MARITIME GAS-COOLED REACTOR PROGAM QUARTERLY PROGRESS REPORT FOR THE PERIOD ENDING DECEMBER 31, 1959

Technical Report · Tue Oct 31 00:00:00 EST 1961

OSTI ID:4123999

8 8 : 8 7 < 6 9 4 : = F J ;= 7 : : 6 5 % el element was used for determining the friction factor and "hot channel" factor of 19 closely spaced rods. A preliminary study of the effect of reactor support mounting on the stresses and reactions of the main coolant piping was conducted to determine whether or not the reactor could be mounted on fixed supports or whether or not it must be flexibly mounted with a single degree of freedom fore and aft, and to determine the values of stresses and reactions for the fixed and flexible mountings. An analysis to determine fuel-element conditions for a range of core pressure drops for the Mark I reactor design point was completed. For a selected Mark I design, the effect of changes in the dimensions of the hot channel was calculated. A study was made of the feasibility of returing the high-temperature core outlet gas back through a vertical center duct in the core to a point below the core. The design investigation program for the control-rod drive mechanism is discussed. Fluid Systems and Plant Arrangment: The isothermal pressure-drop tests for air and helium were completed. The data, which represent approximately 7OO individual runs, covered a range of Reynolds numbers from 25,000 to 500,000. A feasibility study was made for a system of separate reactor inlet and outlet ducts, in which the high-temperature outlet duct is cooled by an extenal tracer flow of helium taken from the high-pressure compressor discharge. Twenty-one conceptual designs were prepared for the main-coolant stop valve, and two valve types were selected for further engineering study. A more precise estimate was made of the maximum surface temperatures that would be reached in the reactor core as a result of a loss-of-coolant accident at fuel power. The computer program for the solution of the system transient equations was made operational. An evaluation study of separate- and concentric-duct arrangements for the Maritime Gas-Cooled Reactor (MGCR) plant was completed. right piping-arrangement cases were run on the IBM 704 computer to determine pipe stresses and reactions. Rotating Machiner: Performance characteristics of the MGCR turbines were computed in accordance with an IBM program for off-design-point conditions, and turbine maps were compiled on the basis of these results. Further study on the prototype low-pressure turbine design resulted in a reduction of the connecting shaft overhang from 48 to 24in., with accompanying changes in critical speed. Three seal configurations were tested: the floating oil-gas seal, the Stein seal assembly, and the Stein seal with floatingring breakdown seal. Investigations of the operation of the constant-speed turbocompressor established an allowable speed variation of plus or minus 5%. Other studies indicated that the turbocompressor system has satisfactory steady-state characteristics over a range of sea-water temperature from 28 to 100 deg F and at reduced reactor power levels, reduced plant inventories, and reduced reactor outlet temperatures. A reflux heat dump was considered for removiug heat from the power-turbine bypass stream. Reactor Physics: Because BeO can be Produced only in small blocks ( approximately 60 in./ sup 3/) and because the grinding of large quantities of BeO blocks to close tolerances is prohibitively expensive, an aluminum structure was designed to support the individual moderator blocks (BeO or graphite) and fuel elements. Calculations were made on the reactivity of the BeO critical assembly under various conditions, for inclusion in the safeguards report. These calculations showed that there is not much difference in the worth of the controlling device for the BeO- and graphite-moderated cores. The temperature behavior of the proposed BeO-moderated critical facility was investigated under various transient conditions. The Adler-Nordheim-Hinman resonance integral codes were reprogrammed for the IBM 704. This was

• Research Organization: General Atomic Div., General Dynamics Corp., San Diego, Calif.
• DOE Contract Number: AT(04-3)-187
• NSA Number: NSA-15-001049
• OSTI ID: 4123999
• Report Number(s): GA-1195
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-61
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
ACCIDENTS
AIR
ALUMINUM
ALUMINUM ALLOYS
ANALOG SYSTEMS
BERYLLIUM MODERATOR
BERYLLIUM OXIDES
CARBON MONOXIDE
CHROMIUM ALLOYS
COMPRESSORS
COMPUTERS
CONTROL ELEMENTS
COOLANT LOOPS
COOLANTS
COOLING
CRITICALITY
EXCURSIONS
FLUID FLOW
FUEL ELEMENTS
GAS COOLANT
GAS FLOW
GASES
GRAPHITE MODERATOR
HEAT TRANSFER
HELIUM
HIGH TEMPERATURE
IBM
IBM 704
INCONEL ALLOYS
LATTICES
LEAKS
MECHANICAL STRUCTURES
MGCR
NICKEL ALLOYS
NIOBIUM ALLOYS
OILS
PIPES
PRESSURE
PROGRAMMING
PROPULSION
REACTIVITY
REACTOR CORE
REACTORS
RESONANCE NEUTRONS
RODS
SAFETY
SEA
SEALS
SHIPS
SIMULATORS
SLOWDOWN
STRESSES
SURFACES
TEMPERATURE