From Foreword: "Report containing information about ongoing research and development taking place in the Oak Ridge National Laboratory's Fuels and Materials Development Program. This includes a series of test reports covering such subjects as reactor fuels, the development of nitride fuels, and the effects of radiation on structural materials."

Report containing ongoing research and experiments of the Oak Ridge National Laboratory's Instrumentation and Controls Division.

From introduction: "Evaluation of trade-off and optimization studies in which numerous design parameters and economic ground-rules are varied to determine the effect on the economics of the reference concept. Selected safety studies."

From abstract: "explain all important features pertaining to a new pulsed irradiation reactor, the Capsule Driver, Core, and to analyze the potential problems and hazards of operating this reactor in the existing Spert IV facility."

The experiments described in this report were performed for two purposes: (1) to find a chemical, or combination of chemicals, that would be effective for inhibiting corrosion of steel in static portions of the Hanford water systems, and (2) as screening tests to select candidate mixtures for further testing to replace sodium dichromate as the inhibitor of aluminum and steel corrosion in the Hanford single-pass reactor cooling water.

From introduction: "Summary report of the 1000 MWe Boiling Water Reactor Plant Feasibility Study performed by the General Electric Company."

From introduction: "Presents the detailed description of a 1000 MW Electric Power Plant employing one or two Boiling Water Reactors as the steam source."

From introduction: "Contains the appendices and a complete set of drawings related to the 1000 MWe Boiling Water Reactor Plant Feasibility Study performed by the General Electric Company for the United States Atomic Energy Commission."

From abstract: "Verify rod mechanism characteristics reported by the United Shoe Machinery Corporation, demonstrate that rod mechanism characteristics were compatible with reactor kinetics and that over-all reactor system behavior was stable, and to recommend system modifications needed for satisfactory performance."

Chemical engineering analysis and stress analysis for design of the 3 megawatt intermediate heat exchanger and steam generator for service with a liquid sodium heat transfer fluid.

Abstract: The natural circulation, direct cycle, boiling water reactor reference design presented in this technical report is the alternate to the preferred preliminary design developed under Phase I of the PL-3 contract. The report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and training program outline.

Abstract: Test procedures for special tests involving in-core SM-1 temperature and flow instrumentation are described (Task XIV Package Tests). These tests involve in-core steady state flow and temperature measurements, loss of flow transients, load transients, reduced primary system pressure operations and reduced element flow. The thermal and hydraulic conditions prevailing in these tests, including steady state and transient burnout rations, are developed. The effects of reduced system pressure and flow on the burnout ratios are determined as are the expected stuck rod conditions when Task XIV test elements are installed. The effect on the maximum credible accident is included and a recommendation to conduct these Task XIV package tests is made.

Abstract: The pressurized water reactor preliminary design presented in this volume is the preferred design developed under Phase I of the PL-3 contact. This technical report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and a training program outline.

"This report presents the conceptual design of a 75 MWe prototype Mixed Spectrum Superheater power plant. The scope of the work has emphasized primarily the design, performance, and cost information on the nuclear portion of the plant. The research and development programs required to insure plant feasibility are also present."--Intro.

Abstract: This technical report summarizes all core physics experiments performed on SM-1 Core I and SM-2 Rearranged and Spiked Core I throughout core life. These measurements were obtained on site during the 16.4 MWYR lifetime of SM-1 Core I and 1.6 MWYR lifetime of SM-1 Rearranged and Spiked Core I. SM-1 Core I was the first stainless steel - UO2 dispersion core to burn out in the Army Nuclear Power Program. Experimental techniques are described and a complete history of fuel movements in the core presented. Measurements include control rod bank positions during all core conditions, temperature and pressure coefficients, control rod calibrations, transient poison effects, source multiplication, and stuck rod positions. The effect of core rearrangement and spiking on core reactivity and core life is also reported. Applicable measurements from the SM-1 zero power experiments are included.

Abstract: This technical report evaluates hazards involved with SM- 1 plant response and system performance tests (Task XII), to determine the response of the SM-1 plant to step electrical and steam load changes. The report describes the changes in plant equipment and operating procedures for this task and evaluates these changes for possible additional hazards to those described in APAE No. 2, Revision 1, "Hazards Summary Report for the Army Package Power Reactor SM-1."

Report discussing the Plutonium Recycle Critical Facility (PRCF). The reactor, building, associated equipment, operating program, safety, and operating procedures are described.

From introduction: "Describes progress in the development of radial vane primary separators following initial work reported in GEAP 3564."

Experiment to study temperature conditions during a five-minute heating cycle in the annealing of uranium sheets used as fissile material.

Abstract; The fission product data obtained during SM-1 Core I operation (June 1957 - May 1960) is reviewed briefly and interpreted. Evidence is presented to indicate that a fuel element defect was responsible for the high fission product activity level observed in the primary coolant. Relative escape coefficients are calculated and the defect size estimated. Anticipated fission product levels during SM-1 Core II and SM-1A Core I operation are estimated from alpha surface contamination data on completed fuel elements. The importance of in-line sampling for monitoring fission product activity is stressed as well as the need for failed fuel element detection methods.

Introduction: This is a final report on the evaluation of Kanigen electroless nickel plating for surfaces in contact with water and steam i a sodium heated AISI Type 316 stainless steel steam generator. The purpose of the coasting was to afford protection from stress corrosion cracking originating on the water-steam side of the unit. It has been concluded that the kanigen coating does not afford adequate protection for the services condition intended. This work was performed as part of the research and development program for the United States Atomic Energy Commission sodium Components Design Project.

Report containing ongoing research and experiments of the Oak Ridge National Laboratory's Instrumentation and Controls Division.

Report discussing the evaluation of the "behavior of austenitic and ferritic stainless steels, iron-chromium-aluminum base alloys and nickel-base superalloys, in medium and high temperature simulated reactor environments" (p. 2).

Chain reacting conditions for plutonium nitrate in water solution have been examined experimentally for a variety of sizes of spheres and cylinders.