nureg/cr-6700 - Oak Ridge National Laboratory

Section 5
Radiation Shielding
The effect of assembly type (PWR or BWR) on the characteristics of high-burnup spent fuel was investigated
by regenerating the nuclide importance rankings for an 8 × 8 BWR fuel assembly that incorporated four
burnable poison rods, each containing 4.0-wt % Gd 2 O 3 . The results of the shielding importance rankings for
the moderate-enrichment and moderate-burnup fuel (3-wt % and 20-GWd/t) and higher-enrichment and
higher-burnup fuel (5-wt % and 70-GWd/t) are listed in Table 12. The BWR fuel results for 100-years
cooling are almost identical to those for the PWR fuel. After 5-years cooling, the rankings are affected by
the larger 60 Co contribution in the PWR assembly compared to the BWR assembly due to the significantly
larger cobalt impurity level associated with the PWR assembly model. In addition, 252 Cf appears in the list of
dominant BWR nuclides due to its larger production in the BWR assembly. A number of nuclides, most
notably 252 Cf, 248 Cm, and 246 Cm, were found to exhibit large relative changes (up to 60%) in their fractional
contributions to the total dose rate but have little impact on the total dose rate due to the generally low
importance of these nuclides. All other changes in nuclide composition between the assembly types resulted
in changes that were less than 1% for the cask models studied.
The most dramatic change in the relative radionuclide importance with burnup is the increasing importance
of 244 Cm to the total dose rate. At shorter cooling times, the dose rate is dominated by fission products and
the 60 Co activation product (note, however, that the trace level of 59 Co used in this study was highly
conservative). Contributions from 134 Cs and 154 Eu, both generated by fission product capture as opposed to
direct production from fission, generally increase with burnup, while 144 Pr exhibits a steady near-exponential
decrease in importance with increasing burnup due to the increasing contribution from other radionuclides.
At longer cooling times (>5 years) the contribution from 244 Cm becomes significant in high-burnup fuel.
The rapid increase in the 244 Cm component results in the decrease in the relative contribution of many fission
products and 60 Co.
A significant trend in the shielding results is the dramatic increase in the neutron source, and consequently
the neutron component of the total dose rate for higher-burnup spent fuel. The increase with burnup is
dependent on both the cooling time and the cask design. In high-burnup fuel, spontaneous-fission neutrons
generated primarily by 244 Cm dominate the neutron source for cooling times between about 2 and 50 years.
At the shorter cooling times, 242 Cm becomes increasingly important, and at 100 years contributions from
240 Pu (at low burnup) and 246 Cm (at high burnup) represent a significant fraction of the spontaneous-fission
neutron source and thus the neutron dose rate. At the longer cooling times (>50 years) the (α,n) component
of the neutron source increases and becomes significant only for relatively low-burnup fuel.
At short cooling times, the fission products 144 Pr, 106 Rh, 134 Cs, and activation product 60 Co dominate the total
dose rate. However, the impurity cobalt level assumed in the assembly structural material is likely many
times higher than that used in current assembly designs. Therefore, the fractional contribution from 60 Co is
likely significantly overestimated by this study when considering newer fuel assembly designs.
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