Abstract: This technical report elucidates principles of hazards evaluation. The concept of hazards potential is introduced and utilized to show how a reactor system perturbation will influence its nuclear safety. Literature relating to reactor safety is referenced to provide the sources of information required for hazards analysis and show how they influence a hazards evaluation. A checklist of items which should be considered in evaluating a change, test, or modification is presented.

First volume of a compilation of information about reactor containment, written by experts from a variety of organizations.

Second volume of a compilation of information about reactor containment, written by experts from a variety of organizations.

Critical review of the information on zirconium, with estimates of the uncertainty in the values.

Symposium proceedings regarding long-term storage of highly radioactive wastes resulting from nuclear reactors.

From introduction: "The purpose of this report is to describe the observations made during the post-irradiation examination of HPD-2S, and to discuss possible modes of failure.

From summary: "An industrial piping failure survey covering 701 contacts in electric utilities, petroleum refineries, chemical processing, marine applications, architect-engineers, component manufacturers, piping fabricating and erectors, insurance companies, and others was conducted as part of the Reactor Primary Coolant Rupture Study."

From abstract: "This report presents the design, fabrication, assembly, operating procedures, and start-up data for six experimental test loops to examine the effect of steel exposed to sodium at temperatures as high as 1300 F."

Report documenting the ongoing research of the Oak Ridge National Laboratory's Chemical Technology Division. This report includes tables, diagrams, graphs, and articles related to chemical technology.

Report containing ongoing projects and experiments undertaken by the Oak Ridge National Laboratory's Molten-Salt Reactor Program.

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 abstract: "This report has been prepared as a design reference document describing the calculation methods and engineering results for shielding analysis work done during the design of the Advanced Test Reactor."

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."

Abstract: A survey of the problem of reactor system contamination by radioactive material and methods that have been employed to remove the material was carried out. Following this survey, an investigation of chemical solutions was undertaken to find one which might be successfully employed in the decontamination of a stainless steel steam generator. From a preliminary screening, the most promising chemical method from the view point of minimum corrosion and maximum decontamination is a caustic permanganate treatment followed by and acid rinse.

Abstract: This technical report summarizes the results of research and development tasks performed during Phase I of this contract and presents recommendations for future development work. Work is reported in the areas of plant assembly and relocation, housings and footings, waste heat dissipation, instrumentation, refueling systems, waste disposal, shielding, core nuclear thermal and hydraulic studies, gaseous waste processing (BWR), and critical experiments on a 5 x 5 array of Type 3 fuel elements.

Metabolic properties of plutonium and allied materials, biological studies of radiation effects, and biological effects of radiation from external and internal sources.

Annual progress report of safety program for several laboratory divisions.

Progress report discussing projects and work completed by the University of Michigan Department of Chemistry 1957-1958.

Investigation conducted to determine the cause of cracking, during tension, on the surface of tensile specimens of uranium-molybdenum alloys.

Analysis of a class of steady magnetohydrodynamic channel flow problems when natural, as well as forced, convection is important.

Derivation and basic formulas the physicist or nuclear engineer needs to do the necessary nuclear calculation for a reactor design.

The purpose of this study was to provide a means for evaluating the various parameters involved in the design and performance of levitation coils.

A study on radium and mesothorium poisoning and dosimetry and instrumentation techniques in applied radioactivity.

Abstract: A cyclic stress analysis of the SM-1 primary system was carried out. Problems encountered in the fabrication of PM-2A Core II and SM-lA Core II are described, and the results of an examination of damaged SM-lA Core I stationary fuel elements reported. A preliminary study of the radiation damage to SM-1 reactor vessel was made and the possibility of annealing the vessel discussed. Performance analyses are presented for five cores: SM-1 Core, SM-1 Core 1 rearranged and spiked, SM-1 Core II with special components, PM-2A Core 1, and SM- 1A Core 1. Preliminary critical experiments were made with SM-2 elements in a SM- 1 core configuration and nuclear and thermal analyses of the use of SM-2 elements in SM-1, SM-1A, and PM-2A completed. A throttling steam calorimeter was selected for measuring moisture carry-over on the PM-2A steam generator. Test procedures for evaluating the shielding of the SM-1, SM-lA, and PM-2A plants are summarized. Radiochemical and chemical analyses of SM-1 coolant and crud are summarized, and methods of activity control discussed. Preliminary results of studies of the properties of reactor pressure vessels under irradiation and no irradiation conditions are summarized briefly.