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Title: MARITIME GAS-COOLED REACTOR PROGRAM. Quarterly Progress Report for Period Ending June 30, 1960

Abstract

: 9 ; 9 ? < A > = 6 8 8 = moved from the Hanford in-pile loop after approximately 1000 hrs irradiation. Examination indicated no dimensional changes or discoloration of the model. Some minute cracks were located in the cladding. The fission products which were detected in the loop during irradiation are attributed to these cracks. Design of the reactor vessel is continuing. More detailed calculations of fuel element performance are being carried out using computer codes to determine the effect of different flux distributions. Calculations are also under way to determine the effect on fuel element temperature of perturbations on coolant flow. Work is under way to establish means of analyzing and predicting details of the heat transfer process in fuel elements. This includes meaurement of flow and pressure profiles in large scale models and heat and mass-transfer experiments. Burst tests of 3/8-in. dia, 0.010-in. thick wall Inconel and Hastelloy X fuel cladding thbes indicated bursting pressures greater than 300 psi at 1900 deg F. Detail design of the reactor core is continuing. Temperature in the moderator is being calculated as a function of variation in coolant bleed flow. Stresses in the core support grid are beingmore » analyzed by machine computation. Detail design of the control rod drive mechanism is essentially complete. Tests were started with a ball-nut lead screw running unlubricated in hot helium. No significant wear of the ball-nut device appeared after 5000 cycles at 300 deg F. Work is continuing on heat transfer tests for the propulsion plant heat exchangers. It was determined that the heat exchanger tubes can be placed on a closer pitch which will enable the heat exchangers to be made smaller. Power plant kinetics calculations included the effect of lead variations on turbomachine speeds. Step changes in auxiliary power from zero to full power or vice versa produce changes of approximately 1 per cent in turbine-compressor speed. The low-pressure turbine bypass control permits output power to vary from zero to full power within five seconds. Power control by bypassing to the regenerator rather than a heat dump was found to be practical. Assembly of the full size bearing and seal test rig was completed. Conceptual design of the sealing and lubrication system was completed and is ready for development testing in the full size test rig. Thermodynamic cycle calculations continued with studies of operation at 1500 deg F. The 1300 deg F cycle machinery can be operated at l500 deg F without over- speeding by reblading the high-pressure turbine and increasing the cooling bleed flow. Compressor model tests were continued to evaluate the effect of Reynold's number on stage performance. The expected effect did not occur and efforts are being made to refine the loop instrumentation to permit more accurate measurements. The MGCR critical facility was completed and fuel loading begun. Initial power is limited to one watt. Later operation will be at powers up to 500 watts. Experiments are being carried out in the LRL "Hot Box" facility to evaluate the resonance absorption of neutrons in U/sup 238/ at MGCR temperatures. A series of experiments to determine neutron thermalization and Fermi age in beryllium oxide was run in the linear accelerator. An important preliminary result of these experiments is that the Fermi age in BeO is significantly less than that calculated from the age of beryllium. Reactor nuclear calculations were completed for the preliminary design of the Beryllia Reactor Experiment (BRE). The number of fuel elements, fuel loading, and flux distribution for four alternate configurations was determined. The effect of water flooding on reactivity and the necessary control rod worths was calculated. Nuclear calculations for the MGCR prototype consisted of studies of fuel programs to obtain better fuel economy. It was determined that a 3-zone loading program could approach the« less

Publication Date:
Research Org.:
General Atomic Div., General Dynamics Corp., Sans Diego, Calif.; General Dynamics Corp. Electric Boat Div., Groton, Conn.
OSTI Identifier:
4820804
Report Number(s):
GA-1532
NSA Number:
NSA-16-006212
DOE Contract Number:  
AT(04-3)-187
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-62
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; ALUMINUM ALLOYS; BYPASS; CANNING; CHROMIUM ALLOYS; COMPRESSORS; COMPUTERS; CONFIGURATION; CONTAMINATION; CONTROL ELEMENTS; CONTROL SYSTEMS; CRACKS; DEFORMATION; DISTRIBUTION; EQUATIONS; EXPANSION; FAILURES; FISSION PRODUCTS; FRICTION; FUEL CANS; FUEL ELEMENTS; GAS COOLANT; GAS FLOW; HASTELLOY; HEAT EXCHANGERS; HEAT TRANSFER; HELIUM; HIGH TEMPERATURE; HOMOGENEOUS REACTORS; IN PILE LOOPS; INCONEL ALLOYS; INSTRUMENTS; IRRADIATION; LEAD; LEAKS; LUBRICATION; MACHINE PARTS; MATERIALS TESTING; MEASURED VALUES; MECHANICAL STRUCTURES; MOCKUP; MOLYBDENUM ALLOYS; MONITORING; MOTORS; NEUTRON FLUX; NICKEL ALLOYS; NIOBIUM ALLOYS; NUMERICALS; PERFORMANCE; P

