ISO 23018:2022
(Main)Group-averaged neutron and gamma-ray cross sections for radiation protection and shielding calculations for nuclear reactors
Group-averaged neutron and gamma-ray cross sections for radiation protection and shielding calculations for nuclear reactors
This document provides guidance in the preparation, verification, and validation of group-averaged neutron and gamma-ray cross sections for the energy range and materials of importance in radiation protection and shielding calculations for nuclear reactors[1], see also Annex A. [1] This edition is based on ANSI/ANS-6.1.2-2013[1].
Sections efficaces multigroupes neutrons et gammas pour les calculs de radioprotection associés aux réacteurs nucléaires
General Information
Standards Content (Sample)
INTERNATIONAL ISO
STANDARD 23018
First edition
2022-05
Group-averaged neutron and gamma-
ray cross sections for radiation
protection and shielding calculations
for nuclear reactors
Sections efficaces multigroupes neutrons et gammas pour les calculs
de radioprotection associés aux réacteurs nucléaires
Reference number
ISO 23018:2022(E)
© ISO 2022
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ISO 23018:2022(E)
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© ISO 2022
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ISO 23018:2022(E)
Contents Page
Foreword .iv
1 Scope . 1
2 Normative references . 1
3 Terms and definitions . 1
4 Abbreviations and acronyms . .2
5 Preparation of group-averaged neutron and gamma-ray cross sections .3
5.1 Evaluated nuclear data files. 3
5.2 Checking of evaluated nuclear data files . 3
5.3 Energy ranges and materials of importance . 3
5.4 Group-averaging techniques . 3
5.4.1 General . 3
5.4.2 Fine-group structures . 3
5.4.3 Pointwise weighting function . 3
5.4.4 Self-shielding treatment . 4
5.4.5 Collapsed-group cross sections . 4
5.5 Upscattering cross sections . 4
5.6 Legendre order of scattering . 4
6 Verification and validation of cross sections . 4
6.1 Verification of cross-section sets . 4
6.2 Validation of cross-section sets . 5
Annex A (informative) Information on group-averaged neutron and gamma-ray cross
section verification, validation, libraries, and generation . 6
Bibliography . 9
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ISO 23018:2022(E)
Foreword
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This document was prepared by Technical Committee ISO/TC 85, Nuclear energy, nuclear technologies,
and radiological protection, Subcommittee SC 6, Reactor technology.
Any feedback or questions on this document should be directed to the user’s national standards body. A
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INTERNATIONAL STANDARD ISO 23018:2022(E)
Group-averaged neutron and gamma-ray cross sections for
radiation protection and shielding calculations for nuclear
reactors
1 Scope
This document provides guidance in the preparation, verification, and validation of group-averaged
neutron and gamma-ray cross sections for the energy range and materials of importance in radiation
1)
protection and shielding calculations for nuclear reactors , see also Annex A.
2 Normative references
The following documents are referred to in the text in such a way that some or all of their content
constitutes requirements of this document. For dated references, only the edition cited applies. For
undated references, the latest edition of the referenced document (including any amendments) applies.
ISO 12749-5, Nuclear energy, nuclear technologies, and radiological protection — Vocabulary — Part 5:
Nuclear reactors
3 Terms and definitions
For the purposes of this document, the terms and definitions given in ISO 12749-5 and the following
apply.
ISO and IEC maintain terminological databases for use in standardization at the following addresses:
— ISO Online browsing platform: available at https:// www .iso .org/ obp
— IEC Electropedia: available at https:// www .electropedia .org/
3.1
cross-section processing code
computer code that converts evaluated nuclear data in a specified format and procedure into a form
that is appropriate for use in applications
Note 1 to entry: A cross-section processing code performs calculations such as resonance reconstruction,
Doppler broadening, and multigroup averaging.
3.2
ENDF/B
United States of America evaluated nuclear data file (3.3) prepared and reviewed by subject matter
experts that is coordinated and maintained by CSEWG and NNDC at Brookhaven National Laboratory
3.3
evaluated nuclear data file
nuclear reaction database stored using a specified format and procedure
[2] [3] [4]
EXAMPLE ENDF/B , JEFF , and JENDL .
[1]
1) This edition is based on ANSI/ANS-6.1.2-2013 .
