Analysis With Uranium in Solution cont'd

DOE-STD-3007-93

where the effect of spatial shelf shielding of the U-238 absorption resonances dominates over

the shelf shielding of the low energy U-235 fission cross section. This effect is shown in

Figure 6 between a U-235 enrichment of 0.7% and 5.0%

2. The heterogeneous fuel geometry is assumed to have an optimum fuel rod diameter, optimum

spacing, and optimum water moderation and reflection, so that the heterogeneous K-eff starts

as high as possible. This is true for the heterogeneous case in Figure 6.

It also needs to be recognized that uranium and plutonium that is dissolved will be swept away

from the fuel rods, and will be diluted in the solution in the entire dissolver, i.e. the mixing rate

will be much greater than the dissolution rate of the uranium and plutonium from the EBR-II and

TRR fuel rods. This will result from the convective currents within the dissolver, as well as from

sparging from the bottom of the dissolver. Therefore, the fuel rods from the EBR-II bundle and

the TRR bundles will be critically safe during any stage of dissolution.

The criticality safety of TRR rods in the dissolver was shown in Ref. 1. All cases that were

considered obtained K < 0.95, except for the case listed in Table 2 of Ref. 1, where it was

assumed that the entire inner annulus was filled with TRR rods with optimized diameters and

spacing. Since this condition was unrealistically conservative, Ref. 1 was accepted as proving the

criticality safety of TRR dissolution. Of particular interest is that the K-eff decreased by 0.05

(Ref. 1, Table 2 vs. Table 4) when 400 grams of natural uranium containing 0.129 wt. % Pu-239

was dissolved per liter of solution in the dissolver. It would have decreased by even more if the

neutron absorption in the nitrogen in the UO2(NO3)2 in solution were also included in the

calculation. The equivalent U-235 enrichment for this calculation (from Table 1 of Ref. 1) was

0.705 + (2.0 x 0.129) = 0.9635 %. Therefore, the effect on K-eff of ignoring old fissile material,

with an equivalent enrichment < 0.9635 %, in the dissolver solution is conservative. Since, from

Table 1, the equivalent U-235 enrichment for the EBR-II rods and the TRR rods that are being

considered in this NCSE is only 0.77 % and 0.79 %, respectively, dissolved material from these

rods will also cause the K-eff of the EBR/TRR rod lattice to decrease. This is the result of the

low fissile enrichment of the EBR/TRR material that is being dissolved. Therefore, it is

conservative to ignore the effect of the dissolved EBR/TRR material on the criticality safety of the

system. This is true for EBR/TRR rods as they dissolve, and for a new batch of rods that are

placed into old solution, provided that the enrichment of the old fissile material in solution is <

0.9635 %. Fissile enrichments > 0.9635 % in solution are considered in Section 5.

Another very significant result in Ref. l (Table 6, case 7) is the case for TRR rods in inserts in the

dissolver, including the effect of absorption in the insert wall material. This case calculated K =

0.488, which shows the criticality safety margin that actually exists for TRR dissolution. Since, as

shown in Table 1, the equivalent U-235 enrichments for the EBR-II rods and for the TRR rods

are both lower than for previous TRR calculations in Ref. 1 (0.9635 % enriched), criticality safety

will be assured for loading and dissolving the EBR/TRR bundles in the dissolver.

According to Table 2 of ANSI-8.1 (Ref. 6), an infinite homogeneous solution of uranium nitrate

[UO2(NO3)2] in water, without a fuel rod lattice, will be critically safe as long as the U-235

enrichment is < 1.96 wt. %. The limiting U-235 enrichment is as high as this value (1.96 wt. %)

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