MODERATOR HEAT TRANSFER ANALYSIS FOR SRE-PEP THIRD CORE

Technical Report · Sun Sep 01 00:00:00 EDT 1963

OSTI ID:4072518

Johnson, C G

Moderator coolant gap widths for SRE-PEP are based upon a two-region 33 fuel element core with 3 special test fuel elements and 30 standard fuel elements having radial power peaking factors of 1.707 and 1.160, respectively, and operating at a mixed-mean coolant outlet temperature of 1192 deg F. Operating moderator coolant gap widths of 0.210-inch and 0.165-inch for the seven central test fuel elements (3 special plus 4 standard) and twenty-six standard driver fuel elements, respectively, will allow adequate moderator element cooling for the SRE-PEP core operating at 30 Mw(t) power. Between the first and second row elements, the gap width will be 0.1875-inch. The core tankreflector annulus exit temperature must be maintained at 1132 deg F for standard 30 Mw(t) reactor operation. For reactor power operation above 30 Mw(t), the required mixedmean coolant outlet temperature and annulus exit temperature for adequate moderator cooling are shown. Analysis of the division of heat flow between the main sodium and moderator coolants indicates that on the average, 65.7% of the heat generated within the graphite moderator is transferred to the moderator coolant. Therefore, curves, tables and detailed calculations hgave been based on a 2/3, 1/ 3 power split between the moderator coolant and main coolant, respectively. This result is independent of uniform graphite thermal conductivity changes due to cumulative exposure to fast neutrons. The effect of uniform changes in thermal conductivity within the graphite is merely to alter the magnitude of the temperatures and not the distribution of temperature. This is indicated by the isotherms given for graphite conductivities of 12 and 42 Btu/hr-ft- deg F, respectively. Studies of the central moderator element indicate that maximum graphite temperatures will eventually reach about 1260 deg F near the top of the element as graphite thermal conductivity is reduced by fast neutron exposure. According to the isotherms, a maximum graphite temperature increase of 144 deg F above the datum of 925 deg F may be expected at the midpoint of the central moderator element. At the beginning of life when graphite moderator thermal conductivity is approximates 42 Btu/hr-ft- deg F, the overall core average moderator temperature is 938 deg F. After five years of reactor irradiation, damage by fast neutrons reduces the conductivity to approximates 12 Btu/hr-ft- deg F and average moderator temperature increases to 971 deg F. The maximum horizontal temperature variation across the flats of the outside Zircaloy-2 clad was estimated to be 23.8 deg F. This variation occurs at the outlet to the central moderator element. The moderator coolant temperature difference at this point is 26.2 deg F. The moderator volume fraction at various temperatures is given for the central element and maximum power standard moderator element, respectively. The moderator cooling analysis was based upon 5.85% of total fission power deposited as heat in the moderator. (auth)

• Research Organization: Atomics International. Div. of North American Aviation, Inc., Canoga Park, Calif.
• DOE Contract Number: AT(11-1)-GEN-8
• NSA Number: NSA-18-015448
• OSTI ID: 4072518
• Report Number(s): NAA-SR-Memo-9183
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-64
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
COOLANTS
COOLING
DIAGRAMS
DISTRIBUTION
FAST NEUTRONS
FUEL CANS
FUEL ELEMENTS
GRAPHITE MODERATOR
HEAT TRANSFER
HEATING
HIGH TEMPERATURE
IRRADIATION
MATERIALS TESTING
MODERATORS
OPERATION
POWER
REACTOR CORE
REACTORS
SODIUM
TABLES
TEMPERATURE
THERMAL CONDUCTIVITY
THERMODYNAMICS
TRANSIENTS
VARIATIONS
ZIRCALOY
ZONES