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Radial neutron flux profile: Figure 3.11 shows the 2 dimensional radial neutron flux profile for TRU fueled core. The flux profile is similar to that of UOX and MOX cores.

Figure 3.11. 2D-radial neutron flux profile for TRU-1/25-4.5/40-4.2/24- TRU.

The flux is again uniformly distributed throughout the core with the maximum flux at the central fuel assembly. The maximum to average flux ratio was calculated as 2.535. Reactor physics parameter: The delayed neutron fraction for TRU fuel was calculated as 0.004822 as shown in Table 3.9 which is similar to MOX fuel due to the presence of Pu239 in the TRU fuel assemblies.

The control worth calculated for TRU core is calculated as 0.23961. It is observed the control rod worth or the shutdown margin for the TRU fuel is similar to that of MOX fuel. This is because the higher plutonium content in the TRU fuel which also exhibits a higher thermal neutron absorption cross section thus resulting in the decrease of neutron absorption by control rods or burnable poison.

The reactivity co-efficient values for fuel and moderator for TRU core are -1.43E- 05 δk/o_{C and -1.135E-03 δk/}o_{C as shown in Table 3.8. Compared to UOX fuel the reactivity} co-efficient for TRU cores are also less negative. This is again because of low capture resonance and high absorption resonance for Pu239 _{which results in positive reactivity thus} making the reactivity co-efficient less negative.

Table 3.9. Reactor physics parameters for cores with TRU fuel arrangement.

Details Core 1 : TRU-1 /25-4.5/40-4.2 /24- TRU

Multiplication Factor 1.21939 ± 0.00012

Maximum to Average Flux Ratio 2.535

Delayed Neutron Fraction _0.004822

Control Rod Worth 0.23961

Fuel Co-efficient of Reactivity

(δk/oC) -1.430E-05

Moderator Co-efficient of

Reactivity (δk/oC) -1.135E-03

Critical boron concentration: Figure 3.12 shows the soluble boron (B10_{) concentration in} ppm corresponding to the keff values for TRU fueled cores. The critical boron concentration for reactor criticality for TRU core is 1547.807 ppm. It is observed that the critical boron concentration for TRU core is greater than that to the UOX and MOX cores to overcome the presence and formation of plutonium isotopes and the long lived fission products (i.e. trans-uranic actinides) in the fuel assembly.

Figure 3.12. The k-effective vs boron (B10) concentration for TRU fueled core.

Equilibrium cycle and spent nuclear fuel analysis: Figure 3.13 shows the multiplication factor variation for TRU fueled core for once, twice and equilibrium burn cycle.

The burn characteristic for the TRU fueled core stands in good agreement with the UOX fueled cores. From the figure, the cycle length for the TRU core is 1000 days (~33 months) for once burn cycle, 500 days (~16 months) for twice burn cycle and 360 days (12 months) for equilibrium cycle.

Table 3.10 shows the fuel utilization at the end of equilibrium cycle in the TRU fuel assembly, whereas Table 3.11 shows the non-actinide composition in the spent fuel for the TRU fueled core. The burnup analysis for the TRU fuel is also different compared to UOX and MOX cores in terms of its actinide composition.

Table 3.10. Initial and final actinide composition for the TRU fuel assembly.

Fission Products Initial Composition Final Composition

Mass (kgs) Mass (kgs) Np237 _{10.36 2.25} Pu238 _{2.941 7.842} Pu239 135 4.303 Pu240 42.3 13.39 Pu241 14.34 9.254 Pu242 7.703 17.53 Am241 9.916 1.079 Am242 0.00669 0.00667 Am243 1.297 5.291 Cm242 0.00002 0.6154 Cm243 0.00283 0.04654 Cm244 _{0.188 5.671} Cm245 0.1134 0.4734 Cm246 0.0009 0.2623

Just like the MOX assembly, the high initial Pu240 content decreases the actinide composition in the TRU fuel and also enhances the Am241 _{poisoning. It is observed from} the table that the total Am241 consumption at the end of equilibrium cycle is more than 89 percent. This is important in criticality analysis for storage and disposal of spent fuel due to high radioactivity effects of Am241 and Np237 formed due to its α-decay which induces a large negative reactivity in the spent fuel. The overall Pu239 consumption of more than 94 percent. The total consumption for the Am and Np consumption was 43 percent and 78 percent respectively. These values were in good agreement with the reference values based on the burnup characteristics for inert matrix TRU fuels.

