To illustrate the characteristics of different methods of fuel management, the example of the 1054-MWe PWR designed by the Westinghouse Electric Company for the Donald C. Cook Nuclear Plant of the American Electric Power System [A1 ] has been used.

We shall be interested in estimating the fuel-cycle performance of this reactor during a steady-state cycle, in which fresh fuel has a 235 U enrichment of 3.2 w/o 235 U and spent fuel has a 23SU enrichment of around 1.0 w/o and also contains around 0.6 w/o fissile plutonium. The reference design condition used to evaluate effective neutron cross sections and other reactor physics parameters during irradiation is taken to be 2.7 w/o 235 U. This value, slightly higher than the arithmetic mean of the fissile content of fuel at the beginning and end of irradiation, is intended to reflect the higher cross section of fissile plutonium compared with 235 U.

Table 3.11 gives the properties of each region of the core lattice for the reference design of this reactor. Information to be used directly in neutron balances and other fuel-cycle calculations are the volume fraction of each region v, the concentration N of each molecular or nuclear species in that region, and the ratio of the thermal flux in that region to the thermal flux in the fuel region ф. The rate of reaction of thermal neutrons with species ; in region j per unit lattice volume is Ща^ф/ф, where

Ny is the number of atoms or molecules of species / per unit volume of region /

Oj is the effective cross section of species і for thermal neutrons Vj is the volume fraction of region j in the lattice

ф/ is the ratio of thermal-neutron flux in region / to thermal flux in the fuel ф is the thermal-neutron flux in the feel region

Characteristics assumed for this reactor in the reference design condition are listed in Table 3.12.

Table 3.13 gives effective cross sections for thermal neutrons and other nuclear properties of the materials in the core of this reactor. These effective cross sections have been calculated by the procedure recommended by Westcott, which has been outlined in Chap. 2, from data provided by Westcott [W3] and Critoph [С1]. To obtain appropriate nuclear reaction rates, these effective cross sections are to be multiplied by the thermal-neutron flux, ямвіі, where Имв is the density of neutrons in the Maxwell-Boltzmann part of the spectrum and 0 is the average speed of the Maxwell-Boltzmann neutrons.

(3.37)

The value of 80 b given for fission-product pairs is an approximate, constant value to be used independent of the fuel from which the fission products are formed and independent of the flux time to which the fission products are exposed after formation. For the present PWR, the effect of these variables on the cross sections of fission-product pairs, evaluated by extrapolation of Walker’s [Wl] tables, is as follows:

As irradiation progresses, the effective cross section of fission products from each fuel nuclide decreases. This decrease is partially offset by greater production of fission products from plutonium, which have a higher cross section than those from 235 U. The constant value of aF = 80 given in Table 3.13 takes these two effects approximately into account.

Table 3.14 gives the neutron balance for this reactor in the reference design condition, charged with U02 fuel containing 2.7 w/o 435 U uniformly distributed, operating at full power of 3250 MW, and with equilibrium xenon and samarium. Items 1 through 5 deal with events experienced by neutrons with energies high enough to cause fission in 338 U, above 1 MeV. Items 6 through 10 deal with events experienced by neutrons while being slowed down from fission to thermal energies. Items 11 through 23 give the production and consumption of thermal neutrons.

One significant feature of Table 3.14 is that the production of thermal neutrons, item 11, equals the total consumption, item 23. A second point to be noted is that the reactivity is given by the ratio of item 22 to item 23.

0 2112

p = ud = 0A349 (338)

A third point is that the constant term К in Eq. (3.33) is the sum of items 12, 15, 16, 17, 18, 19, and 20.

К = 0.0008 + 0.1431 + 0.0390 + 0.0043 + 0.0216 + 0.0724 + 0.0143 = 0.2955 (3.39)

Finally, the initial conversion ratio ICR, the ratio of the atoms of 239 Pu produced per atom of 235 U consumed, is the sum of items 2, 8, and 15.

ICR = 0.0111 + 0.4619 + 0.1431 = 0.6161 (3.40)

These terms determine the rate at which plutonium builds up during irradiation. One reason for giving the neutron balance in such detail is to be able to evaluate these terms.