This report describes the loan proposals of Messrs. Kingman Brewster, Jr., Charles Killingsworth, Edward Shapiro, and William Vickrey and the more recent proposals of the panel on Educational Innovation. These proposals would permit a student to draw an advance at least equal to the cost of his education without imposing a financial burden on the taxpayer or the federal government.

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In connection with the ORR Hastelloy-N irradiation experiments, a limited number of tensile tests and uniaxial and biaxial stress-rupture tests on S8DR Hastelloy-N heats were conducted at AI to determine the effect of some of the ORNL test conditions. The tests performed at AI were in parallel to the ORNL control tests. The results showed that the effect of the test condition variations between the two tests were generally insignificant. The effect of the test conditions is discussed.

Dosimetry measurements and calibrations have been made at the University of Washington's Fast Neutron Medical Research Facility. The facility is designed to deliver localized neutron doses of a few rads to the testes of volunteer subjects at 2.5 MeV and 14 MeV. Dose measurements were based on the use of tissue equivalent ion chambers. Dosimetry and shielding measurements show that doses accurate to +-10% can be delivered to the volume of interest while maintaining safe dose levels outside the shield and to other portions of the subject. 10 references, 9 figures, 3 tables.

Thermophysical-property specimens of ThO/sub 2/, ThO/sub 2/-10 wt % UO/sub 2/, and ThO/sub 2/-20 wt % UO/sub 2/ were fabricated and characterized. Each material composition was represented by specimens with nominal densities of 95, 85, and 75 percent of theoretical. Thermal-diffusivity, linear-thermal-expansion, and specific-heat measurements were made at 300 to 1900/sup 0/C and used to determine thermal conductivity. Results of linear-thermal-expansion and specific-heat measurements were approximately the same for all materials. Values of linear thermal expansion ranged up to 2.2 percent in heating from room temperature to 2000/sup 0/C. The specific heat ranged from 0.055 cal/(g) (C) at 0/sup 0/C to 0.080 cal/(g) (C) at 2000/sup 0/C. Thermal-diffusivity values ranged from 3.0 x 10/sup -2/ cm/sup 2/ per sec at 300/sup 0/C to 0.5 x 10/sup -2/ cm/sup 2/ per sec at 1900/sup 0/C. Thermal conductivities decreased in regular fashion with increased UO/sub 2/ content and porosity. An equation is given which fits all of the data obtained; it is anticipated that it will fit the data for other compositions within the given limits. (NSA 22: 47703)

With the exception of N Reactor the plutonium production reactors operated by Douglas United Nuclear, Inc., use treated Columbia River water as coolant on a once through basis. Thus, radionuclides formed by neutron activation of Columbia River salts not removed in the water treatment process and water treatment additives are discharged to the river. Although the quantity and possible effects of the radionuclides released are well within nationally accepted limits, emphasis has been placed for some time on reducing the releases to as low a level as possible. More recently increasing concern has been evidenced with regard to the heat which is also discharged to the river. This report discusses concept which not only would drastically reduce the radionuclide content of the river but which would also substantially decrease the heat discharge. This concept is the disposal of the reactor effluent to the ground either to a pond or to a network of trenches.

This report provides data on daily dissolver operation and metal transfers. Sodium hydroxide and sodium nitrate requirements are also given.

Interchannel flow through fuel element junctions can occur in production reactor fuel columns as a result of different pressure loss patterns in the two coolant subchannels. This interchannel flow is most significant during boiling conditions because, generally, the boiling length in each subchannel is different. These different boiling lengths result in divergent pressure loss characteristics in the subchannels and cause differential pressures to exist across the fuel element junctions. These differential pressures are the predominant driving forces for interchannel flow. The extent to which interchannel flow exists, and its effect on the thermal hydraulic characteristics of reactor flow channels has long been a question of concern in the design of electrically heated models of reactor fuel columns for laboratory experiments. Interchannel flow is quite unpredictable and subject to considerable variation due to mechanical and alignment changes at the junction. It is also very important because it significantly affects the total thermal-hydraulic performance of the fuel column. The purpose of this investigation was to analytically evaluate the effects of interchannel flow on the predicted thermal-hydraulic performance of a production reactor fuel channel and to develop accurate methods by which it could be simulated in electrically heated laboratory models.

It is of interest to investigate the time following a complete loss of coolant incident when the pile would be exposed to the ambient air. This is of concern due to the fast propagation of graphite burn-up when brought in contact with air at or above the operating temperature of the graphite (1292{degree}F). One possible way conjectured in which the pile can be exposed to the surrounding environment is for the centering flange gasket (neoprene) or the bellows gas seal (silicone) to fail. This particular study was concerned with the temperature transient of the seals following the extreme accident condition of complete loss of coolant. This problem is one of many being investigated in a concentrated effort to re-evaluate the effectiveness of the confinement system of the smaller reactors. The maximum credible accident for the DUN operated production reactors has been the loss of coolant to a single cross header. The Hanford philosophy of reactor safety has been the provision of reliable methods of cooling fuel in all accident situations short of severance of the primary coolant piping or destruction of the reactor building. However, for the present study the more serious accident of complete loss of coolant was investigated. …

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This report details activities of the Division of Production and Hanford Plant Assistance Programs for the month of October 1967.

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The purpose of this report is to document the evaluation and capabilities of the Core Radial Displacement Transducer. Pertinent details of the transducer history (Measurement Definition, Basic Design Evaluation Program) are included to substantiate the evolution.

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