines discharges at the bottom of the river, approximately normal to the current, and near the main channel. As part of the Division of Production research and development program titled ``The Effects of Reactor Effluent on the Quality of Columbia River Water,`` the temperature measurements recorded in this document were taken over a period of several years. Generally, the exact locations of the ends of the effluent lines could not be determined from the river. The intent therefore was to determine the initial degree of dilution at the point where the mixed effluent first reached the surface, the so-called ``bubble,`` over a range of river flow-rates. A secondary goal was to determine the distance downstream (in quarter-mile increments) at which vertical mixing of the effluent streams could be considered complete. The data reported here have been partially reported previously and used in the development of a mathematical model of plume development. The purpose of this document is to provide a record of the field data in more detail …

The results are reported of a literature search for information concerning the heterogeneous, gas phase, catalytic hydrogen-oxygen recombination. Laboratory scale experiments to test the performance of specific metal oxide catalysts under conditions simulating the atmosphere within a nuclear reactor containment vessel following a loss-of-coolant blowdown accident are suggested.

Process development personnel have been seeking means of improving the (after thermal exposure) adhesion of the S8DR ceramic coating to chromized Hastelloy N tubing. This effort has led to the incorporation of an elevated temperature chromium diffusion treatment prior to the application of the ceramic coating. This report describes the results of tests performed to ascertain the influence of the additional thermal treatment on the 1400/sup 0/F creep-rupture properties of S8DR tubing.

This report itemizes the irradiations performed by Testing for Battelle-Northwest. It lists the material being irradiated, awaiting disposition and material shipped during the report period.

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The design of a Phase II dummy hydraulic core for the CSE reactor simulator is described. The core will be a scale model having features typical of PWR or BWR reactor cores, and will include control rods and associated support mechanisms, fuel elements, a thermal shield, top and bottom core guide and support plates, and flow baffles. The core and simulator vessel will be instrumented to provide measurement of temperature, static and differential pressures, forces on core components, and rarefaction wave pressures during high temperature, high pressure blowdown studies.

It is the purpose of this report to present data obtained from the routine reactor effluent radioactivity monitoring program as well as the results obtained from the research and development program directed toward the reduction of effluent radioactivity.

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This document, dated May 3, 1966, presents data from the zirconium tube sample evaluation program at the Hanford Reservation. Data presented includes: (1) operating history, (2) flow characteristics, (3) time in core, and (4) average temperature for numerous samples. The document consists entirely of data.

A knowledge of the fuel temperature transient following a reactor flow-loss accident is important in reactor confinement studies, since the rate an time of fission product release depend upon this transient. Fuel melting times presented by Ekern, Muraoka, and Zaloudek have been used in past confinement studies. These fuel melting times were based on the solid fuel element design, and the temperature transients were not presented. Furthermore, improved fission product decay heat data and reactor scram transients are now available. Therefore, the ALOC computer code was developed to calculate the available. Therefore, the ALOC computer code was developed to calculate the fuel temperature transient following an accidental loss of coolant. This report presents fuel temperature transients for assumed accidents as calculated by this code. Operating instructions for the ALOC code are given in the appendix.

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The meetings were held on December 6 and 7, 1965 in the 325 Building conference room at PNL. The objectives of the meetings were: (1) To review the status of the ceramic coproduct target element fabrication technology and the subsequent extraction processing technology of irradiated elements as developed at PNL. (2) To discuss the feasibility of performing extraction studies of irradiated coproduct target elements at the SRL pilot plant facilities in demonstrating the capability of coproduct production in N-Reactor. (3) To identify potential problem areas that could arise when the laboratory scale extraction process would be performed on a pilot plant scale in the existing SRL facility.

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