ORNL-4191 - the Molten Salt Energy Technologies Web Site

4 2 5 10 20
transient that followed w;~c; recorded by the computer.
The net heat input to the system was
evaluated several times from the combined effects
of the temperat.ure slope and thermal capacity of
i.he system, the power to the electric heaters, and
the 10 kw oE nuclear power when the reactor was
critical. The fuel was drained shortly after the
final heat input data were taken in the Euel loop
55.5 hr after the scram, Two additional sets of
heat input data were taken in the fuel drain tank
at times of 168.5 and 745 hr, but there was no ex-
perimetatal method to correlate the drain-tank data
with the fuel-loop data because of the difference
in heat losses. The analysis of the experimental
daiita gave the change in afterheat between the
55.5-hr data and the various other sets of data
taken in the fuel loop and between the two sets of
data in the fuel drain tank.
The computer program CALDRON was used to
check the experimental results and to provide
reference points at decay times of 55.5 and 765
hr. ‘The results of the CALDRON calculations
and the afterheat measurements are shown in Fig.
1.11. ’Two sets of CALDKQN calculations are
shown, one set without krypton or xenon stripping
and the other with krypton and xenon stripping at
a rate equivalent to the removal from the MSKE
fuel salt, The MSKE experimental data were
normalized to the 55.5- and 745-hr CALDRON cal-
culations that included stripping. (The heat
losses required to make the observations agree
with the calculation at these points were assumed
to exist at all other times in the same system.)
We had hoped to obtain useful data within about
10 min after the scram. However, there was ap-
25
TIMF AFTFR SUI lTTln\Alkl fhrl
ORNL - DWG 67-11787
1 MSRE MEASIIREMfNIS NOR
50 1GO 200 500 1000
Fig. 1.11. Results of MSRE Aftarheat Meosurernenr.
parently an air leak through the radiator enclosure
which healed itself in about 1.5 to 2 hr. Since
the calculation procedure required that the heat
losses froin the reactor system be nearly con-
stant, the first 2 hr of data could not be used.
Actually the calculations made 2 hr after shutdown
also appear to be somewhat low, as seen in Fig.
1.11.
The thermal capacity of the fuel and coolant
systems for the afterheat calculation was cali-
brated during the run 12 startup. The temperature
transient was recorded following a step increase
in nuclear power of 148 kw. The thermal capacity
of the system was found to be 16.22 Mw-sec/”F.
1.4 EQUIPMENT PERFORMANCE
Heat Transfer
C. W. Gabbard
The monitoring of the heat transfer performance
of the salt-to-salt heat exchanger continued, both
by periodic measurement of the heat transfer coef-
ficient and by practically continuous observation
of the “heat transfer index.” (The heat transfer
index is defined as the ratio of react or power to
the temperature difference between the fuel leav-
ing the core and the coolant leaving the radiator.)4
Six sets of data were taken for evaluation (if the
heat transfer coefficient during rutis 11 and 12.
41bid., p. 21.