prEN ISO 19226

SLOVENSKI STANDARD
01-marec-2026
Jedrska energija - Ugotavljanje pretoka nevtronov in premikov na atom (dpa) v
reaktorski posodi in vgrajenih delih (ISO/DIS 19226:2026)
Nuclear energy - Determination of neutron fluence and displacement per atom (dpa) in
reactor vessel and internals (ISO/DIS 19226:2026)
Kernenergie - Bestimmung der Neutronenfluenz und Verschiebungen pro Atom (dpa) im
Reaktordruckbehälter und Einbauten (ISO/DIS 19226:2026)
Énergie nucléaire - Détermination de la fluence neutronique et des déplacements par
atome (dpa) dans la cuve et les internes de réacteur (ISO/DIS 19226:2026)
Ta slovenski standard je istoveten z: prEN ISO 19226
ICS:
27.120.10 Reaktorska tehnika Reactor engineering
2003-01.Slovenski inštitut za standardizacijo. Razmnoževanje celote ali delov tega standarda ni dovoljeno.

DRAFT
International
Standard
ISO/DIS 19226
ISO/TC 85/SC 6
Nuclear energy — Determination of
Secretariat: DIN
neutron fluence and displacement
Voting begins on:
per atom (dpa) in reactor vessel and
2026-01-06
internals
Voting terminates on:
2026-03-31
Énergie nucléaire — Détermination de la fluence neutronique et
des déplacements par atome (dpa) dans la cuve et les internes du
réacteur
ICS: 27.120.10
THIS DOCUMENT IS A DRAFT CIRCULATED
FOR COMMENTS AND APPROVAL. IT
IS THEREFORE SUBJECT TO CHANGE
AND MAY NOT BE REFERRED TO AS AN
INTERNATIONAL STANDARD UNTIL
PUBLISHED AS SUCH.
This document is circulated as received from the committee secretariat.
IN ADDITION TO THEIR EVALUATION AS
BEING ACCEPTABLE FOR INDUSTRIAL,
TECHNOLOGICAL, COMMERCIAL AND
USER PURPOSES, DRAFT INTERNATIONAL
STANDARDS MAY ON OCCASION HAVE TO
ISO/CEN PARALLEL PROCESSING
BE CONSIDERED IN THE LIGHT OF THEIR
POTENTIAL TO BECOME STANDARDS TO
WHICH REFERENCE MAY BE MADE IN
NATIONAL REGULATIONS.
RECIPIENTS OF THIS DRAFT ARE INVITED
TO SUBMIT, WITH THEIR COMMENTS,
NOTIFICATION OF ANY RELEVANT PATENT
RIGHTS OF WHICH THEY ARE AWARE AND TO
PROVIDE SUPPORTING DOCUMENTATION.
Reference number
ISO/DIS 19226:2026(en)
DRAFT
ISO/DIS 19226:2026(en)
International
Standard
ISO/DIS 19226
ISO/TC 85/SC 6
Nuclear energy — Determination of
Secretariat: DIN
neutron fluence and displacement
Voting begins on:
per atom (dpa) in reactor vessel and
internals
Voting terminates on:
Énergie nucléaire — Détermination de la fluence neutronique et
des déplacements par atome (dpa) dans la cuve et les internes du
réacteur
ICS: 27.120.10
THIS DOCUMENT IS A DRAFT CIRCULATED
FOR COMMENTS AND APPROVAL. IT
IS THEREFORE SUBJECT TO CHANGE
AND MAY NOT BE REFERRED TO AS AN
INTERNATIONAL STANDARD UNTIL
PUBLISHED AS SUCH.
This document is circulated as received from the committee secretariat.
IN ADDITION TO THEIR EVALUATION AS
BEING ACCEPTABLE FOR INDUSTRIAL,
© ISO 2026
TECHNOLOGICAL, COMMERCIAL AND
USER PURPOSES, DRAFT INTERNATIONAL
All rights reserved. Unless otherwise specified, or required in the context of its implementation, no part of this publication may
STANDARDS MAY ON OCCASION HAVE TO
ISO/CEN PARALLEL PROCESSING
be reproduced or utilized otherwise in any form or by any means, electronic or mechanical, including photocopying, or posting on
BE CONSIDERED IN THE LIGHT OF THEIR
the internet or an intranet, without prior written permission. Permission can be requested from either ISO at the address below
POTENTIAL TO BECOME STANDARDS TO
WHICH REFERENCE MAY BE MADE IN
or ISO’s member body in the country of the requester.
NATIONAL REGULATIONS.
ISO copyright office
RECIPIENTS OF THIS DRAFT ARE INVITED
CP 401 • Ch. de Blandonnet 8
TO SUBMIT, WITH THEIR COMMENTS,
CH-1214 Vernier, Geneva
NOTIFICATION OF ANY RELEVANT PATENT
Phone: +41 22 749 01 11
RIGHTS OF WHICH THEY ARE AWARE AND TO
PROVIDE SUPPORTING DOCUMENTATION.
Email: copyright@iso.org
Website: www.iso.org
Published in Switzerland Reference number
ISO/DIS 19226:2026(en)
ii
ISO/DIS 19226:2026(en)
Contents Page
Foreword .iv
Introduction .v
1 Scope . 1
2 Normative references . 1
3 Terms and definitions . 1
4 Transport theory calculational models . 3
4.1 General .3
4.1.1 Output requirements .3
4.1.2 Methodology: transport calculations with fixed sources .3
4.2 Transport calculation .3
4.2.1 Input data .3
4.2.2 Discrete ordinates (SN) method .4
4.2.3 Monte Carlo transport method .4
4.3 Validation of neutron fluence calculational values .5
4.4 Determination of calculational uncertainties .5
5 Reactor pressure vessel neutron dosimetry measurements . 6
5.1 Introduction .6
5.2 General requirements for reactor vessel neutron metrology.6
5.3 In-vessel surveillance capsules .7
5.4 Ex-vessel surveillance capsules .7
5.5 Uncertainty estimates and measurement validation in standard neutron fields .8
6 Comparison of calculations with measurements. 8
6.1 Introduction .8
6.2 Direct comparison of calculated activities with measured sensor activities .8
6.3 Comparison of calculated rates with measured average full-power reaction rates .9
6.4 Comparison of the calculations against measurements using least-squares methods .9
7 Determination of the best-estimate fluence . 9
8 Calculational methods for dpa and gas production . 9
8.1 Displacements per atom (dpa) .9
8.2 Gas production . .10
Bibliography .11

