Nuclear energy - Determination of neutron fluence and displacement per atom (dpa) in reactor vessel and internals (ISO 19226:2017)

ISO 19226:2017 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, between the ends of active fuel assemblies, given the neutron source in the core.
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).
ISO 19226:2017 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.

Kernenergie - Bestimmung der Neutronenfluenz und Verschiebungen pro Atom (dpa) im Reaktordruckbehälter und Einbauten (ISO 19226:2017)

Dieses Dokument stellt bei gegebener Neutronenquelle im Kern ein Verfahren zur Auswertung von Bestrahlungsdaten im Bereich zwischen dem Reaktorkern und der Innenfläche des Kernbehälters, durch den Reaktordruckbehälter und die Reaktor¬grube, auf der Höhe der Brennstoffsäule dar.
ANMERKUNG   Diese Bestrahlungsdaten könnten Neutronenfluenz oder Verschiebungen pro Atom (dpa) und Helium¬produktion sein.
Die Auswertung erfolgt sowohl mit Neutronenflussberechnungen als auch mit Messdaten aus der Behälter  und Hohlraumdosimetrie. Dieses Dokument gilt für Druckwasserreaktoren (DWR), Siedewasser¬reaktoren (SWR) und Druckschwerwasserreaktoren (PHWR).
Dieses Dokument enthält außerdem ein Verfahren zur Bewertung der Neutronenschädigungseigenschaften am Reaktordruckbehälter und der Kerneinbauten von DWR, SWR und PHWR. Die Schädigungs¬eigenschaften konzentrieren sich auf atomare Verschiebungsschäden durch direkte Verschiebungen von Atomen aufgrund von Kollisionen mit Neutronen und indirekte Schäden durch Gasproduktion, die beide stark vom Neutronenenergiespektrum abhängig sind. Daher sind Berechnungen der Gesamtzahl der Atomverschiebungen für eine gegebene Neutronenfluenz und ein gegebenes Neutronenenergiespektrum wichtige Daten, die für das Management der Reaktorlebensdauer verwendet werden.

Énergie nucléaire - Détermination de la fluence neutronique et des déplacements par atome (dpa) dans la cuve et les internes du réacteur (ISO 19226:2017)

Le présent document fournit une procédure d'évaluation des données d'irradiation dans la région située entre le cœur du réacteur et la surface interne de la cuve, à travers la cuve sous pression et la cavité du réacteur, entre les extrémités des assemblages combustibles, pour une source donnée de neutrons dans le cœur.
NOTE       Ces données d'irradiation peuvent être la fluence neutronique ou les déplacements par atome (dpa), et la production d'Hélium.
Cette évaluation s'appuie à la fois sur des calculs de flux de neutrons et sur des données de mesures de dosimétrie à l'intérieur de la cuve et de la cavité, selon les cas. Le présent document s'applique aux réacteurs à eau sous pression (Pressurized Water Reactors, PWR), aux réacteurs à eau bouillante (Boiling Water Reactors, BWR) et aux réacteurs à eau lourde pressurisée (Pressurized Heavy Water Reactors, PHWR).
Le présent document donne également une procédure d'évaluation des endommagements dus aux neutrons sur la cuve sous pression du réacteur et les composants internes des PWR, BWR et PHWR. Les endommagements sont axés sur les dommages de déplacements atomiques causés par le déplacement direct des atomes dû aux collisions avec les neutrons, et sur les dommages indirects causés par la production de gaz, les deux types de dommages étant fortement dépendants du spectre d'énergie des neutrons. Pour une fluence neutronique et un spectre d'énergie des neutrons donnés, le calcul du nombre cumulé total de déplacements atomiques est donc une donnée importante à utiliser pour la gestion de la durée de vie du réacteur.

Jedrska energija - Ugotavljanje pretoka nevtronov in premikov na atom (dpa) v reaktorski posodi in vgrajenih delih (ISO 19226:2017)

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Publication Date
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SLOVENSKI STANDARD
SIST EN ISO 19226:2020
01-maj-2020
Jedrska energija - Ugotavljanje pretoka nevtronov in premikov na atom (dpa) v
reaktorski posodi in vgrajenih delih (ISO 19226:2017)

Nuclear energy - Determination of neutron fluence and displacement per atom (dpa) in

reactor vessel and internals (ISO 19226:2017)

Kernenergie - Bestimmung der Neutronenfluenz und Verschiebungen pro Atom (dpa) im

Reaktorbehälter und Einbauten (ISO 19226:2017)

Énergie nucléaire - Détermination de la fluence neutronique et des déplacements par

atome (dpa) dans la cuve et les internes du réacteur (ISO 19226:2017)
Ta slovenski standard je istoveten z: EN ISO 19226:2020
ICS:
27.120.10 Reaktorska tehnika Reactor engineering
SIST EN ISO 19226:2020 en,fr,de

2003-01.Slovenski inštitut za standardizacijo. Razmnoževanje celote ali delov tega standarda ni dovoljeno.

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SIST EN ISO 19226:2020
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SIST EN ISO 19226:2020
EN ISO 19226
EUROPEAN STANDARD
NORME EUROPÉENNE
February 2020
EUROPÄISCHE NORM
ICS 27.120.10
English Version
Nuclear energy - Determination of neutron fluence and
displacement per atom (dpa) in reactor vessel and
internals (ISO 19226:2017)

Énergie nucléaire - Détermination de la fluence Kernenergie - Bestimmung der Neutronenfluenz und

neutronique et des déplacements par atome (dpa) dans Verschiebungen pro Atom (dpa) im Reaktorbehälter

la cuve et les internes du réacteur (ISO 19226:2017) und Einbauten (ISO 19226:2017)

This European Standard was approved by CEN on 6 January 2020.

