Suppose a pressurized water reactor is loaded with 200 tons of 3% enriched uranium. The reactor has been designed to run with an average thermal-neutron flux of ??_n??= 1.5?10¹³ cm^?2 s^?1 for a period of four years before refueling. You can assume that ?_n is held constant the whole time, ignore neutrons losses due to the finite size of the reactor, and ignore any contribution of ²³⁹Pu fission to the reactor power. What is the thermal power output of the reactor immediately after it is started? How does the power output decrease as a function of time over the four years? Use thermal-neutron cross sections from?

Table 19.1.?

Os Nucleus Density g/cm3 of 232 Th 233U 235 U 238 U 11.7 < 10-6 7.34 7.34 13.0 18.7 531 45.3 576 12.2 18.7 585 98.7 684 15.1 18.9 1.68 x 10-5 2.68 2.68 9.30 U* 18.9 4.21 3.37 7.59 9.34 natural 239Pu 19.8 748 271 1019 7.99