Interim report regarding an exploration of the Jo Dandy area, Montrose County, Colorado. Concerns work done on the behalf of the Division of Raw Materials of the U.S. Atomic Energy Commission.

This report deals with a group of both theoretical and experimental investigations which have been carried out, utilizing Core 15, one in a series of critical assemblies that have been constructed at Atomics International's Epithermal Critical Experiments Laboratory (ECEL).

Abstract; This technical report describes the changes in the SM-1 incurred by the experiment, Core Temperature and Flow Instrumentation (Task XIV), and evaluates the possible hazard involved in these changes. Temperature and flow measurements will be taken on a Task XIV instrumented stationary fuel element, instrumented control rod fuel element and other selected points in the SM-1 core to provide data on the core steady state and transient performance. The hazards evaluation consists of a nuclear evaluation, thermal and hydraulic analysis, description of tests to be performed, and discussion of containment integrity and maximum accident considerations.

Abstract: This technical report contains a summary of experimental and analytical work performed on cold clean SM-1 Core II with special components and PM-2A Core I each with thermal shield and vessel mockups. The purpose of this work was to define the neutron intensity and spectra to which the pressure vessels are subjected in order to assess the effect of irradiation on the vessel lifetimes. The radial distributions of the energy dependent neutron flux per unit power in the core and vessel mockups are presented. The uncertainties associated with the measurements are also given. In addition to graphical and tabular presentations of the test results, discussions are provided on the experimental techniques and theory. A review of the analytical models is provided in addition to the calculational results. A special analytical study of the the PM-2A Core I power distribution and absolute power level determination is given. A comparison of the experimental and analytical results is made and conclusions and recommendations presented.

Abstract: This technical report covers the zero power experiments performed on PM-2A Core II at the Alco Critical Facility. PM-2A Core II is the first replacement core for a portable pressurized reactor at Camp Century, Greenland. Core II is the same as Core I with the exception that Core II has an increased burnable poison (B-10) content. The zero power experiment consisted of fuel element uniformity test; core assembly test, development of an on-site loading procedure and an analysis of experimental data. Physical characteristics determined include distribution of fuel and B-10 in the fuel plates, minimum critical mass, control rod bank calibration, and integral rod worth. The report concludes with an analysis of the experimental data including estimated uniform and non-uniform burnup rates.

This quarterly progress report details the ongoing research and experiments at the Oak Ridge National Laboratory as part of the Aircraft Nuclear Propulsion Project. The first part of this report discusses reactor theory and design. The second part of this report is not included. The third part of this report discusses materials research. The fourth part of this report includes appendixes

Abstract: This technical report contains a comprehensive analysis of the nuclear characteristics of the SM-1 Core I. Comparison of analytical and experimental results for neutron ages and core reactivities of a variety of cases investigated shows that the MUFT III with P-1 slowing down approximation gives the best results. At startup the core reactivity and rod bank worth under various operating conditions are investigated and compared to experiment. Core lifetime was calculated to be 16.8 MWYR compared to 16.4 MWYR experimental. The temperature coefficient has been calculated and compared to experiment as function of burnup. In Appendix A, flux distribution, temperature coefficient, effective delayed neutron fraction and core life are analyzed by Dr. R. L. Murray by one and two group modified theory series expansion calculations.

The purpose of this manual is to help an engineer or scientist to design a practical shield by making available to them the techniques and data developed by the Naval Reactors Program and Pressurized Water Reactor Program.

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.

From introduction: "This investigation was undertaken with the general objective of determining whether uranium core could be sheathed with aluminum by an extrusion-cladding process. The ultimate objective of this work was to determine the feasibility of extrusion cladding flat plates of uranium."

Abstract: "The possibility of transient temperature variations during the self-heating of a plasma by thermonuclear reactions was investigated. Calculations were performed on an IBM Card Program Calculator. Estimates of the time scale for self-heating were obtained. It was concluded that it would be unsafe to assume that the nuclear temperature is always equal to or greater than the electron temperature in a discussion of plasma instabilities for a plasma with 10 percent or greater tritium concentration."

Abstract: "Two parallel wires close together with an arc between them form a hairpin-type of current configuration which should magnetically propel the plasma of the arc. By the time-of-flight measurements with probes it has been demonstrated that directed translational energies of 290 ev can be imparted to deuterons in a plasma produced by a high-current pulsed arc at the tips of two titanium wires loaded with deuterium. It is believed that three deuterons are actually magnetically propelled."

Abstract: "The deuterium plasma which is ejected by a 6000-ampere pulsed current through a rail-type spark source has a yield in the forward direction which is a factor of 5 to 10 greater than that in the perpendicular direction. This effect is believed to be due to the magnetic propulsion of the plasma."

Introduction: This report describes a program to develop a nondestructive (ultrasonic) tester and testing procedure to perform integrity tests on isotope heat source containers. The increased application of isotope heat sources has brought more attention to the engineering problems encountered during fabrication. One of the most important of these problems is insuring the isotope containment during the long unattended periods of use. To determine the integrity of the heat source container, a nondestructive testing procedure must be adopted to test each heat source after fabrication.

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 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.

From foreword: This report presents the preliminary design of a gamma irradiator intended for use as an experimental hospital sterilizer.

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.

Report describing technical progress on the design and construction of an Organic Moderated Reactor Experiment (OMRE), to be operated at the National Reactor Testing Station in Arco, Idaho. "This is the first report of the series and coves the period from the initiation of the project to July 31, 1956. Also included as an appendix to the report is a detailed description of the OMRE facility" (p. 3).

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: Impact tests on the Mark III Hanford safety element indicate that the design is mechanically adequate and does not constitute a personnel hazard.

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

From introduction: This report details the results of the post-irradiation examination of a test involving one defected rod and one undefected rod.