From abstract: A program to investigate the mechanism of the critical heat flux condition from the standpoint of flow regimes has been initiated at Dynatech for the AEC. This report covers the work done on this investigation in the first year.

Technical report presenting preliminary results obtained in the static corrosion test program to screen promising construction materials for a proposed liquid fuel power breeder reactor.

Abstract: Irradiated fuel elements from the Organic Moderated Reactor Experiment (OMRE) first core loading have been examined and evaluated to determine: (1) the stability of the floating plate fuel element design, (2) the stability of the stainless steel clad UO2 - stainless steel cermet core fuel plates under irradiation and exposure to the organic coolant, (3) the extent and nature of deposits on the fuel element services, and (4) the distribution of burnup in the fuel elements.

Introduction: The development and qualification of the system acceptance test heaters and the reactor simulator heater are described in this progress report.

From abstract: The response of the reactor to forced oscillations has been investigated with the pile oscillator technique and the data are presently being analyzed.

From summary: The full 140 element loading of the core was completed on October 10, 1962. At this point, critical operation was begun for operator training and post-critical testing purposes.

Abstract: This report presents an analysis of coolant boiling in sodium graphite reactors.

Introduction: Measurements of polonium content in solids are usually performed by dissolving the material and preparing a thin sample which can be measured with one of the available instruments. A measurement of the radiation emerging from the thick sample should be sufficient to evaluate the polonium content. For this reason a calculation has been performed and the results have been compared with those obtained by the thin sample method. Obviously this proposed method using thick samples would be faster.

This report summarizes the results and conclusions of a study made to evaluate the merits of using zirconium hydride as a solid moderator in an integral boiling water-nuclear superheat reactor of the pressure vessel type.

Report discussing two trial runs for a grannular resistance processing furnace, in which the goal was to produce high purity graphite. These tests were a continuation of a series of trial runs that had previously yielded inconsistent results.

Abstract: "A phase diagram for the uranium-cobalt system has been constructed, based on thermal, microscopic, X-ray and chemical data. The results indicate the existence of three compounds and two eutectics, one of which shows a lowering of the melting point of uranium by approximately 400 C."

Technical report describing that voids were measured in a ½-inch by 1-3/4-inch channel with the S-1 insert (B(0)/B(1) = 0.4, L(0) = 0.1 inch), at 2 inches ahead of the insert (position A), ½-inch past the insert (position B), 5 inches past (position C), and 12 inches past (position D). The conditions were: P – 1000 psia, G = 1.00 x 10(6) lb/h-ft(2), and x = 18.8 percent. Average void and void distribution at position A are the same as for flow in a straight channel. Void distribution at position B shows that the stagnation region downstream of the inserts contains a high fraction of voids. Average void and void distribution at positions C and d show that the two-phase mixture becomes strongly mixed (homogenized) as a result of passing through he contraction-expansion inserts. Distribution at position D approaches the distribution at position A; i.e., the straight channel distribution.

A report about the fabrication of dummy fuel-pls assemblies which are to be used in the engineering test loop at the MTR.

A report about the design and construction of equipment for thermal-conductivity measurements of irradiated encapsulated uranium. Metallographic techniques used to confirm solubility limits of uranium in thorium, previously determined by electrical resistivity and X-ray data, have been unsuccessful.

A report about aluminum-clad fuel elements, fuel-element development, zirconium uranium alloys, corrosion of zirconium, and reactor material development.

Abstract; The heat exchanger and steam generator for the U.S. Atomic Energy Commission Sodium Components Project will be constructed entirely of type 316 stainless steel. Because of the susceptibility of this alloy to stress corrosion cracking, it is proposed to clad all areas of the steam generator with Inconel where the stainless steel will be exposed to water and steam. This report includes a discussion of the work by numerous investigators that justify the selection of Inconel for this service. A discussion of Inconel type welding alloys is also included.

The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.

Technical report covering chemistry of source, fissionable, and structural elements, nuclear chemistry, radio-organic chemistry, chemistry of separations processes, chemical physics, radiation chemistry, and instrumentation at the Oak Ridge National Laboratory for the period ending September 30, 1950. [From Abstract]

Quarterly report describing progress on the EVESR Nuclear Superheat Fuel Development Project.

Report documenting the Halogen Collector Test program, which was undertaken to provide "methods of removing trace quantities of radioactive iodine from an air stream" (p. 1). The report includes designs of the program and the test facility, test results, analysis of these results, and encountered problems. Appendices begin on page 38.

From introduction: "This report provides background information regarding the study results presented in the first volume of this report and provides information resulting from the studies made on the gas cooled reactor concept."

Study of factors, such as the average particle size and particle-size distribution, which affect the fluidization properties of uranium trioxide powders.

"We will consider two finite difference schemes for the solution of the Navier-Stokes equations in one space dimension. The first is a method used by Lax and Wendroff. The second method is based on an approximate solution of the centered, implicit difference equations. Since this approximate solution is obtained by successive substitutions, we will refer to this second method as an iterative method."

"The energy method is applied to study the stability of two types of difference approximations to parabolic partial differential equations, the alternating direction methods Douglas, Peaceman, and Rachford, and a new semi- explicit method. Each difference scheme is proved to be unconditionally stable. These results apply to parabolic equations with variable coefficients, defined in cylindrical domains with an essentially arbitrary bounded base."