Carbide Fuel Development: Phase 1 Report, Period of May 15 to September 15, 1959 Page: 8
viii, 100 p., some folded : ill. ; 28 cm.View a full description of this report.
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The results of heat transfer calculations for the EBR-I indicate that six uranium carbide fuel
rods, with a maximum temperature of 1700 F (930*C), may be substituted for the 36 metallic urani-
um rods in each fuel subassembly. (See Fig. 3.6.) Since the EBR-I is a relatively low specific
power, low temperature reactor, there is less incentive to utilize the high specific power, high
temperature potential of carbide fuel in this reactor.
The thermal conductivity of a number of the fuels is unknown. It was necessary to make some
assumptions of thermal conductivity values, in order to do the above heat transfer analyses. The
thermal conductivity of UC is known only to 1400 *F (760C). Since the variation between room
temperature and 1400 F is slight, the same thermal conductivity was assumed up to 3600*F (19800C).
The PuC-UC thermal conductivity is unknown, and it was assumed to be similar to that of UC.
The PuO2-UO2 thermal conductivity is also unknown, and it was assumed to be similar-to that of
UO2. In the event the thermal conductivities are found to be appreciably different from the as-
sumptions, the heat transfer analyses would-change accordingly.
Effect of Fuel to Clad Gap
A 0.5 mil radial helium gap between the carbide and inner clad surface was assumed in ob-
taining the above heat transfer results. A larger gap would result in an excessive temperature
drop across the helium, which would reduce the efficient utilization of the high fuel temperature.
(See Table 3.2.) A smaller gap is impractical. The temperature drop across the cladding and
the coolant film is comparatively small and its influence is therefore subordinated to that of the
helium gap.
Several approaches can be taken to design the fuel to clad gap. Initial ceramic fuel element
designs, such as PWR UO2 rods, considered the relative expansion of fuel and cladding. By speci-
fying very close dimensional tolerances on the OD of the fuel ( 0.0005 in. for PWR) and ID of the
cladding, the fuel and cladding were designed to just touch at operating conditions. The fuel rods
designed on this principle performed well, but were expensive to fabricate.
Recent irradiation experience on UO2 with less stringent dimensional tolerances has served
as the basis for a less conservative approach in the design of the fuel to clad gap; Runnallsi and
Robertson have reported that the surface temperature of UO2 starting with a 0.017 in. diametral,
cold, fuel-clad gap did not rise during irradiation more than 210 F (100*C) over that of UO2 starting
with a 0.005 in. diametral, cold, fuel-clad gap. Tests were with 0.67 in. diameter pellets clad in
Zircaloy-2; center temperatures were near the melting point. The explanation proposed was
that cracked segments of oxide shift radially outwards and contact the cladding. The fuel-to-clad
temperature drop then becomes a function of contact pressure and the surface condition of fuel
and cladding. Ability of fuel to relocate itself from an unfavorable heat transfer position to a more
favorable one has been noted by Bates and Roake.2 They irradiated UO2 powder packed to 4 g/cc
in Zircaloy-2, at heat fluxes as high as 700,000 Btu/hr/ft2 at central melting temperatures. The
fuel sintered to at least 9.3 g/cc during irradiation, and the shrinkage was taken axially. The
dense UO2 filled the cladding radially. It is reasonable to assume that UC will behave in the same
manner as UO2. The coefficient of expansion of UO2 is similar to that of UC, and although the
thermal conductivity of UC is considerably better than that of UO2, UC is still expected to crack
at operating temperature gradients (see next section). The decreased allowable tolerances would
decrease the fuel cycle costs below the estimates made in Section 3.3.
An additional method of circumventing the problem of strict tolerance is the use of a low melt-
ing point metal bond between the fuel and cladding. Sodium or NaK are compatible with fuel and
cladding; initial experiments with lead show some promise. The experience with low melting8
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Bolomey, R.; Lazerus, S.; Sapir, J.; Sofer, G.; Steinmetz, H.; Strasser, A. et al. Carbide Fuel Development: Phase 1 Report, Period of May 15 to September 15, 1959, report, October 15, 1959; Washington D.C.. (https://digital.library.unt.edu/ark:/67531/metadc502678/m1/18/: accessed July 16, 2024), University of North Texas Libraries, UNT Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.