International Reactor Dosimetry File 2002 - IAEA Publications

Such approximate calculations are not necessarily applicable to all
material types used for a specific reactor dosimetry application. The adequacy
of the approximation may also be tested by using samples of different
thickness, or comparing the results from dilute and non-dilute types of
material. There are also a number of other effects that can have a significant
impact on the calculation of activation rates, such as flux depression or the
partial shielding of one sample by an adjacent sample. Such effects may not add
linearly, especially when neutron scattering effects are significant.
II.1. USEFUL APPROXIMATIONS FOR COVER FOILS
The attenuation of neutrons in a neutron beam is given simply by the
equation:
F = exp(–x)
where x = Nst, s is the total neutron absorption cross-section and t is the
thickness of the cover foil. This equation can be integrated over all angles for
an isotropic neutron flux:
F = E 2 (x)
where E 2 is the second exponential integral. Such a correction can then be
applied to the neutron cross-section for any given activation reaction in each
neutron energy group.
II.2. NEUTRON GROUP STRUCTURES FOR COVER OR
SELF-SHIELDING CORRECTIONS
Since many neutron resonances are very narrow in width, a computer
code such as LINEAR is required to process the point cross-sections so that the
narrow neutron resonances will be adequately represented. Using a fixed
group structure (such as the 640 groups in IRDF-2002) may not be adequate
for reactions with narrow resonance structures. The cover and self-shielding
calculations should be performed for each neutron energy group in the fine
structure that results from the LINEAR processing code. Doppler broadening
must also be taken into account at the temperature of the reactor experiments,
using computer codes in the PREPRO2002 library available on the IAEA web
site [II.1]. After the neutron self-shielding corrections have been applied to the
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