International Reactor Dosimetry File 2002 - IAEA Publications

3. SELECTION OF CANDIDATE CROSS-SECTIONS
FOR IRDF-2002
E.M. Zsolnay, H.J. Nolthenius
As stated in Section 1, IRDF-2002 contains cross-section data for 66
dosimetry reactions along with their related uncertainty information [3.1]. These
data have been selected from the most recently available cross-section libraries and
new evaluations. The procedure for selecting the best quality data for IRDF-2002
began with detailed analyses of the contents of the cross-section files of interest.
Prior to the Technical Meeting on International Reactor Dosimetry File:
IRDF-2002 (held at the IAEA in Vienna from 27 to 29 August 2002), a supplementary
workshop on benchmarks took place at the 11th International
Symposium on Reactor Dosimetry, Brussels, 18–23 August 2002 [3.2].
Agreement was reached that only those cross-sections accompanied with
adequate uncertainty information in the form of covariance matrices would be
accepted for IRDF-2002. The primary basis for the selection of the crosssections
for IRDF-2002 was comparison of the data with the experimental
results obtained from four standard neutron fields (thermal Maxwellian, 1/E,
252 Cf fission and 14 MeV neutron field), taking into consideration the corresponding
uncertainty information.
Detailed analyses of the data were followed by comparisons of the
integral values of the candidate cross-sections with the experimental data
obtained in the above mentioned standard neutron fields. C/E values were
determined and evaluated, together with the corresponding uncertainty data.
The original cross-section information was available in the ENDF-6
format for all the libraries investigated. These data have been converted to a
SAND II type 640 group cross-section form. A neutron temperature of 300 K
and a ‘flat’ weighting spectrum were applied in the conversion procedure. All
the calculations for the cross-section and related uncertainty information were
performed using the 640 energy group structure.
The following sections contain details of the work outlined above, and the
results obtained.
3.1. ANALYSIS OF THE DATA FROM RECENT NATIONAL
REACTOR DOSIMETRY FILES AND NEW EVALUATIONS
As part of the procedure for updating IRDF-90, data in the reactor
dosimetry files JENDL/D-99 [3.3] and RRDF-98 [3.4], and new evaluations
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