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since that time by addition of 84 g of beryllium
metal. The method of addition was described
previously.
The low corrosion sustained by the MSIiE fuel
circuit, which is in general accord with the re-
sults from a wide variety of out-of-pile corrosion
tests, might have been expected to be greater
during the first 10,000 Mwhr of operations be-
cause the UF, concentration of the fuel was
markedly less than was intended. According to
Grimes,’ “The lack of corrosion in the MSRE
melts which appear to be more oxidizing than
intended can be rationalized by the assumption
(1) that the Hastelloy N has been depleted in Cr
(and Fe) dt the surface so that only Mo and Ni
are exposed to attack, with Cr (and Fe) reacting
only at the slow rate at which it is furnished to
the surface by diffusion, or (2) that the noble-
metal fission products are forming an adherent
and protective plate on the reactor metal.”
If it is assumed that corrosion of the fuel pro
duced 3.3 equivalents of UF, and that fission has
resulted in the oxidation of 0.8 equivalent of UF,
‘R. E. ‘l‘homa and W. R. Grimtxs, MSR Program
Semznnn. Pro&. Rept. Feb. 28, 1967, ORNL-4119,
p. 123.
’W. R. Grimes, Chemical Research and Develop-
ment for Molten-Salt Breeder Reactors, ORNL-~~S~~
p. 70 (June 6, 1967).
Sample
No.
9-4
10-25
11-5
11-13
11-32
11-38
11-49
12-6
12-1 1
12-21
111
per gram-atom of fissioned uranium, ’ the addi-
tions of beryllium metal to the fuel have been fol-
lowed by the concentrations listed in Table 8.3.
The calculated values are for the most part higher
than the measured values. The cause of this
disparity is not currently understood, but is under
investigation.
All dissolutions of beryllium metal into the fuel
salt proceeded smoothly; the bar stock which was
withdrawn after exposure to the fuel was observed
to be smooth and of symmetrically reduced shape.
No significant effects on reactivity were ob-
served during or following: the beryllium additions,
nor were chemical analyses indicative that such
additions were made until after run 12 was begun.
The additions preceding run 12 had increased the
U3+ concentration in the total uranium by ap-
proximately 0.6%. Four exposures of beryllium
were made at close intervals during the early
part of that run. Samples taken shortly after the
last of these four exposures (FP12-16 et seq.,
Table 8.1) began to show an unprecedented in-
crease in the concentration of chromium in the
specimens, followed by a similar decrease during
the subsequent sampling period.
Previous laboratory experience has not dis-
closed comparable behavior, and no well-defined
‘Ibid., p. 65.
Table 8.3. Concentration of UF3 in the MSRE Fuel SaltR
___I____
Uranium Uranium + u3 Net
u3 +/&J u3 +/xu
iLlwhr Burned Burned Oxidized He Added ne Added Equivalents Calculated Analysis
(9) (moles) (moles) (‘) Reduced
(7%) (X)
10,978
16,450
17,743
20,386
25,510
27,065
30,000
32,450
33,095
35,649
5.54
829
953
1029
1287
1365
1514
1637
1670
1798
2.34 1.87
3.50 2.80
4.02 3.20
4.34 3.46
5.43 4.34
5.76 4.60
6.39 5.10
6.91 5.50
7.05 5.60
7.59 6.10
0 0
l(i.28 3.61
16.28 3.61
27.04 6.20
27.94 6.20
27.94 6.20
36.34 8.06
36.34 8. Oh
54.11 12.01
74.12 16.45
3.13
5.81
5.21
8.74
6. 86
6.60
7.96
7.76
11.40
15.35
RThese numbers assume that the salt originally was 0.16% reduced; that the increase in Cr from 38 to 65 ppm was
real, occurred before 11-14-66, and resulted in reduction of U“’ to U”; that each fission lesults in oxidation of 0.8
atom of U”; and that there have been no other losses of 1J3+.
0.31
0.58
0.53
0.88
0.69
0.66
0.80
0.77
1.14
I .5
0.1
0.5
0.37
0.42
0.34
0.37
1.2
0.5