630A MARITIME NUCLEAR STEAM GENERATOR. Progress Report No. 7

Technical Report · Mon Mar 16 00:00:00 EST 1964

DOI:https://doi.org/10.2172/4004790· OSTI ID:4004790

This progress report covers the period from November 15, 1963 to February 15, 1964. A study indicated that the most desirable type of blower drive turbine is one using main turbine throttie steam conditions and exhausting to the main turbine cross-over line. Preliminary planning for the initiation of a dynamic structural analyses of the overall steam generator was completed. External pressure loading and thermal stress calculations show that the calandria has a suitable design margin. A revised fuel latch operable from the rear face of the core was designed. A study was initiated to determine the feasibility of substituting Zircaloy for the stainless steel tubing within the active core. Preliminary sizing of control rod extensions and gang plates was completed. Initial loading of the second configuration of the 630A critical experiment reactor was completed. Detailed power distributions were measured in the 11 typical positions. Subcritical and critical rod worth curves were obtained in the critical experiment with up to 132 shim rods in the core. Moderator temperature coefficient measurements were made and agreed well with analytical data. Critical experiment correlation of fine radial power calculations in the revised mock-up showed good agreement. Performance specifications were prepared for a 1-Mw(e) power plant. Parametric thermal analyses, based on various tube sheet thicknesses and heating rates, were completed for the top and bottom tube sheets. The shield plug was redesigned to accommodate additional shim rods and to facilitate fabrication and moderator flow adjustments. Calculations of the operating and shutdown dose at the nuclear sensor location was started. Studies were performed to determine the size of the port openings needed to prevent buckling in the containmert vessel in the event of ship sinking. A blower shaft static seal was built and tested with satisfactory results. Manufacture and procurement of all parts for two developmental dynamic rod actuators were completed. The two-tube boiler was modified and operated satisfactorily on automatic control down to 20 percent power. A reactor control system that meets response and accuracy requirements was simulated on the analog computer. Investigation of a possible test site for power testing the prototype reactor was continued with cost estimates and a facility layout drawing being made for the IET facility at ITS. Specimen ORF-4 was removed from test after 4800 hours of actual operation at 1300 deg F, equivalent to approximately 15,000 hours of full- power 630A operation. An estimated 40 percent burnup of the /sup 235/U atoms occurred. Creep tests on warm-finished fuel sheet have accumulated more than 12,000 hours at 630A stress and temperature conditions with no greater than 0.75 percent plastic creep. Cladding-stock creep tests have surpassed 15,000 hours at various combinations of stress and temperature with satisfactory results. (auth)

• Research Organization: General Electric Co. Advanced Technology Services, Cincinnati
• DOE Contract Number: AT(40-1)-2847
• NSA Number: NSA-18-021668
• OSTI ID: 4004790
• Report Number(s): GEMP-274
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-64
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
ACCIDENTS
ANALOG SYSTEMS
BOILING
BUCKLING
BURNUP
CANNING
COMPRESSORS
COMPUTERS
CONFIGURATION
CONTROL
CONTROL ELEMENTS
CONTROL SYSTEMS
CREEP
CRITICAL ASSEMBLIES
DIAGRAMS
DISTRIBUTION
ECONOMICS
ENVIRONMENT
FABRICATION
FLUID FLOW
FUEL ELEMENTS
FUELS
GENERATORS
HEAT EXCHANGERS
HEAT TREATMENTS
HEATING
HIGH TEMPERATURE
MATERIALS TESTING
MATHEMATICS
MEASURED VALUES
MECHANICAL STRUCTURES
MOCKUP
MODERATORS
MONITORING
MOTION
OPERATION
PERFORMANCE
PLANNING
PLATES
POWER
POWER PLANTS
PRESSURE
PRESSURE VESSELS
PROPULSION
RADIATION DOSES
REACTIVITY
REACTOR CORE
REACTOR FUELING
REACTORS
RODS
SEALS
SHEETS
SH