ORNL-4191 - the Molten Salt Energy Technologies Web Site

0 9 N L- DWG 67- 1 183 3
I
23 43 63 a3 40 3 42 3
LOOP POSIlION-DISIANCE FROM REACTOR TANK TO LOOP
CORE CFNTER (in)
Fig. 15.3. Nuclear Heat Generation in Molten-Snlt
Loop 2.
recovered during postirradiation disassembly of
the loop and from the fission product activity in
the final salt samples, as discussed below.
Calculatioiis involving activity of type 304
stainless steel monitor wires or of fission products
in salt samples required taking into account the
relative flux history for the particular isotopc as
described in the section on activity calculation.
Thermal neutron flux levels calculated from type
304 stainless steel monitor wires attached to
various regions on the outside of the loop were
(in units of neutrons sec- '>: core sliell,
front, 4.4 to 5.5 x cole shell, bottom (rear to
front), 1.7 to 3.8 x 1013; central well in core
graphite, 2.7 to 3.4 x core shell, rear, 1.4
to 1.9 x IOi3; gas separation tank, 1.4 to 2.6 x
10 13. Applying a calculated attenuation factor
of 0.6 and a fuel blackness factor of 0.8 to a
181
~ .
Table 15.2. Mean Neutron Flux in Salt as Calculated
Isotope
from Fission Product Activity
Flux Estimate
(neutrons cm-' sec-')
__.__.I_._
Sample 23 Sample 26
. . . ... .. . . . .. . . . . . ... . ..... .-..... .
30-y 137Cs (6.0%)
281-d 144Ce (5.6%)
65-d "2, (6.3%)
58.3-d 'lY (5.8%)
50.4-d "Sr (4.79%)
32.8-d 141Ce (6.Ouibj
12.8-d 140€3a (6.4%)
11.1-d 14'Nd (2.6%)
10'3 x 1013
0.82
0.93
0.65
1.14
0.61
0.64
0.27
0.59
0.90
1.05
0.96
0.58
0.76
0.70
0.28
0.66
weighted mean monitor flux of 2.4 x 1013 results
in an estimated mean flux available to the fuel of
1.15 10'3.
The flux was estimated on the basis of the
activity of vari-ous fission products in final salt
samples after accounting for the relative flux
history of the salt. Table 15.2 shows flux estimates
based on the activity of a number of isotopes
that might be expected to remain in the salt.
As is frequently the case, the flux estimates
based on fission product activity in salt samples
are somewhat lower than the values obtained in
other ways. Generally, chemical separations are
used prior to counting. Because of favorable halflives,
well-established constants, low neutron
cross sections, good separations, and less interference
from other isotopes in counting, ' 37Cs,
'44Ce, and "Zr are regarded as the more reliable
measures of fissions (or flux). The average of the
flux estimates for these three isotopes is 0.88 x
1013.
The results from the different methods of esti-
mating flux are compared in Table 15.3. The
value of 0.88 x IOi3 obtained from the fission
product activity is the most direct measure of
fissions in the fuel, and, therefore, it will be used
in estimates of the expected activity of other
fission products produced in the hop.
Evidence of corrosion was obtained from chemi-
cal analysis of salt samples withdrawn from the