Construction and operation of a simplified ionization chamber.

Resolution of the products of the reactions of DPN and ATP with ultraviolet light by techniques of partition paper chromatography is reported in this paper.

Report discussing the relative thermal conductivities of liquid lithium, sodium, and eutectic NaK, and the specific heat of liquid lithium, as well as the methods and materials used to determine this information.

From introduction: "Five appendices, each of which pertains to a separate phase of the Filtration of Radioactive Aerosols by Glass Fibers program."

Use of Crout Method in floating-decimal notation in IBM Model I Card-Programmed Calculator.

From summary: "This report was prepared for the Advisory Committee on Reactor Safeguards for their review and consideration of the potential hazards concerning the Battelle Memorial Institute Research Reactor."

Processing of Florida leached zone material for the recovery of alumina.

Discussion of the economic considerations for atomic power in merchant ship propulsion.

Discussion of the economic considerations for atomic power in merchant ship propulsion.

Purex process which utilizes solvent extraction to separate and purify uranium and plutonium from each other and from fission products contained in irradiated uranium fuel elements.

More detailed study of an advanced engineering test reactor.

Quarterly technical report discussing progress and activities of the Liquid Metal Fuel Reactor Experiment.

Engineering study on underground construction of nuclear power reactors.

Abstract: An analysis and a summary are.given of the radioactivity buildup program at the Army Package Power Reactor during the last nine months. · Due to the wide fluctuation of water and crud results, only general interpretations can be made with this data. Metal test coupon data indicate a substantially greater buildup on Croloy 16-1 metal than on Type 304 stainless steel. Coupled with the decreased ability to remove the radioactivity buildup on Croloy 16-1 by conventional descaling techniques, the implication is that this metal might pose a serious problem for use as steam generator material. It was emphasized, however, that results to date ae only preliminary and extensive additional experimentation would be required to reach more definite conclusions.

Liquid Metal Fuel Reactor Experiment program quarterly technical report.

Introduction: Seldom does the actual arrangement of control elements in a nuclear reactor confers to the ideal and convenient mathematical array. In order to achieve shim control. safety and regulation, it is desirable to design with rods of different sizes and materials. With given fuel element arrangement, typically in square or hexagonal lattice spacing, there will be rods located at different distances form the center of the core and from each other. As the reactor operates, absorbers will be withdrawn, leaving further asymmetries in the location of those remaining. It is the purpose of this report to develop in detail a two-group diffusion theory with as complete generality as possible. The method is as yet restricted to the unreflected core, or to the reflected core by use of reflector savings and bare equivalent geometries.

Preliminary study to determine the technical and economic feasibility of a 315 eMW High-Temperature, Integral, Liquid Metal Fuel Reactor (HTIR).

Abstract: This technical report summarizes the data obtained in a recent literature survey conducted to determine the effects of neutron irradiation on the impact and other mechanical properties of both ferritic steels and austenitic stainless steels. The survey was primarily aimed at obtaining sufficient data on the behavior of pressure vessel steels at high integrated neutron flux levels in order that a reference material of construction could be selected for the SM-2 (APPR-1B) reactor vessel. Materials studied in this literature survey included carbon and low alloy steels such as: ASTM A-212B, ASTM A-201, ASTM A-301B (CR-Mo), ASTM A-106 (coarse and fine grained), ASTM A-285, ASTM A-302B (Mn-Mo), ASTM A-353, ASTM A-203 Grade D, E-7016 carbon steel weld metal, USS Carilloy T-1, HY-65 and HY-80. In addition, Types 304 and 347 stainless steels were also investigated as representative austenitic materials which might be used in pressure vessel construction. A careful evaluation was made of the irradiation induced changes in the mechanical properties of the above materials. The ferritic steels were evaluated primarily on the basis of increases in transition temperature due to irradiation and decreases in the amount of maximum energy absorbed prior to ductile failure. Factors such as industrial experience, …

From introduction: "The contract covers final design of the SM-2 Nuclear Power (PWR) Steam Electric Station, described in the preliminary design report, APAE 40, Volumes I and II."

Summary: Prototype testing of the safety and control rod drives indicated that both units functioned properly. No major problems were encountered during testing. Seal leakage data collected indicated that the seal units were performing satisfactorily. Scram times during both cold and hot testing were excellent and actually better than expected.

This report summarizes only that portion of the injection-casting experiments in which the castings were made with air pressure.

From abstract: "This report presents an IBM digital computer program that solves numerically and simultaneously the equations that describe gross reactor behavior under runaway conditions."

An engineering-scale demonstration of the Sulfex process indicated that semi-continuous decladding of unirradiated stainless steel-clad UO2 or U02ThO2 fuels is feasible.

Abstract: This technical report describes the changes incurred in the SM-1 by the insertion of the SM-1 Core II without special components. The SM-1 Core II components were made to specifications very nearly identical to those of SM-1 Core I. The differences consist of europium absorber sections, internal europium flux suppressors in the control rod fuel elements, and low impurity cladding. Each of the SM-1 Core II components with the exception of the five absorber sections new in SM-1 Core I were subjected to a Zero Power Experiment at the Alco Critical Facility. The results of this experiment indicate that the SM-1 Core II will have nuclear characteristics very similar to that of the SM-1 Core I. Since SM-1 Core II will be operated with the same mode of rod control, in the same core support structure, and with the same primary coolant flow conditions, the thermal characteristics should be essentially identical to that of SM-1 Core I. Also, all kinetic characteristics of SM-1 Core II should be identical to those of SM-1 Core I. This report demonstrates that there is no increase in potential for a hazardous situation at SM-1 due to the replacement of SM-1 Core I by …