Thermal equivalent neutron flux

In order to estimate the induced activity, the neutron dose value and the effect of interfering reactions, the distributions of thermal, epithermal, and fast neutron flux densities should be determined in a bulk media or a phantom. The most commonly used dosimetry reactions are given in O Sect. 36.3 of the Appendix of this Volume. In the case of a phantom, the combination of the activation detectors and the physical integration methods is recommended for the determination of the volume averaged dose equivalent rates and through it the total dose absorbed by a given organ (Csikai et al. 1988). 

The average spatial thermal flux is actually the total neutron density times 2200 m/sec. This is true because the absolute thermal flux is measured with l/v-absorber, using its cross section at this standard velocity. The macroscopic fission cross section of U is that for the same velocity. Actually, since this cross section does not have a l/v-dependence, a correction can be made to make it an equivalent l/v-cross section, which will give the correct fission rate in Eq. (4). This correction is based on a Maxwellian neutron distribution and is a function of the neutron temperature. The temperature of the neutrons in the AGN-201 core is approximately 350 K, which corresponds to a correction factor of 0.98. 

Equations (l), (2) and (3) are correct only if each atom of the sample is exposed to the same thermal flux. If the saimple is in the usuaj form of a foil which has appreciable thickness, the thermal flux is decreased inside the foil due to the absorption of thermal neutrons as they traverse the foil. This self-shielding effect will be compensated for in computing the equivalent activity the foil would have had if the thermal flux were not depleted by the self-absorption of the foil. 

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