Abstract; Technical report describing a low-flow autoclave system developed for out-of-pile corrosion testing of materials in controlled environment steam up to 500 C. The system has been set up in triplicate to provide for the exposure of various zirconium alloys to steam at 300, 400, and 500 C. The oxygen and hydrogen of the steam were controlled at 25 ppm and 3 ppm, respectively, to simulate the gas conditions from radiolytic water decomposition found in a boiling water reactor. The autoclave internals were so designed to result in a temperature variation between specimens under test of less than 2C.

The General Electric Company is designing and building a 180,000 kilowatt nuclear power plant for the Commonwealth Edison Company at a site 47 miles southwest of Chicago. The construction permit was issued on May 4, 1956 but is subject to submittal to the Commission of a final hazards summary report and a finding by the Commission that the final design provides reasonable assurances that the health and safety of the public will not be endangered by operation of the reactor in accordance with the specified procedures.

From introduction: "This report sets forth the results of a technical and economic analysis of automatic propulsion system control as a possible design improvement in the direct cycle boiling water reactor propulsion system in a T7 tanker."

From introduction: "This is the thirteenth in a series of quarterly reports which will cover the progress and results from the conceptual design, economic evaluations and research and development work performed by the General Electric Company as part of the Nuclear Superheat Project."

Stability tests are to be made in the VBWR (boiling water reactor). The tests are of three types: steady state measurements, rod oscillator tests, and transient tests.

A design and economic analysis is presented for a 44,000 DWT tanker utilizing a 27,300 shp direct cycle boiling water reactor propulsion system. Included are design descriptions and drawings, estimates of operating and capitol costs for the reference ship, estimates of long range potential, descriptions of required research and development programs, a preliminary safeguards analysis, an analysis of ship's motion on boiling water reactor performance, a summary reactor physics analysis, and a discussion of direct cycle accessibility considerations. Also included are summary descriptions for various design alternatives which were considered.

Quarterly report discussing progress on the Fast Ceramic Reactor Development Program. Information is reported on vented fuel production, fuel testing in TREAT, fuel performance evaluation, fast-flux irradiation of fuel, and reactor dynamics and design.

Work performed during the quarter is summarized by: direct measurement of fission gas pressure, loop operations, performance of UO2 fuel, UO2 grain growth and melting studies.

From introduction: "This is the fourth of a series of quarterly reports which will cover the progress and results from the conceptual design, economic evaluations and research and development work performed by the General Electric Company as part of the Nuclear Superheat Project."

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.

Task A: Modification and Preparation of Experimental Facility. Facility engineering and layout is about seventy-five percent complete. Task B: Design and Construction of Test Sections. The major dimensions and characteristics of the metal and glass test sections have been calculated. One feasibility test of the electrically conducting coating on samples of glass tubing has been completed. Task C: Design and Construction of Test Stand, Task E: Pressure and Temperature Instrumentation for Test Section and Task F: Power Supply for Test Section. Preliminary engineering has been initiated on these tasks. The planned approach has been defined in each case. For Task E the transducer specifications have been defined and quotations on and/or sample units of the transducers have been requested. Tasks C and F can proceed with detailing as soon as drafting on Task B is about 50 percent complete. This point is scheduled to be reached during the first part of July. Task D: Void Fraction Instrumentation. The requirements for the x-ray instrumentation have been considered in the course of Task B and the x-ray power supply is presently on hand. The detailed engineering effort on this task is not scheduled to begin before July.

Quarterly progress report discussing activities related to the Vallecitos Boiling Water Reactor (VBWR) and related facilities.

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.

From introduction: "This is the eighth of a series of quarterly reports which will cover the progress and results from the conceptual design, economic evaluations and research and development work performed by the General Electric Company as part of the Nuclear Superheat Project."

This report is submitted to the U. S. Atomic Energy Commission as a final summary safeguards and hazards evaluation of a proposed test reactor at its Vallecitos Atomic Laboratory in Alameda County of California. It is the purpose of this report to provide sufficient data to obtain an AEC facility license for the reactor.

From introduction: "This report provides the preliminary design for the nuclear portion of a 300 MWe Boiling Water Separate Superheater Plant to be in operation by 1967."

This report contains technical specifications and supporting data for the Critical Experiment Facility, information to justify operation of the facility, procedural control to be used to ensure safe operation, changes in neutron instrumentation and safety system, an evaluation of the safety of the Facility. It also contains descriptions of the site, the facility and the Critical Assemblies, operating standards and procedures. Amendment no.18 to license application for Critical Experiment Facility is also attached.

From introduction: "This is the sixth of a series of quarterly reports which will cover the progress and results from the conceptual design, economic evaluations and research and development work performed by the General Electric Company as part of the Nuclear Superheat Project."

From introduction: "This is the twelfth of a series of quarterly reports which will cover the progress and results from the conceptual design, economic evaluations and research and development work performed by the General Electric Company as part of the Nuclear Superheat Project."

The 300 series stainless steels had been selected originally as the reference fuel cladding material for utilization in several superheat reactor (SHR) systems being built as part of the United States Atomic Energy Commission (USAEC) program. The adequacy of the general corrosion resistance of Type-304 stainless steel for superheat fuel cladding was confirmed in the Phase I portion of the study reported previously. Fuel jacket failures that occurred in Type-304 stainless clad fuel elements exposed in the Vallecitos boiling water reactor superheated steam loop indicated the questionable dependability of such stainless steels for this SHX fuel cladding application. The following conclusions are based on the out-of-pile general corrosion evaluations completed to date on Incoloy-800 as a fuel cladding for nuclear superheat applications: 1. The corrosion data from 4000-hour heat transfer tests indicate good corrosion resistance up to at least a 1300°F metal temperature. By use of a devised method of data treatment, the general corrosion for three years exposure at 1300°F can be calculated to average 0.0016 inch with an upper 95 percent confidence limit of 0.0033 inch. 2. A compositionally-disturbed layer develops at the metal-oxide interface. The disturbed layer depth is a function of time and temperature of exposure. …

From introduction: "This is the eleventh of a series of quarterly reports which will cover the progress and results from the conceptual design, economic evaluations and research and development work performed by the General Electric Company as part of the Nuclear Superheat Project."

From introduction: "this is the fifth of a series of quarterly reports which will cover the progress and results from the conceptual design, economic evaluations and research and development work performed by the General Electric Company as part of the Nuclear Superheat Project."

This report presents the reactor description and safeguard evaluation for an open pool research and test reactor being supplied to the State College of Washington, Pullman, Washington, by the General Electric Company.

The raw materials and fabrication procedures employed in preparing thirty two zirconium alloy compositions for evaluation as described in GEAP-3979 are reported. Considerations involved in the extension of reported laboratory procedures to larger scale production are discussed.