iii
ISO/DIS 19226:2026(en)
Foreword
ISO (the International Organization for Standardization) is a worldwide federation of national standards
bodies (ISO member bodies). The work of preparing International Standards is normally carried out through
ISO technical committees. Each member body interested in a subject for which a technical committee
has been established has the right to be represented on that committee. International organizations,
governmental and non-governmental, in liaison with ISO, also take part in the work. ISO collaborates closely
with the International Electrotechnical Commission (IEC) on all matters of electrotechnical standardization.
The procedures used to develop this document and those intended for its further maintenance are described
in the ISO/IEC Directives, Part 1. In particular the different approval criteria needed for the different types
of ISO documents should be noted. This document was drafted in accordance with the editorial rules of the
ISO/IEC Directives, Part 2 (see www.iso.org/directives).
Attention is drawn to the possibility that some of the elements of this document may be the subject of patent
rights. ISO shall not be held responsible for identifying any or all such patent rights. Details of any patent
rights identified during the development of the document will be in the Introduction and/or on the ISO list of
patent declarations received (see www.iso.org/patents).
Any trade name used in this document is information given for the convenience of users and does not
constitute an endorsement.
For an explanation on the voluntary nature of standards, the meaning of ISO specific terms and
expressions related to conformity assessment, as well as information about ISO's adherence to the World
Trade Organization (WTO) principles in the Technical Barriers to Trade (TBT) see the following URL:
www.iso.org/iso/foreword.html.
This document was prepared by Technical committee ISO/TC 85, Nuclear energy, nuclear technologies, and
radiological protection, Subcommittee SC 6, Reactor Technology.
This document is based on the ANSI/ANS 19.10-2024 but extends to cover the evaluation of irradiation
damage due to neutron fluence.

iv
ISO/DIS 19226:2026(en)
Introduction
This document is intended for use by
a) those involved in the determination of exposure parameters for the prediction of irradiation damage
to the vessel and to the internals of a nuclear reactor, where the exposure parameters can be neutron
fluence and/or displacements per atom (dpa),
b) those involved in the determination of material properties of irradiated reactor vessels and reactor
internals,
c) regulatory agencies in licensing actions such as the writing of Regulatory Guides, analysis of reports
concerning the integrity and material properties of irradiated pressure vessels and reactor internals.