Citation Formats

. MARITIME GAS-COOLED REACTOR PROGRAM. Quarterly Progress Report for Period Ending June 30, 1960. United States: N. p., 1962. Web.
. MARITIME GAS-COOLED REACTOR PROGRAM. Quarterly Progress Report for Period Ending June 30, 1960. United States.
. Wed . "MARITIME GAS-COOLED REACTOR PROGRAM. Quarterly Progress Report for Period Ending June 30, 1960". United States.
@article{osti_4820804,
title = {MARITIME GAS-COOLED REACTOR PROGRAM. Quarterly Progress Report for Period Ending June 30, 1960},
author = {},
abstractNote = {: 9 ; 9 ? < A > = 6 8 8 = moved from the Hanford in-pile loop after approximately 1000 hrs irradiation. Examination indicated no dimensional changes or discoloration of the model. Some minute cracks were located in the cladding. The fission products which were detected in the loop during irradiation are attributed to these cracks. Design of the reactor vessel is continuing. More detailed calculations of fuel element performance are being carried out using computer codes to determine the effect of different flux distributions. Calculations are also under way to determine the effect on fuel element temperature of perturbations on coolant flow. Work is under way to establish means of analyzing and predicting details of the heat transfer process in fuel elements. This includes meaurement of flow and pressure profiles in large scale models and heat and mass-transfer experiments. Burst tests of 3/8-in. dia, 0.010-in. thick wall Inconel and Hastelloy X fuel cladding thbes indicated bursting pressures greater than 300 psi at 1900 deg F. Detail design of the reactor core is continuing. Temperature in the moderator is being calculated as a function of variation in coolant bleed flow. Stresses in the core support grid are being analyzed by machine computation. Detail design of the control rod drive mechanism is essentially complete. Tests were started with a ball-nut lead screw running unlubricated in hot helium. No significant wear of the ball-nut device appeared after 5000 cycles at 300 deg F. Work is continuing on heat transfer tests for the propulsion plant heat exchangers. It was determined that the heat exchanger tubes can be placed on a closer pitch which will enable the heat exchangers to be made smaller. Power plant kinetics calculations included the effect of lead variations on turbomachine speeds. Step changes in auxiliary power from zero to full power or vice versa produce changes of approximately 1 per cent in turbine-compressor speed. The low-pressure turbine bypass control permits output power to vary from zero to full power within five seconds. Power control by bypassing to the regenerator rather than a heat dump was found to be practical. Assembly of the full size bearing and seal test rig was completed. Conceptual design of the sealing and lubrication system was completed and is ready for development testing in the full size test rig. Thermodynamic cycle calculations continued with studies of operation at 1500 deg F. The 1300 deg F cycle machinery can be operated at l500 deg F without over- speeding by reblading the high-pressure turbine and increasing the cooling bleed flow. Compressor model tests were continued to evaluate the effect of Reynold's number on stage performance. The expected effect did not occur and efforts are being made to refine the loop instrumentation to permit more accurate measurements. The MGCR critical facility was completed and fuel loading begun. Initial power is limited to one watt. Later operation will be at powers up to 500 watts. Experiments are being carried out in the LRL "Hot Box" facility to evaluate the resonance absorption of neutrons in U/sup 238/ at MGCR temperatures. A series of experiments to determine neutron thermalization and Fermi age in beryllium oxide was run in the linear accelerator. An important preliminary result of these experiments is that the Fermi age in BeO is significantly less than that calculated from the age of beryllium. Reactor nuclear calculations were completed for the preliminary design of the Beryllia Reactor Experiment (BRE). The number of fuel elements, fuel loading, and flux distribution for four alternate configurations was determined. The effect of water flooding on reactivity and the necessary control rod worths was calculated. Nuclear calculations for the MGCR prototype consisted of studies of fuel programs to obtain better fuel economy. It was determined that a 3-zone loading program could approach the},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1962},
month = {10}
}

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