Table 3.11. Total non-actinide composition for the TRU fueled core.

Fission Products TRU-1 /25-4.5/40-4.2 /24- TRU Mass (kgs) Mo95 13.07 Tc99 18.34 Ru101 19.17 Rh105 11.93 Ag109 2.379 Cs133 26.51 Nd143 18.54 Nd145 _15.49 Sm147 _1.51 Sm149 _0.08416 Sm150 6.353 Sm151 0.3163 Sm152 2.59 Gd155 1.904 Gd155 0.001301 Total 138.1878

The burnup characteristics for the fission products (i.e. non-actinides) in the TRU cores is similar to that for MOX cores. This is because of high initial plutonium content particularly for Pu239, Pu240 and Pu241 isotopes in the TRU fuel assemblies. The total composition of the fission products as well for important individual isotopes were in good agreement with that for MOX assemblies.

4. CONCLUSION

The Westinghouse SMR was modelled and analyzed with flexible fuel configurations using the Monte Carlo N-Particle code. The fuels investigated in the study were uranium-oxide, mixed-oxide and trans-uranic fuels. First, the SMR core was referenced with the UOX fuel assembly; and on confirming good performance it was then used for the comparative study with other fuel options. The study was carried out by determining the reactor physics parameters (i.e. the radial flux profile and the delayed neutron fraction) and the reactor safety related parameters (i.e. the temperature co-efficient, the soluble boron and control rod worth) at BOL. Furthermore, the spent fuel composition for each of these oxide fueled cores was also analyzed.

For the UOX fueled cores, the parameters were in good agreement with that to the values for a standard pressurized water reactor with UO2 fuel. The radial neutron flux profile was uniformly distributed throughout the core. The control rod worth were sufficient to safely shutdown the reactor during normal as well as cold operating conditions. Furthermore, the soluble boron concentration was found to be within permissible limits thus controlling the criticality during reactor operation without resulting in corrosive damage to the reactor components. The reactivity co-efficient values for the fuel and moderator were also largely negative indicating a large negative feedback for any positive reactivity insertion thus controlling and safely shutting down the reactor.

Upon successful referencing the UOX fueled SMR, the core was analyzed with MOX and TRU fuels using the same procedure. Comparatively, the reactor physics parameter as well as the reactor safety parameters were found to be in good agreement with that to the UOX core. The delayed neutron fraction for the MOX and TRU fuel was lower

due to high initial plutonium content. Furthermore, the control rod and absorber worth was increased whereas the reactivity co-efficient were less negative. This was because of neutron spectrum hardening behavior exhibited by plutonium isotopes which decreased the neutron absorption by control rods and soluble boron thereby leading to positive reactivity insertion and reducing their worth.

The equilibrium cycle for each cores were also determined using a three batch refueling strategy. For the UOX cores, the equilibrium cycle length was 14 to 17 months whereas for MOX and TRU cores were 14 to 15 months and 12 months respectively. The 24 month refueling defined in the objective can be achieved by adopting a different core arrangement or by adjusting the boron concentration in the burnable absorbers or by increasing the fissile loading in the core. In the spent fuel, the composition for fissile material, the long lived and short lived isotopes were determined. This is highly recommended for the criticality analysis required for the storage of spent fuel which as a part of the future works to be conducted. In the MOX fueled cores, the reactor grade Pu consumption was 54 percent whereas weapon grade Pu consumption was 74 percent. The presence of high Pu20 content enhanced the Am241 poisoning but below the 3 percent safety criteria thus making it possible to separate and the recycle the spent fuel for MOX fuel fabrication. In TRU fuel, with the IMF based fuel Pu239 consumption of 94 percent was achieved whereas the total Am and Np consumption were 43 percent and 78 percent respectively.