18.08.2015   NUCLEAR CHEMICAL ENGINEERING

THE FISSION PROCESS

5.1 Fissile Materials

Table 2.8 lists capture and fission cross sections for the four nuclides fissile with thermal neutrons and gives the average number of neutrons produced per nuclide fissioned (n) and per

Table 2.8 Properties of fissile nuclides for 2200 m/s neutrons’*’

233 U

235 и

239 Pu

241 Pu

Cross sections, b

Fission Of

531.1

582.2

742.5

1009

Capture ac

47.7

98.6

268.8

368

Absorption oa

578.8

680.8

1011.3

1377

a = ajar

0.0898

0.169

0.362

0.3647

Neutrons produced

Per fission V

2.492

2.418

2.871

2.927

Per neutron absorbed rj

2.287

2.068

2.108

2.145

^From App. C.

neutron absorbed (»?). These properties are needed to calculate reactor neutron balances, evaluate fuel reactivity, and work out fuel cycles.