**Acknowledgements**

**Step Time (days) Flux (n/cm2**

52 Nuclear Material Performance

**Table 6.** Neutronics and burnup data for UOX fuel.

**Step Time (days) Flux (n/cm2**

**Table 7.** Neutronics and burnup data for MOX fuel.

**Step Time (days) Flux (n/cm2**

**Table 8.** Neutronics and burnup data for CEU fuel.

**4. Conclusions**

 0 2.06 × 1014 0 30 2.17 × 1014 1.16 60 2.20 × 1014 2.32 70 0 2.32 100 2.23 × 1014 3.47 130 2.26 × 1014 4.63 160 2.31 × 1014 5.79 190 2.35 × 1014 6.95 220 2.40 × 1014 8.10

7 190 2.81 × 1014 12.00 8 220 2.91 × 1014 14.00

 0 1.85 × 1014 0 30 1.96 × 1014 2.05 60 2.00 × 1014 4.11 70 0 4.11 100 2.05 × 1014 6.16 130 2.09 × 1014 8.21 160 2.14 × 1014 1.03 190 2.19 × 1014 1.23 220 2.25 × 1014 1.44

**-s) Burnup (GWD/MTU)**

**-s) Burnup (GWD/MTU)**

**-s) Burnup (GWD/MTU)**

Parallel computing of fuel burnup in nuclear reactors give important insight about the chain reaction, the fuel consumption and buildup of radionuclides. The comprehensive calculations clearly present the characterization of the neutron flux, which is essential in the reactor design. This work is supported by the National Science Foundation (NSF) through the Center for Energy and Environmental Sustainability (CEES), a NSF CREST Center (Award NO. 1036593).
