International Reactor Dosimetry File 2002 - IAEA Publications

from ENDF/B-VI Release 8, JEFF-3.0 and CENDL-2 [3.5], were assessed and
analysed.
Analysis began with a survey of the plots of the relevant cross-sections in
order to detect discontinuities and other obvious discrepancies in the crosssection
data. The numerical characterization of the cross-sections of interest
required that the spectrum averaged cross-section values be calculated for
three theoretical spectrum functions (Maxwellian thermal spectrum at a
neutron temperature of 293.58 K, 1/E spectrum from 0.5 eV to 1.05 MeV and
Watt fission spectrum). A three group structure was used for the representation
of the uncertainty information, with energy boundaries of 10 –4 eV, 0.5 eV,
1.05 MeV and 20 MeV. A typical materials testing reactor (MTR) spectrum
available in 640 SAND II group format [3.6] (Fig. 3.1) was used as a weighting
spectrum in the input of the cross-section uncertainty processing code.
Cross-section values and the related uncertainty information were investigated
(including detailed analyses of the relevant covariance matrices).
Corresponding data from the different libraries were compared, along with the
equivalent data of IRDF-90. The results, together with the detected errors,
discrepancies and shortcomings (which could be related to the physics and/or
mathematics content, or to the format of the data), were presented in the form
of progress reports [3.7, 3.8] and communicated to the evaluators of the
libraries via the IAEA. Some 180 different cross-sections were analysed (some
several times due to revisions (see below)). For several reactions, no better
quality cross-section evaluations are available in the literature than the data in
IRDF-90. Only a limited number of new evaluations accompanied by
uncertainty information (the majority of them for the RRDF) have been made
in the energy region from thermal to 20 MeV over the previous decade.
As a result of the analysis outlined above [3.7], the evaluators revised and
modified selected data from JENDL/D-99 and RRDF-98, and a number of new
cross-section evaluations have been included from Refs [3.9, 3.10].
Examination of the revised data and analyses of the new data [3.8] led to the
preparation of a new set of cross-sections. These cross-sections were candidates
for inclusion in IRDF-2002, and are listed in Table 3.1 [3.8, 3.11].
The cross-sections and their uncertainty information (as listed in
Table 3.1) were the best quality data available in the literature before the end
of 2004, and therefore the cross-section data for IRDF-2002 are taken from
these sources. There are some reactions that are of interest for dosimetry applications
with insufficient cross-section information, while no suitable crosssection
data were found in the literature for others. These reactions are also
listed in Table 3.1.
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