AN EMPIRICAL PREDICTION OF TEMPERATURE GRADIENTS IN MODERATOR GRAPHITE WITHIN OPERATING NUCLEAR REACTORS

Technical Report · Thu Feb 15 04:00:00 EST 1962

OSTI ID:4807955

Carter, R L

The thermal conductivity of artificial graphite as a function of temperature, integrated fast neutron irradiation experience, and temperature during irradiation was deduced by a semi-empirical procedure. Within a section of circular symmetry centered upon a fuel element of a graphitemoderated reactor core, the volume heat generation by gamma and neutron degradation were estimated as a function of a single radial position variable. Assuming the heat flow from the graphite moderator volume to be entirely radial, self-consistent calculations were made for temperature distributions at various positions within the reactor core at various times after reactor startup. Results are given for the Sodium Reactor Experiment (SRE), for a calandria-core 255 Mwe SGR, and for the reactor at the Hallam Nuclear Power Facility. (auth)

• Research Organization: Atomics International. Div. of North American Aviation, Inc., Canago Park, Calif.
• DOE Contract Number: AT(11-1)-GEN-8
• NSA Number: NSA-16-011232
• OSTI ID: 4807955
• Report Number(s): NAA-SR-6616
• Country of Publication: United States
• Language: English

Related Subjects

ABSORPTION
CONFIGURATION
DISTRIBUTION
EQUATIONS
FAST NEUTRONS
FUEL ELEMENTS
GAMMA RADIATION
GRAPHITE MODERATOR
HEAT TRANSFER
HEATING
HIGH TEMPERATURE
INTERACTIONS
IRRADIATION
LIQUID METAL COOLANT
MEASURED VALUES
NEUTRON FLUX
NEUTRONS
NUMERICALS
POWER PLANTS
REACTOR CORE
REACTOR TECHNOLOGY
SCATTERING
SLOWDOWN
SODIUM
SRE
STARTUP
TEMPERATURE
THERMAL CONDUCTIVITY
VOLUME
ZONES