v
DRAFT International Standard ISO/DIS 19226:2026(en)
Nuclear energy — Determination of neutron fluence and
displacement per atom (dpa) in reactor vessel and internals
1 Scope
This document provides a procedure for the evaluation of irradiation data in the region between the reactor
core and the inside surface of the containment vessel, through the pressure vessel and the reactor cavity.
NOTE These irradiation data could be neutron fluence or displacements per atom (dpa), and Helium production.
The evaluation employs both neutron flux computations and measurement data from in-vessel and cavity
dosimetry, as appropriate. This document applies to pressurized water reactors (PWRs), boiling water
reactors (BWRs), and pressurized heavy water reactors (PHWRs).
This document also provides a procedure for evaluating neutron damage properties at the reactor pressure
vessel and internal components of PWRs, BWRs, and PHWRs. Damage properties are focused on atomic
displacement damage caused by direct displacements of atoms due to collisions with neutrons and indirect
damage caused by gas production, both of which are strongly dependent on the neutron energy spectrum.
Therefore, for a given neutron fluence and neutron energy spectrum, calculations of the total accumulated
number of atomic displacements are important data to be used for reactor life management.
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.
ANSI/ANS 19.10, Methods for determining neutron fluence in BWR and PWR pressure vessel and reactor
internals
ASTM E170-16a, Standard Terminology Relating to Radiation Measurements and Dosimetry
3 Terms and definitions
For the purposes of this document, the terms and definitions given in ANSI/ANS 19.10, ASTM E170-16a 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
accuracy of a measured/calculated value
difference between the “real” and the measured/calculated value, typically due to systematic errors in the
measurement/calculation procedure
3.2
benchmark experiment
well-defined set of physical experiments with results judged to be sufficiently accurate for use as a
calculational reference point
Note 1 to entry: The judgment is made by a group of experts in the subject area.

ISO/DIS 19226:2026(en)
3.3
best-estimate fluence
most accurate value of the fluence based on all available measurements, calculated results, and adjustments
based on bias estimates, least-squares analyses, and engineering judgment
3.4
calculational methodology
mathematical equations, approximations, assumptions, associated parameters, and calculational procedure
that yield the calculated results
Note 1 to entry: When more than one step is involved in the calculation, the entire sequence of steps comprises the
“calculational methodology.”
3.5
code benchmark
comparison to the results of another code system that has been previously validated against experiment(s)
3.6
continuous-energy cross-section data
cross-section data that are specified in a dense point-wise manner that comprises the energy range
3.7
dosimeter reaction
neutron-induced nuclear reaction with a product nuclide having sufficient activity to be measured and
related to the incident neutron fluence
3.8
displacements per atom (dpa)
Consequence-based radiation exposure measurement unit representing the total number of displacements
per atom of a material as a result of neutron irradiation over a specified time
3.9
least-squares adjustment procedure
method for combining the results of neutron transport calculations and the results of dosimetry
measurements that provides an optimal estimate of the fluence by minimizing, in the least-squares sense,
the calculation-to-measurement differences
3.10
multigroup cross-section data
cross-section data that have been determined by averaging the continuous-energy cross-section data over
discrete energy intervals using specified weighting functions to preserve reaction rates
3.11
neutron fluence
time-integrated and energy-integrated neutron fluence rate (i.e. the time-integrated and energy-integrated
neutron flux) as expressed in neutrons per square centimeter
3.12
precision of a measured/calculated value
standard deviation (if available from a set of repeated measurements/calculations) of the distribution of the
measured or calculated physical value
3.13
reactor internals
reactor structure components that are within the pressure vessel such as the core baffle, core barrel,
thermal shield, lower and upper core plates in PWRs and BWRs

ISO/DIS 19226:2026(en)
3.14
solution variance
measure of the statistical variance of the Monte Carlo transport solution due to a finite number of particle
histories
Note 1 to entry: Mathematically, it is the second central moment of the distribution about the mean value, which is
used to measure the dispersion of the distribution about the mean.
4 Transport theory calculational models
4.1 General
4.1.1 Output requirements
The transport calculations need to be able to determine accurately the neutron flux or fluence distributions,
and/or other response parameters such as reaction rates or dpa for the analysis of integral dosimetry
measurements and for the prediction of irradiation damage to reactor pressure vessels and its internals.
Calculation methodologies described in this document focus on neutron fluence for determining radiation
embrittlement of reactor vessel materials.
While neutron fluence (E > 1,0 MeV) (where neutron fluence (E > 1,0 MeV) represents the fluence of neutrons
with energy above 1,0 MeV) has frequently been selected as the exposure parameter for determining
radiation embrittlement of reactor vessel materials, the procedures in this document extend to include
fluence spectrum above 0,1 MeV, in addition to thermal fluence below 0,625 eV.
Some parameters of the calculations would be determined based on
— direct use of the results: design or comparison to measurements (which imply envelope or best-estimate
results, respectively),
— required response functions: (E > 1,0 MeV) neutron flux, (E > 0,1 MeV) neutron flux, thermal neutron flux
(E < 0,625 eV), dpa/s, dosimeter reaction rates;
NOTE The figures for flux, given as examples of upper or lower limit, depend on the application.
— location(s) of interest: fineness of the spatial meshing.
4.1.2 Methodology: transport calculations with fixed sources
In the practice suggested in this document, a source distribution throughout the core is prepared using the
results of core physics calculations; multidimensional transport theory calculations then are performed to
propagate the neutrons to regions outside the core.
This document uses codes based on transport theory to determine multigroup three-dimensional flux
distributions and to evaluate the reaction rates of dosimetry materials or dpa properties through proper
use of response functions or cross sections.
[2]
Transport theory calculations should be performed using deterministic discrete ordinates (S ) or
N
[3]
statistical Monte Carlo approaches as discussed in 4.2.2 and 4.2.3, respectively. Other transport methods
may be used if they are part of a benchmarked methodology.
4.2 Transport calculation
4.2.1 Input data
The four major types of input required are.

