The alternative to Section 4.4.1 of this chapter — fuel enrichment and fast neutrons — is to retain natural uranium as the fuel but combined with slow neutrons, and thus take advantage of the greatly increased likelihood of U-235 fission occurring. Figure

1.6 (c) shows that the value of ar is several hundred times greater for slow neutrons than for fast.

Instead of the П] fast neutrons of Section 4.3, let n ] slow neutrons be introduced into the infinite mass of natural uranium. The calculation in Section 4.3.4 of this chapter to obtain the next generation of П2 neutrons may now be repeated, but of course there is no possibility with slow neutrons of inelastic collisions occurring. The neutrons are simply either captured or cause fission.

The necessary microscopic cross-section values for natural uranium may be obtained by taking the weighted average of the values for U-235 and U-238:

a (natural uranium) =

1 138

_ a (U-235) + — a (U-238)

139 139

The values given in Table 1.2 are applicable to neutrons of the ‘standard’ thermal energy 0.025 eV. Thus:

П2 = Оf nj/(<7f + Oc)v

= 4.15 П|/(4.15 + 3.43) 2.43 = 1.33 П|

Therefore к® = 1.33 > 1.

Hence natural uranium fuel will give k® > 1, theo­retical maximum value 1.33, provided the fissioning neutrons are of thermal energy. This is the basis of the class of reactors known as thermal reactors, per­haps with some slight enrichment of the fuel.

The flaw in the foregoing is that, following the absorption of the ni thermal neutrons, the next generation of П2 neutrons are as a result of fission events and are therefore fast neutrons. These must be converted — moderated — into thermal neutrons if the chain reaction in the natural uranium is to be sustained. This is achieved by having a second material, a moderator, in the reactor in which the neutrons released by fission may lose their energy and become thermal neutrons by successive elastic collisions with the moderator nuclei.