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ISO 23018:2022(E)
3.4
experimental benchmark
experiment for which conclusions can be drawn as to the accuracies of computational models and the
underlying nuclear data
Note 1 to entry: An experimental benchmark contains the following:
— a complete description of the conditions under which the experiment took place, including input data such
as reactor geometry, material compositions, core power distribution, relevant material temperatures, and
experimental conditions specified in sufficient detail to model or to replicate the experiment;
— measured data and their associated uncertainties.
Note 2 to entry: An experimental benchmark can provide “integral” or “differential” metrics; ”integral” pertains
to integral quantities such as reaction rates, while “differential” provides energy-dependent spectral information
such as time-of-flight measurements.
3.5
group-averaged cross section
cross section averaged over energy groups (intervals) as weighted by specified functions
3.6
JEFF
evaluated nuclear data file (3.3) produced via an international collaboration of NEA Data Bank
participating countries
3.7
JENDL
Japanese evaluated nuclear data file (3.3) for fast breeder reactors, thermal reactors, fusion neutronics
and shielding calculations, and other applications
3.8
neutron and gamma-ray cross section
cross section for the interactions of neutrons and gamma-rays with matter, including cross section for
the secondary emission of neutron and gamma-ray as well as cross section for the effects of neutron
and gamma-ray on materials (e.g. heating or helium generation)
3.9
numerical benchmark
specification of a set of input quantities (e.g. composition and geometry of bulk material and radiation
sources) and of reference calculated output quantities relevant to the benchmark (e.g. spatial and energy
dependence of neutron or gamma-ray fluence profiles) in detail sufficient to determine the accuracies
of a specified computational method when applied to modelling of the same input specifications
4 Abbreviations and acronyms
CSEWG Cross Section Evaluation Working Group
KERMA Kinetic Energy Released per unit Mass
LWR Light Water Reactor
NEA Nuclear Energy Agency
NNDC National Nuclear Data Center
OECD Organisation for Economic Co-operation and Development
ORNL Oak Ridge National Laboratory
RSICC Radiation Safety Information Computational Center
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ISO 23018:2022(E)
5 Preparation of group-averaged neutron and gamma-ray cross sections
5.1 Evaluated nuclear data files
Evaluated nuclear data files shall be derived from documented and reviewed information, including
basic experimental data, nuclear models, and systematics. The evaluated microscopic cross sections
shall be expressed as unique physical parameters and piecewise-continuous functions of incident
particle energy, of secondary particle energy, and of secondary particle angle with respect to the
incident particle direction. The evaluation shall be in sufficient detail for shielding applications, shall be
reviewed and documented, and should be tested against benchmark experiments.
5.2 Checking of evaluated nuclear data files
Before processing an evaluated nuclear data file, it should be checked for format conformation, data
[5]
validity, recommended procedure conformation, and physics content (e.g. using ENDF Utility Codes ).
If KERMA values are computed for response functions, cross-section processing code outputs (showing
kinematics limits from the total momentum conservation) should be analysed to avoid KERMA
[6]
calculation problems .
5.3 Energy ranges and materials of importance
-5
The evaluated nuclear data files should cover energy ranges (~10 eV to ~20 MeV for neutrons and
~1 keV to ~30 MeV for gamma-rays), and materials (shield materials as well as other materials required
for calculation of radiation sources) of importance in radiation protection and shielding calculations for
nuclear reactors.
5.4 Group-averaging techniques
5.4.1 General
Evaluated nuclear data files shall be averaged over energy groups by numerical techniques that do not
significantly degrade the accuracy of the evaluated nuclear data files for the application of interest.
Weighting functions and energy group structures should be appropriate for the application. The group-
averaging process should be carried out by tested, verified, and validated computer codes that have
been documented and reviewed.
5.4.2 Fine-group structures
The number and energy boundaries of fine energy groups shall be specified in detail sufficient to yield
adequate accuracy for radiation protection and shielding calculations, even if the number of fine groups
would impose significant computation time, memory, and hard disk space requirements. A fine-group
structure shall be used to determine the sufficiency of a coarser collapsed-group structure derived
from that fine-group structure. The fine-group structure shall accommodate weighting functions that
will be applied to the collapse of the fine-group data to coarse groups. The fine-group structure should
be insensitive to the weighting f
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