Technical report presenting the results of a study of the AEC thorium production facilities at Iowa State College, Ames, Iowa. Operating procedures, descriptions of equipment and a production cost analysis have been included in this study.

This memorandum presents the process design for the solvent reast area of the Feed Materials Production Center. The purpose ofthe transmittal is to provide the basis for the detailed mechanical design and layout of this section of the refinery. Drawings, flowsheets, and specifications are included.

Annual report describing progress on unclassified research programs funded by the U.S. Atomic Energy Commission during the 1964-65 fiscal year.

Quarterly report describing progress on unclassified research programs funded by the U.S. Atomic Energy Commission during the first quarter of the 1966 fiscal year.

Annual report describing progress on unclassified research programs funded by the U.S. Atomic Energy Commission during the 1965-66 fiscal year.

The following progress report describes the examination of elements removed in comparison to a previous study on the power facility reactor operations.

The following progress report describes the examination of elements removed in comparison to a previous study on the power facility reactor operations.

A report regarding carnotite resources of Outlaw Mesa, Mesa County, Colorado.

"his interim report presents the grade and tonnage data collected during exploration at the Lost Creek schroeckingerite deposit, Sweetwater County, Wyoming, Between July 1951 and February 1952. Discussions of he geologic results of exploration (lithology, mineralogy, structure, origin, etc.) are not included in this report and consequently the detailed lithology and structure have been omitted from the illustrations (figs. 1-7)"

"This report describes second-cycle postirradiation examination and AIROX reprocessing-refabricating of uranium dioxide irradiated to an additional 10,000 Mwd/MTU burnup."

A report regarding the irradiation behavior of Unalloyed hypostoichiometric uranium carbide experiment AI 3-11 and review

"The purpose of this report is to examine the safety aspects of PNPF restart on continued operation, after completion of the core cleanup and system modifications."

The development of a high heat flux heater for use in corrosion rate and stress-rupture tests of fuel cladding materials in a sodium environment is described.

The following handbook was made for the purpose of: (1) to give a semidetailed description of the testing reactor, and (2) to explain, in so far as possible, the reasons for the design.

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.

The following report follows the study on economics of nuclear power generation using a light water reactor.

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.

Abstract: "The mechanical properties of a large number of zirconium alloys have been investigated. These alloys include zirconium-tin and zirconium-tin-uranium. Tensile tests have been run on these alloys at room temperature, 500 F, and 600 F. Hot-hardness data have been obtained from room temperature to 1600 F, with particular emphasis on the hot-rolling range. Creep tests and fatigue tests have been run on a number of zirconium-tin alloys at 500 F."

Abstract: "The mechanical properties of a large number of zirconium alloys have been investigated. These alloys include zirconium-tin and zirconium-tin-uranium. Tensile tests have been run on these alloys at room temperature, 500 F, and 600 F. Hot-hardness data have been obtained from room temperature to 1600 F, with particular emphasis on the hot-rolling range. Creep tests and fatigue tests have been run on a number of zirconium-tin alloys at 500 F."

Abstract: "The ductility of Type 310 stainless steel T-joints brazed with GE-62 brazing alloy was measured at room temperatures 1200, 1650, and 1800 F. The measure of ductility was taken as the plastic axial strain required to crack braze fillets in T-section tensile specimens. At elevated temperatures, the ductility of as-brazed joints approximated that of the stainless steel, but at room temperature the brazed joints had only one-tenth the ductility of the base metal. Annealing for 16 hr at 1800 F in air was found to triple the room-temperature ductility of the brazed joint."

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: "Aluminum-clad nickel-plated uranium is corrosion resistant in boiling water. Samples with intentional defects in the aluminum-nickel clodding layers, after testing for more than 300 hr. showed no sign of rapid failure. The study of electroplated metals on uranium for aluminum cladding has defined the effects of several processing variables on the corrosion resistance. The quality of the uranium and the quality of the aluminum-nickel bond were important. Heat treatment and vacuum outgassing of the nickel-plated uranium prior to aluminum cladding were beneficial, and minimizing the amount of air present in the hot-press assembly was desirable. Variation in the steps for electroplating on uranium did not affect the corrosion resistance of aluminum-clad uranium."

This report covers the progress made on an intensive program to develop and evaluate UO2 as a possible fuel element for the PWR.

In studies of the fabrication of fuel subassemblies for the supercritical-water reactor, the limited ductility of stainless steel-UO₂ fuel sheet made the fabrication of corrugated-flat-plate-type subassemblies impractical. A fabricable MTR-type assembly, 5/8 in. square and 30 in. long, was developed. Methods for edge cladding fuel sheet were developed and a brazing alloy, GE-75, was found to be corrosion resistant in supercritical water.