International Reactor Dosimetry File 2002 - IAEA Publications

cross-sections in this manner, the very fine group cross-sections can then be
collapsed to coarser group structures, which may be used in the neutron
adjustment codes. Alternatively, computer codes have been developed to
perform calculations of the neutron self-shielding corrections using the neutron
resonance parameters directly. Although this process may be time consuming,
the set of shielded activation cross-sections can be used routinely as long as the
same geometry foils or wires are used, regardless of the application.
II.3. TABLES OF NEUTRON SELF-SHIELDING CORRECTIONS
Neutron self-shielding correction factors have been experimentally
determined by irradiating foils or wires of varying thickness, with and without
cadmium covers. The relevant tables can be found in the literature, for example
in ASTM Standard Test Method E262 for Determining Thermal Neutron
Reaction and Fluence Rates by Radioactivation Techniques. However, these
tables list the neutron self-shielding corrections separately for the thermal and
resonance integrals of only a few types of material, including cobalt, gold and
indium. While such data can be used to determine corrections to very simple
reactor dosimetry experiments involving only these types of material or for
estimating the magnitude of such corrections, they are not generally applicable
to reactor dosimetry applications. Furthermore, this approach is not
appropriate to neutron spectral adjustment procedures, since the tabulated
data can only be used to correct reaction rates prior to spectral adjustment. The
neutron cross-sections are preferably shielded rather than the reaction rates.
Spectral adjustments will then not depend on any prior assumptions concerning
the thermal or epithermal neutron flux.
II.4. APPROXIMATION FORMULAS COMMONLY USED FOR
NEUTRON SELF-SHIELDING
Reactor dosimetry measurements are frequently performed with
relatively small foils or wires of such a size that the mean free path for neutron
scattering tends to be larger than the sample dimensions. Therefore, the
neutron self-shielding factor can be approximated by neglecting the neutron
scattering effects. Formulas can then be derived to determine the neutron selfshielding
for any given geometry, assuming an isotropic neutron flux. Such an
assumption is generally acceptable for the thermal and epithermal neutron
flux, and the derivation of such formulas is given in Refs [II.2, II.3]. Neutron
self-shielding calculations should be performed for each neutron cross-section
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