Fission energy, average table

In these very heavy atoms, spontaneous fission is often the predominant mode of decay, and it is also the process most liable to large errors in half-life prediction. Nix, in a letter of considerable practical interest, computes the total energy release in fission, the average number of neutrons per fission (v), and the neutron energy for a few superheavy nuclides. Some of these results are shown in Table 1 the inclusion of a plutonium isotope provides a comparison with a known actinide. 

In determining the biological hazard from spontaneous fission, it is assumed that each fission produces an average of three neutrons, each with an energy of 2 MeV. Estimated surface fluxes and dose rates are given in Table 8.18. 

The actual value of v for a given fissionable material can be obtained from experiment. In such measurements, even if the incident neutrons were all of some fixed energy, the number of neutrons produced by fission reactions would be a statistical quantity and would vary from one fission to the next. As a matter of fact, the actual number of neutrons released by any one reaction is of little interest in reactor physics, and one would certainly prefer an average value which could be assigned to a large number of such reactions. Measurements of this type have been made, and the presently acceptable values of v (the average number of neutrons per fission) for the three principal nuclear fuels are listed in Table 1.2. 

An important quantity in reactor kinetics is neutron lifetime, l, the average time between the release of a neutron in a fission reaction and its loss from the system by absorption or escape. For convenience of calculation, the lifetime in a thermal reactor may be divided into two parts (a) slowing down time, the mean time required for fission neutrons to slow down to thermal energies, and (b) diffusion time, the average time that thermal neutrons diffuse before being lost in some way. Table 3.6 provides values of these parameters for several different media. 

When averaged over all the modes of disintegration, the energy release from slow neutron fission of is approximately 205 MeV. The way in which this release is distributed between the various processes involved is shown in Table 2.1. 

As was illustrated earlier (Fig. 4.1), the value of vj = v/ I + a) for both and Pu is appreciably higher in a fast spectrum. The fission and capture cross sections themselves are of course much smaller at the higher energies (see, for example. Fig. 2.8 for U ). Table 11.1 gives values of the important parameters of the fissile isotopes averaged over a typical fast reactor spectrum. 

Table I indicates that the average number of neutrons that are released during fission is greater than 2, although it is dependent on the energy of the neutron causing fission. Of these two or more neutrons one will have to be used to maintain the chain reaction, the remaining one or more can be used for (1) leakage from the reactor core, (2) capture in structure and moderator materials, and (3) production of fissile atoms from fertile atoms.

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