From Introduction: "The principal objective of this investigation is to promote fundamental understanding of the mechanism of solidification of aqueous solutions, and the effects of externally applied electrical and magnetic fields on solidification."

Abstract: A materials exposure program has been established in the Garigliano Nuclear Power Plant to measure the effect of neutron irradiation and time at temperate on the mechanical properties of the reactor pressure vessel steel. Base metal specimens were made from portions of the pressure vessel steel, and weld heat-affected zone and weld metal samples were taken from a weldment made from the pressure vessel steel and simulating a pressure vessel circumferential weld since there are no longitudinal welds in the forged ring shell. The specimens were sealed in helium-filled capsules and placed in the reactor vessel at locations where they will be exposed to a variety of conditions. Tensile property changes will be measured by pre- and post-irradiation tests on small tensile specimens. Fracture characteristic changes will be measured in similar fashion by Charpy V-notch impact tests. The program is planned to cover a 32-year period, with specimens to be removed for test at intervals of 1, 2, 4, 8, 16, and 32 years.

From foreword: The results of an engineering study directed toward the evaluation and selection of a nuclear reactor system designed to produce electricity and furnish steam to a water desalination plant in the Florida Keys are given in this report.

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.

From Abstract: "An electronic technique was used in determinations of transference numbers in aqueous electrolyte solutions by the moving-boundary method. Data for 0.1 N NaCl at temperature from 15 degrees to 125 degrees C are presented in this report. Probable causes of difficulty with the method were identified. and recommendations are made in this report for further research to overcome the difficulties."

Abstract: This report constitutes the RFD-1 Safety Flight Plan for investigating the disassembly, distraction, and disposal of the SNAP-10A reactor during re-entry.

From Introduction: "The research work described in this report consisted of two phases: analytical and experimental. Because of the complexity of the problem of the interaction between acoustical vibrations and heat transfer, the effort to analyze this problem theoretically began with a search for simplified "mathematical models" of the actual physical processes involved."

From abstract: "The objective of this program is to produce a set of pulse-height spectra representing the response of a standard gamma-ray scintillation spectrometer to radiations emitted in the decay of neutron-deficient nuclides."

Summary: Data were obtained in a two-road test section which consisted of two 7/16-inch diameter heater rods inside a roughly rectangular flow area. The heated length of the rods was 30 inches, with a 15-inch unheated calming length preceding it. Heater wall temperatures were recorded while the heater tubes were trans-versing the critical heat flux and transition boiling; these temperatures were used to calculate heat transfer coefficients. The following general results were obtaining: (a) Pressure has very little effect on the heat transfer coefficient in transition an film boiling. (b) Heat transfer coefficients during film boiling increase with mass velocity and steam quality. (c) The range of film boiling convective heat transfer coefficients observed was 364 to 1150 Btu/h-ft(2)-degrees F. (d) Temperature oscillations occur during transition boiling with a magnitude of as much 700 degrees F, at a frequency of about 1/2 cps. These temperature oscillations are reduced in magnitude as the steam quality and mass velocity are increased, becoming small (~20 degrees F) at high qualities and mass velocity. (e) A preliminary correlation of heat transfer coefficient data correlates the experimental data within about 20 percent. (f) Temperatures rises during transition boiling can be described analytically.

Report discussing the "problem of obtaining a curve of the heat flux vs. temperature difference for boiling liquid metals" (p. 3). Contains discussions of a variety of heating techniques, though it focuses on electron-bombardment heating.

"This report covers fire tests conducted on an experimental design of a wooden-jacket shield for radioisotope shipping containers" (p. i).

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.

Report that investigates the concept of using ammonium fluoride-oxidant systems to dissolve spent U-Zr-Sn reactor fuels in both batch and continuous operation.

Report evaluating the design characteristics and limitations of mobile blankets for breeder reactors. This also includes economic considerations for each tested blanket. Appendices begin on page 40.

From Introduction: "The purpose of this study is to investigate some of the hydrodyamic aspects of two-phase heat transfer in the lower quality range. The results of this study should contribute to a better and more thorough understanding of the processes and mechanisms affecting two-phase heat transfer, and in this way help in improving its prediction."

Report containing an aeroradioactivity survey of Puerto Rico made on behalf of the Division of Biology and Medicine, United States Atomic Energy Commission. Topics include the types of bedrock noted in the area as well as their levels of radioactivity.

Map with graded color shading to show levels of naturally-occurring gamma aeroradioactivity within Puerto Rico. Details about the survey and map creation are printed at right. Scale 1:250,000.

Abstract: This report is a synthesis of published and unpublished data on the rocks of the Anadarko basin.

Report containing the summation of the desalination studies carried out by The University of Oklahoma Research Institute, including their theories of carbon demineralization, electrode development, and the cost estimates for a demineralization plant.

Report containing information regarding the demineralization of water via carbon electrodes. Includes steps taken in the experiment, advancements made in the types of carbon material used, and cost estimates.

Technical report describing a UO2-Mo cermet fuel assembly fabricated for long-term irradiation performance testing in the Vallecitos Boiling water Reactor. The design and fabrication histories of this assembly are described and pre-irradiation data on each individual rod are presented. Molybdenum was added to improve the bulk thermal conductivity of the fuel, so that fuel temperatures would remain comparatively low during high-power level operation of the fuel element. The molybdenum was incorporated into the compacts either as fibers or as a thin coating on individual UO2 particles. Fuel pellets were produced from these materials by vacuum hot pressing. The distribution of the molybdenum in both types of cermet fuels appeared favorable to good heat transfer. The fibers were oriented predominantly in the radial planes of the pellet as a result of the uni-directional compaction during the hot-pressing operation. In the pellets made from the coated particles, a continuous network of molybdenum occurred as a result of the coating welding together during the hot-pressing operation. The test assembly contains eight fuel rods; three contain UO2-Mo cermet, three contain the cermet produced from the coated particles, and two are for reference and contain the conventional sintered UO2 pellet fuel. The nominal outside diameter …

The 300 series stainless steels had been selected originally as the reference fuel cladding material for utilization in several superheat reactor (SHR) systems being built as part of the United States Atomic Energy Commission (USAEC) program. The adequacy of the general corrosion resistance of Type-304 stainless steel for superheat fuel cladding was confirmed in the Phase I portion of the study reported previously. Fuel jacket failures that occurred in Type-304 stainless clad fuel elements exposed in the Vallecitos boiling water reactor superheated steam loop indicated the questionable dependability of such stainless steels for this SHX fuel cladding application. The following conclusions are based on the out-of-pile general corrosion evaluations completed to date on Incoloy-800 as a fuel cladding for nuclear superheat applications: 1. The corrosion data from 4000-hour heat transfer tests indicate good corrosion resistance up to at least a 1300°F metal temperature. By use of a devised method of data treatment, the general corrosion for three years exposure at 1300°F can be calculated to average 0.0016 inch with an upper 95 percent confidence limit of 0.0033 inch. 2. A compositionally-disturbed layer develops at the metal-oxide interface. The disturbed layer depth is a function of time and temperature of exposure. …

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

From abstract: "Fuel materials proposed for use in the Advanced Test Reactor were evaluated."