The FuTure oF nuclear Fuel cycle - MIT Energy Initiative

Table 6.3 HLW in the Fuel Cycle Options
hlW
onCe-throuGh
CyCle
tWiCe-throuGh
SCheme (lWr-mox)
SCenario
FaSt reaCtorS
SCheme (lWr-Fr)
tWo-tier SCheme
(mox Fr)
Spent uo 2 fuel na na na
FP in spent uo 2 fuel
na
Ma in spent uo 2 fuel na na na
MoX fuel fabrication losses na na
Spent uo 2 /MoX fuel reprocessing losses
na
Spent MoX fuel na na na
Fr fuel fabrication losses na na
FP in spent Fr fuel na na
Fr spent fuel reprocessing losses na na
Spent Fr fuel na na na na
As high-level wastes have various levels of decay heat and radiotoxicity, both varying over
time, one cannot simply aggregate their masses to make comparisons among the scenarios.
However, “Densification factors” may be used to aggregate the different types of wastes in
order to compare the total repository requirements. A densification factor can be defined as
[BCG, 2006]: “the quantity of HLW or used fuel that can be disposed per unit length of (disposal
drift in) Yucca Mountain is … referred to as the “drift loading factor” and is expressed
in MTHM/m YM . …The densification factor is the ratio of the drift loading factor of HLW to
the drift loading factor of used fuel 4 ”.
Two potential constraints are considered: volume and heat (taking the waste package into
account). In all studies, a total cooling time of 25 years prior to disposal in repository is assumed,
and the repository is assumed to be ventilated for 75 years after it is fully loaded. The
values of the densification factors are also sensitive to the assumptions about the burnup,
the cooling time before reprocessing (e.g the build-up of 241 Am from 241 Pu decay drives up
long-term heat) and above all to the amounts of TRU, cesium and strontium remaining in
the spent fuel, as shown by Figure 6.4 [Wigeland, 2006]. In this figure, the impact on the
repository capacity for waste storage of removal of Cs and Sr as well as transuranic elements
(Pu, Am and Cm) is shown. The cumulative densification factor is somewhat correlated
with repository costs, as discussed in Chapter 7. 5
In CAFCA, it is assumed that 99.9% w of the Pu or TRU (depending on the scenario) is removed
from spent fuel during the reprocessing process 6 . The fission products are assumed
to remain in the waste. However, various densification factors can be found in the literature,
for both the Pu (or TRu) and fissions Products (FP):
1. The densification factor found in [BCG, 2006] for the FP/MA mix resulting from spent
UO 2 fuel reprocessing in the TTC scenario is ~ 4.
2. The [Shropshire et al., 2009] study suggested densification factors from 2 to 10 for the
FPs alone (separated from spent UO 2 fuel), with an effective value of 2.5. This value is
more pessimistic about the benefit of separations than the [BCG, 2006] estimation. In
chapter 6: analysis of Fuel cycle options 79