ORNL-1816 - the Molten Salt Energy Technologies Web Site

ANP QUARTERLY PROGRESS REPORT
systems were compiled, and appatutus was de-
veloped for measuring the over-all radial tempera-
ture difference (or heat transfer coefficient) for a
modified version of the flat-plate system in which
it will not be necessary to measure fluid temper-
atures.
Several approxi mate mathematical temperature
solutions for forced-flow volume-heat-source en-
trance systems were developed which can be used
to predict the temperature structure of struts or
screens located in circulating-fuel reactor flow
passages. The question as to whether electric
currents, which generate heat in the circulating
f I u i ds of experi menta I volume- h eat-source s y stems,
affect the fluid flow characteristics was investi-
gated. It was found that the hydrodynamic structure
was not influenced by the presence of the electric
currents.
The enthalpies and heat capacities of NaF-ZrF,-
UF, (53-43-4 mole %) were determined; the heat
capacity in the solid state over the temperature
range 70 to 525OC was found to be 0.18 cal/g.OC,
and the heat capacity in the liquid state over the
temperature range 570 to 885T was found to be
0.26 cal/go°C. The enthalpies and heat capacities
of LiF-KF-UF, (48-48-4 mole %) were also ob
tained; the heat capacity in the solid state over
the temperature range 125 to 465OC was found to
be 0.234 + (0.95 x lO-,)t cal/gSoC, and in the
liquid state over the temperature range 565 to
88OOC it was found to be 0.657 - (3.93 x 10-4)t
cal/g. OC. The viscometry equipment used earlier
was modified so that the accuracy of liquid vis-
cosi ty measurements was significantly increased.
The more accurate measurements obtained for
NaF-ZrF,-UF, (53.5-40-6.5 mole %) were 30%
lower than the preliminary values reported previ-
ously. The thermal conductivity of molten NaF-
ZrF,-UF, (53.5-40-6.5 mole %) was found to be
1.2 Btu/hr.ft. OF. The thermal conductivities of
the alkali fluoride mixture NaF-KF-LiF with and
without UF, were compared in the liquid and solid
states.
9. Radiation Damage
The program of MTR irradiations of lnconel
capsules containing fused fluoride fuels has been
continued, and irradiated capsules containing both
UF,- and UF,-beoring mixtures have been ex-
ined. There was practically no corrosion of the
irradiated UF3-bearing capsules nor of one of the
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irradiated UF4-bearing capsules, There were no
significant differences in the iron, chromium, or
nickel contents of the irradiated fuels, as com-
pared with the starting fuel batches, and there was
no evidence of segregation of either uranium or
impurities, An improved version of the capsule
irradiation facility has been put into service, and,
to speed up the program, arrangements have been
made for irradiating two capsules simultaneously
in separate, but adiacent, facilities in the MTR.
Examinations of a welded nickel capsule con-
taining a UF,-C,F ,6 solution that was irradiated
in the ORNL Graphite Reactor indicated the un-
suitability of the solution for use as fuel in a
mockup of a circulating-fuel reactor.
A miniature in-pile loop designed for insertion
in a vertical hole in the LITR was bench tested
and was found to be satisfactory, The bench test
included four freezing and melting cycles in 260 hr
of operation at a maximum temperature of 466OF.
Since it was possible to freeze and melt the fused
salt without causing failure of the loop, it may be
advisable to fill the in-pile loop before it is in-
serted in the reactor and then to melt the fuel
mixture after the loop is in position, A Delco
motor has been rebuilt to withstand radiation and
the high temperatures to be used for the in-pile
loop. Heat transfer calculations that predict the
thermal behavior of the loop were completed. A
second model of the loop for insertion in the LlTR
is being constructed.
Plans were completed for an in-pile study of the
removal of xenon from molten fluoride fuels under
ART conditions.
The horizontal type of fuel-circulating loop
designed for irradiation in the LlTR has been
operated out-of-pile with a non-uranium-bearing
salt and is now being inserted in the HB-2 facility
of the LITR. The creep-test apparatus for testing
lnconel at high temperatures in the MTR is being
bench tested. The stress-corrosion apparatus for
LlTR operation has been successfully bench tested,
and an in-pile apparatus is being constructed.
Remote metallographic studies of solid fuel ele-
ments were continued, and additional information
on the relationship between UO, particle size and
radiation damage was obtained.
10. Analytical Studies of Reactor Materials
Methods were developed for the determination
of uranium metal and UF, in NaF-KF-LiF-base
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