Maritime Gas-Cooled Reactor Program (Quarterly Progress Report for the Period Ending, March 31, 1959)

Technical Report · Tue Mar 22 04:00:00 EST 1960

DOI:https://doi.org/10.2172/4200518· OSTI ID:4200518

Turbomachinery considerations indicated that it would be desirable to reduce the cycle pressure from 1,000 to 800 psia. The problem of determining the temperature distribution and the resulting thermal-stress pattern within the graphite was considered. Preliminary designs for a heterogeneous fuel element and a semihomogeneous fuel element were developed. Utilization of the Hanford in- pile gas loop for fuel element testing is discussed. Two-group PDQ calculations were run to estimate control rod worth for the preliminary design core under cold, clean conditions. Curves of rod worth versus position were developed for the hot, clean and the cold, clean preliminary design core. A detailed lifetime calculation was made for the preliminary design heterogeneous core. Fuel cycle costs were estimated on the basis of the effect of B in the fuel elements. The schedule and facilities for the critical experiments are discussed in some detail. The speed of the main turbine shaft was tentatively set at about 12,200 rpm. The design and fabrication of a test stand to evaluate shaft seals and seal systems were completed and trial runs were made. The effects of minor heat transfer due to heat leakage, fluid flow, and thermodynamic phenomena on MGCR full-load cycle performance were studied. Operating characteristics of the heat exchanger test facility are described. A critical review was conducted on the desirability of using concentric ducts and valves. Block diagrams outlining reactor power level, outlet temperature, and plant inventory control are presented. Equations which permit the dynamic analysis of a closed-cycle gas-turbine plant were programed for a digital computer. Descriptions were prepared for fluid-mechanical systems. Several methods of purifying He in both storage bank and main loop were investigated. Investigations into the maximum operating temperatures of the various electrical equipment indicate that temperatures up to 140 °F can be tolerated. Preliminary design work was carried out on an emergency cooldown system which uses a stored inert coolant. Work on the irradiation stability of fuel materials indicated that conversion from the monocarbide to the dicarbide can take place at 2,350 °F in graphite bodies containing UC. Densities as high as 96.6% theoretical were obtained in alumina pellets sintered in H Xenon released from irradiated UO₂ granules was measured. Progress was made in the development of high-temperature x-ray-diffraction techniques. Tests were conducted to investigate the effects of coolant and impurities in the coolant on plant materials and the effects of gettering on inhibiting reactions. Economic studies were continued with the objective of determining the optimum size and power ranges for both nuclear and conventionally powered ships.

• Research Organization: General Dynamics Corp., San Diego, CA (United States)
• Sponsoring Organization: US Atomic Energy Commission (AEC); USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• NSA Number: NSA-14-011295
• OSTI ID: 4200518
• Report Number(s): GA--1030; MA-S89--1; GA--744
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ALUMINUM-- CLEANING-- CONTROL-- COOLANTS-- COOLING-- ECONOMICS-- ELECTRONIC EQUIPMENT-- GRAPHITE-- HELIUM-- HYDROGEN-- IMPURITIES-- MGCR-- POWER-- REACTORS-- SHIPS-- SINTERING-- TEMPERATURE-- TESTING-- URANIUM CARBIDES-- URANIUM OXIDES-- XENON
CONTROL ELEMENTS-- CRITICALITY-- ECONOMICS-- FUEL ELEMENTS-- GAS COOLANT-- GRAPHITE-- LIFETIME-- MGCR-- PLANNING-- POWER PLANTS-- PRESSURE-- REACTOR CORE-- REACTORS-- SHIPS-- TEMPERATURE-- TESTING-- THERMAL STRESSES
FLUID FLOW-- HEAT EXCHANGERS-- HEAT TRANSFER-- LEAKS-- MECHANICAL PROPERTIES-- MGCR-- REACTORS-- SEALS-- SHIPS-- TESTING-- TUBES-- TURBINES-- VALVES
Fuel loading
Nuclear Criticality Safety Program (NCSP)
Reactor Technology
Turbomachinery