Experimental investigations were conducted to obtain additional information on the growth and characteristics of sodium oxide deposits in liquid sodium which could lead to system plugging. These investigations included the removal of sodium oxide from molten sodium by the cold trap method.

VARI solves on the IBM-650 a system of simultaneous, first-order, ordinary, differential equations. The program was written so that a large number of calculations could be done in a reasonable length of time. The program permits the consideration of the production of the isotope by absorption and/or decay of one or more of any of the other isotopes in the chain.

A program of free-surface steam-water separation at 600 psig was carried out in support of the design for operation of the Experimental Boiling Water Reactor at 100 MW thermal. Reduced test data are presented without evaluation since the ANL technical representative received an original copy of the data for evaluation with respect to the EBWR design.

Experimental results on the effect of periodic acceleration on a natural circulation, 1000 psia, two-phase flow loop are presented, and related to the topic of marine reactor design. The initial results of this work were given previously in GEAP 3397.

The fabricability, elevated temperature strength, steam corrosion resistance, and hydrogen pickup of various binary, ternary, and quaternary alloys were studied experimentally.

This report describes progress made in the study of the feasibility of the concept, plant arrangement studies, and considerations pertaining to partial plant attendance. It is concluded that both the indirect and direct boiling cycles are feasible. The recommended reference design is the direct cycle, in a single capsule, with the turbine, generator and pumps on a single shaft.

The reason for initiating the compact drive study for the T7 tanker was to investigate control rod drive size, location, and removal space requirement factors and select the control rod drive mechanism which would allow optimization of the over-all size of the containment vessel. Approximately twelve mechanical/hydraulic control rod drive arrangements were considered during this study.

This document presents the results of investigations concerned with the conceptual design of a 300 KWe "Capsule" nuclear power plant.

The Fuel Cycle Development Program is a basic development program of boiling and other water reactor technology. This is the first report to the U.S. Atomic Energy Commission of activities carried out under this program. It constitutes a review of the progress from April 27, 1959 through July 31, 1960.

Development of nuclear reactor cores having high power density, long fuel life, and low fabrication costs is the objective of this program sponsored by the AEC. It will be achieved by extensive development and testing carried out in conjunction with the design, construction and operation of the Big Rock Point plant of the Consumers Power Company in Michigan. This second quarterly report describes the progress achieved from July through September 1960.

BOUNCE III is a program which was written for the IBM-704 as part of a study of the parameters of the neutron distribution in a large thermal column. The program calculates the eigenvalues and corresponding eigenvectors of the matrix resulting from a diffusion-theory, multigroup description of the thermal neutron spectrum.

Information obtained in a survey of control rod applications and operating experience is reported. This survey covered the Army Nuclear Power Program, the Organic Moderated Reactor Experiment, the Yankee Atomic Power Plant, the Shippingport Pressurized Water Reactor and the Argonne Experimental Boiling Water Reactor.

This study was designed to determine how swaging stainless steel over unground pellets will affect the integrity of fuel rods, and to find maximum pellet diameter variation that can be tolerated.

It is proposed in this report to describe a new two-phase flow mixing length model. It consists of treating the two-phase system as a continuous medium and applying to it the methods and assumptions so widely accepted in single-phase turbulent flow. Because it reduces a two-phase problem to an equivalent one in single-phase flow, the proposed model can be readily used to predict the effects of flow rate, channel geometry, and channel direction in two-phase flow.

This Preliminary Hazards Summary Report has been prepared for submission to the United States Atomic Energy Commission in compliance with Part 50 of the regulations governing the licensing of production or utilization facilities, pursuant to the Atomic Energy Act of 1954, as amended, and contains the general information required by 10 CFR 50.34.

Writing the VARI-II Program was motivated by the need for a method of analyzing the results for the Absorber Burn-Up Experiment in progress at the Vallecitos Atomic Laboratory.

A reference design of a large high power density core has been established representing the available technology as of August, 1960. Reference core performance and cost should improve considerably after incorporation of improvements now under study.

The General Electric Company is proceeding with an irradiation program to proof test a representative array of Savannah I fuel rods. Irradiation of a test assembly containing Savannah I fuel rods has begun and it is proposed that the results of this irradiation will permit an advance evaluation of the fuel performance and fuel burnup in the Savannah I reactor. This report covers the first two quarters of the reporting period. All aspects of the subject program have been consolidated and applicable portions are discussed in some detail.

The objective is to establish, from test results, the relationship between pressure loss and orifice configuration, mass rate, and steam quality, for a Fuel Cycle fuel assembly.

Equipment and techniques for remote examination of irradiated fuel assemblies applicable to the Maritime Program are described. The following subjects are covered: visual and photographic examination, dimensional measurements, gamma activity scanning, fission gas release and fuel rod void volume determinations, density measurements, metallographic examination, and radiochemical burnup analysis.

Experiments were carried out to determine the erosion, oxidation and dimensional characteristics of purposely defected fuel elements containing unsintered UO2 powder prepared by the swaging technique. The experiments were conducted in an out-of-reactor loop under superheat conditions of pressure, temperature, flow velocity and steam chemical composition.

The basic work for development of a scheduling computer for the Consumers' Big Rock Plant is outlined in this report. The computer's purpose is to make feasible higher power densities by operation closer to fuel element burnout limits, and to maximize fuel burnup.

The fourth quarter of effort applied to the Consumers R & D Program is summarized in the following: high power density fuel development, fuel fabrication development, stability, heat transfer and fluid flow, physics development, operation planning and test coordination.

The pressure drop of a new fuel element design concept of spacing rods by means of helical wire springs was investigated experimentally and analytically.