ORNL-4191 - the Molten Salt Energy Technologies Web Site

March 17 following the fission product leak. At
full power, temperatures in the fuel salt ranged
from -,, 545OC in the cold leg return line up to
*72OoC in the core outlet pipe. Since nuclear
heat was removed through the core graphite to the
cooling coils around the outside of the Hastelloy N
core body, the graphite temperature was 55OoC,
or some 7OoC below the average fuel salt temperature
in the core. Temperatures of the metal walls
of the component parts of the loop (Hastelloy N)
ranged from a low of 51OOC in the core body (where
maximum cooling was used) to -73OOC in the core
outlet pipe, where there were no provisions for
cooling.
The temperature distribution of the salt and loop
components was significantly altered when the
reactor was down. Under this condition, salt
temperatures ranged between rv 535OC in the coldleg
return line to -670°C in the top of the core
graphite fuel passages. The core outlet pipe was
67OOC with no nuclear heat, as compared with
"73OoC for full power operation. The core outlet
pipe temperature of 73OOC is believed to have
contributed to the outlet pipe failure as discussed
in a following section.
15.4 SALT CIRCULATION BY CONVECTION
In the first in-pile molten-salt convection loop,
the salt circulation rate of 5 to 10 cm3/min was
substantially below the calculated rate of -45
cm 3/min at operating temperature, and frequent
loss of flow occurred. In order to improve salt
circulation rate and reliability in the second loop,
the salt flow channels at the top and bottom of the
graphite core were redesigned to increase the
flow area and improve the flow path (refer to
Fig. 15.2). Further, the top and bottom of the
core section that were horizontally oriented in
the first loop were inclined So to minimize trap-
ping of any gas released from the salt, since it is
believed that formation of gas pockets often re-
sulted in flow stoppage.
Salt circulation in the second loop was estimated
to be 30 to 40 cm3/min; this was determined by
making heat-balance measurements around the
cold-leg return line and by adding an increment of
heat in a stepwise fashion to one point in the loop
and recording the time required for the heated salt
to traverse a known distance as monitored by
thermocouples around the loop circuit. This flow
180
rate is a fivefold increase over that observed in
the first loop and is attributed to the modification
described. However, occasional loss of flow still
occurred. One possible explanation for this is
that a sufficient temperature difference was not
maintained between the salt in the hot and cold
legs. This is supported by the fact that flow,
when lost, could be restored by adjusting the
temperatures around the loop circuit. Since oc-
casional flow loss did not adversely affect in-pile
operation, this was not considered to be a problem
of any serious consequence.
15.5 NUCLEAR HEAT, NEUTRON FLUX,
AND SALT POWER DENSITY
Nuclear heat was determined at various loop
positions by comparing electric heat input and
cooling rates with the reactor down and at full
power (30 Mw). Figure 15.3 shows the results of
these measurements. Although nuclear heat meas-
urements based on such heat balances are not
precise because of variations in the temperature
distribution around the loop and variations in heat
loss at different loop positions (different power
levels) as well as necessary estimates of the
exact inlet and outlet temperatures of the several
air-water coolant mixtures, they provided a good
basis for determining nuclear heat and resultant
fission-power density during operation. Based on
these determinations, reactor gamma heat with the
loop fully inserted and filled with unfueled salt
was 4200 w. Fission heat was computed by sub-
tracting this value from the total heat generation
obtained when fuel salt was in the loop.
In the earlier part of the operation with fueled
salt (1.73 mole % U), the loop contained 17.84 g
of uranium, 93% enriched, in a salt volume of
76.2 cc. A fission heat of 8600 w in the fully
inserted position was determined, leading to an
estimate of the average thermal neutron flux of
1.18 x 10l3 neutrons cm-* sec-', or an average
fuel-salt power density of 113 w/cc. Assuming
from a neutron transport calculation by H. F.
Bauman that the core/average flux ratio was 1.33,
the average power density in the core salt was
150 w/cc at full power. The power density in the
forward core tubes is estimated by using Bauman's
results to be 180 w/cc.
The average thermal neutron flux in the salt was
also independently determined from flux monitors