MOX fuel is different, I beg to differ. Higher gap fission product migration/releases, etc. Burns different. If melted will release 100 times mores plutonium than standard LEU fuel rods. See NRC Safety Analysis for putting MOX in US reactor:
http://adamswebsearch2.nrc.gov/idmws/DocContent.dll?library=PU_ADAMS^pbntad01&LogonID=026baa2ac948650fffa957db74869764&id=040970215
Excerpt: Table 1: Nominal Unirradiated Fuel Isotopics, %
U.S. European Proposed
Isotope LEU MOX MOX LTA
wt% 234U / U 0.03 ---- ----
wt% 235U / U 3.2 0.24 - 0.72 0.35
wt% 236U / U 0.02 ---- ----
wt% 238U / U 96.75 92.77 95.28
wt% 238Pu / Pu ---- 0.88 - 2.40 0.05
wt% 239Pu / Pu ---- 53.8 - 68.2 90.0 - 95.0
wt% 240Pu / Pu ---- 22.3 - 27.3 5.0 - 9.0
wt% 241Pu / Pu ---- 5.38 - 9.66 1.0
wt% 242Pu / u ---- 2.85 - 7.59 0.1
wt% Pu / HM ---- 4.0 - 9.0 4.37
wt%Fissile / HM 3.2 3.65 - 5.25 #4.15
HM = Pu + U. May not sum to 100% due to rounding and ranges. Derived
from data in licensee submittal, ORNL/TM-2003/2 [Ref.1], NUREG/CR-0200 V1 [Ref.2]
The two MOX fuel types differ in that the relative concentrations of plutonium and uranium and
the distributions of their isotopes differ. Table 1 above compares the distribution of fissile and
non-fissile isotopes in typical LEU fuel, typical commercial reactor-grade MOX fuel, and the
proposed MOX LTAs. The differences in the initial fuel isotopics are potentially significant to
accident radiological consequence analyses since the distribution of fission products created
depends on the particular fissile material. If the fissile material is different, it follows that the
distribution of fission products may be different. For example, one atom of I-131 is created in
2.86 percent of all U-235 fissions, whereas one atom of I-131 is created in 3.86 percent of all
Pu-239 fissions. This is an illustrative example only in that the radionuclide inventory in the fuel
at the end of core life depends on more than fission yield. Nonetheless, this shift in the fission
product distribution needs to be evaluated for its impact on the previously calculated
radiological consequences of DBAs.
The LEU fuel is enriched in the U-235 isotope, an operation that occurs on a molecular scale
while the UO2 fuel is in the gaseous phase. This processing results in fuel pellets with a high
degree of homogeneity and uniform grain sizes. The proposed MOX LTA fuel will be
manufactured in a process that involves blending of UO2 and PuO2 powders to achieve the
desired Pu content. The MOX fuel pellets, therefore, are not as homogeneous as an LEU fuel
pellet. This difference in pellet structure has the potential to affect the diffusion of fission gases
through the fuel pellet and may impact the fraction of the pellet fission product inventory that is
in the fuel rod gap between the pellet outer surface and fuel clad inner surface (i.e., gap
fraction). It is generally understood that the fission gas release (FGR) rate for MOX fuel is
greater than that for LEU fuel, given comparable enrichments and burnups. This behavior is
primarily explained by the lower thermal conductivity of MOX fuel pellets that results in higher
fuel temperatures than in LEU rods. Since the gap fractions are an input to the analyses of
calculated doses from non-core melt DBAs, changes to the gap fractions associated with MOX
fuel need to be considered