ISO/DIS 19226:2026(en)
a) Material composition:
The material compositions should represent the physical configuration as closely as practical. Material
compositions and densities (consistent with the geometric model), coolant and moderator density
(consistent with operating conditions associated with temperature) are required.
b) Geometric model:
The geometric model should represent the physical configuration as closely as practical, including
hot dimensions. Otherwise, “as-built” dimensions of the reactor configuration should be used when
available.
c) Cross-section data:
Appropriate cross-section data should be used. Cross-section sets may be used if they are part of a
benchmarked methodology. Major considerations include:
1) the pertinence of the data evaluation (ENDF/B, JEFF, JENDL…);
2) the energy group structure;
3) the order of the scattering anisotropy (i.e. P expansion);
n
4) the method used for group-collapsing.
d) Core neutron source:
The determination of the neutron source should include the temporal, spatial, angular and energy
dependence together with the absolute source normalization. The spatial distribution(s) of sources
shall be representative of the integrated or averaged distribution(s) during the considered irradiation
duration(s). The angular distribution can be considered as isotropic. The neutron distribution should be
accurate especially at the periphery of the core, in order to properly determine the fluence on the Reactor
Pressure Vessel. Also, the neutron source spectrum (spectra) shall be determined and the average
number(s) of neutrons produced per fission, ν, shall be selected. All these parameters are to be chosen
with regards to the calculated data: representative of irradiation conditions (in case of comparisons to
measurements), or enveloping (in case of design phase for internals and/or vessel analyses).
4.2.2 Discrete ordinates (SN) method
In order to ensure an accurate representation of three-dimensional effects, three-dimensional discrete
ordinates transport calculations should be used when practical. These calculations can be based on methods
such as finite differences, finite elements or method of characteristics (MOC). When three-dimensional
calculations are not practical, a synthesis method may be used to determine the three-dimensional flux or
fluence distribution. In this approach, the fluence distribution is determined by synthesizing the results of
one- and two-dimensional discrete ordinates solutions (see References [4] and [29]). The results depend on
the specific locations where the neutron flux/fluence has to be determined (location of interest), i.e., not
only at the core mid-plane, in general. Note that the use of synthesis technique may lead into inaccurate
results if the material and/or source distributions are highly three dimensional.
4.2.3 Monte Carlo transport method
In addition to the considerations a) to d) in 4.2.1, the Monte Carlo model construction could require a
technique to reduce the score variance. The geometric model used in the Monte Carlo analyses should reflect
the actual physical configuration. The great flexibility in typical Monte Carlo codes allows a very detailed
representation, and this should be used to represent all the important features of the geometry under
consideration and precise dosimeter location. Typically, Monte Carlo codes allow use of either multigroup or
continuous-energy cross sections. Continuous-energy cross sections are recommended. Variance-reduction
techniques that have been validated for these applications may be used to reduce the variance in the Monte
Carlo calculation (some of them are presented in the References [3] and [5]). Techniques that may be used

ISO/DIS 19226:2026(en)
to improve the statistics at locations far from the core include the following, provided that preliminary
checking especially on potential bias on the score has been done:
a) source biasing;
b) splitting with Russian roulette which can be based on weight windows or Adaptive Multilevel
Splitting (AMS);
c) surface restarts;
d) Exponential biasing.Adjoint neutron transport calculations
Adjoint calculations may be performed:
— as upstream calculations, to estimate the space- and energy- dependent importance of the core neutrons
to a specific location (on the vessel or on the considered internal), in order to determine the variance
reduction parameters used in the subsequent Monte Carlo simulation step;
— or else, to replace multiple transport calculations in direct mode:
Because the reactor conditions are generally dependent on the fuel cycle, multiple transport
calculations are required
...