CEN members are bound to comply with the CEN/CENELEC Internal Regulations which stipulate the conditions for giving this

European Standard the status of a national standard without any alteration. Up-to-date lists and bibliographical references

concerning such national standards may be obtained on application to the CEN-CENELEC Management Centre or to any CEN

member.

This European Standard exists in three official versions (English, French, German). A version in any other language made by

translation under the responsibility of a CEN member into its own language and notified to the CEN-CENELEC Management

Centre has the same status as the official versions.

CEN members are the national standards bodies of Austria, Belgium, Bulgaria, Croatia, Cyprus, Czech Republic, Denmark, Estonia,

Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Latvia, Lithuania, Luxembourg, Malta, Netherlands, Norway,

Poland, Portugal, Republic of North Macedonia, Romania, Serbia, Slovakia, Slovenia, Spain, Sweden, Switzerland, Turkey and

United Kingdom.
EUROPEAN COMMITTEE FOR STANDARDIZATION
COMITÉ EUROPÉEN DE NORMALISATION
EUROPÄISCHES KOMITEE FÜR NORMUNG
CEN-CENELEC Management Centre: Rue de la Science 23, B-1040 Brussels

© 2020 CEN All rights of exploitation in any form and by any means reserved Ref. No. EN ISO 19226:2020 E

worldwide for CEN national Members.
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SIST EN ISO 19226:2020
EN ISO 19226:2020 (E)
Contents Page

European foreword ....................................................................................................................................................... 3

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SIST EN ISO 19226:2020
EN ISO 19226:2020 (E)
European foreword

The text of ISO 19226:2017 has been prepared by Technical Committee ISO/TC 85 "Nuclear energy,

nuclear technologies, and radiological protection” of the International Organization for Standardization

(ISO) and has been taken over as EN ISO 19226:2020 by Technical Committee CEN/TC 430 “Nuclear

energy, nuclear technologies, and radiological protection” the secretariat of which is held by AFNOR.

This European Standard shall be given the status of a national standard, either by publication of an

identical text or by endorsement, at the latest by August 2020, and conflicting national standards shall

be withdrawn at the latest by August 2020.

Attention is drawn to the possibility that some of the elements of this document may be the subject of

patent rights. CEN shall not be held responsible for identifying any or all such patent rights.

According to the CEN-CENELEC Internal Regulations, the national standards organizations of the

following countries are bound to implement this European Standard: Austria, Belgium, Bulgaria,

Croatia, Cyprus, Czech Republic, Denmark, Estonia, Finland, France, Germany, Greece, Hungary, Iceland,

Ireland, Italy, Latvia, Lithuania, Luxembourg, Malta, Netherlands, Norway, Poland, Portugal, Republic of

North Macedonia, Romania, Serbia, Slovakia, Slovenia, Spain, Sweden, Switzerland, Turkey and the

United Kingdom.
Endorsement notice

The text of ISO 19226:2017 has been approved by CEN as EN ISO 19226:2020 without any modification.

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SIST EN ISO 19226:2020
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SIST EN ISO 19226:2020
INTERNATIONAL ISO
STANDARD 19226
First edition
2017-11
Nuclear energy — Determination of
neutron fluence and displacement
per atom (dpa) in reactor vessel and
internals
Énergie nucléaire — Détermination de la fluence neutronique et du
déplacement par atome (dpa) dans la cuve et les internes du réacteur
Reference number
ISO 19226:2017(E)
ISO 2017
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SIST EN ISO 19226:2020
ISO 19226:2017(E)
COPYRIGHT PROTECTED DOCUMENT
© ISO 2017, Published in Switzerland

All rights reserved. Unless otherwise specified, no part of this publication may be reproduced or utilized otherwise in any form

or by any means, electronic or mechanical, including photocopying, or posting on the internet or an intranet, without prior

written permission. Permission can be requested from either ISO at the address below or ISO’s member body in the country of

the requester.
ISO copyright office
Ch. de Blandonnet 8 • CP 401
CH-1214 Vernier, Geneva, Switzerland
Tel. +41 22 749 01 11
Fax +41 22 749 09 47
copyright@iso.org
www.iso.org
ii © ISO 2017 – All rights reserved
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SIST EN ISO 19226:2020
ISO 19226:2017(E)
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 ........................................................................................................................................................................ 4

4.2.1 Input data ............................................................................................................................................................................... 4

4.2.2 Discrete ordinates (SN) method ......................................................................................................................... 4

4.2.3 Monte Carlo transport method ............................................................................................................................ 4

4.2.4 Adjoint fluence calculations ................................................................................................................................... 5

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 Stable-product neutron dosimeters ..................................................................................................................................... 7

5.4 Dosimeter response parameters ............................................................................................................................................. 7

5.5 Uncertainty estimates and measurement validation in standard neutron fields .......................... 7

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 .................... 8

6.4 Comparison of the calculations against measurements using least-squares methods ............ 8

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 ........................................................................................................................................................................................ 9

Bibliography .............................................................................................................................................................................................................................11

© ISO 2017 – All rights reserved iii
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SIST EN ISO 19226:2020
ISO 19226:2017(E)
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 19.10-2009 but extends to cover the evaluation of irradiation

damage due to neutron fluence.
iv © ISO 2017 – All rights reserved
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SIST EN ISO 19226:2020
ISO 19226:2017(E)
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 vessel 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.
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SIST EN ISO 19226:2020
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SIST EN ISO 19226:2020
INTERNATIONAL STANDARD ISO 19226:2017(E)
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, between the ends of active fuel assemblies, given the neutron source in the core.

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 http://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
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SIST EN ISO 19226:2020
ISO 19226:2017(E)
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.
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)

mean number of times each atom of a solid is displaced from its lattice site during an exposure to

displacing radiation, as calculated following standard procedures
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 neutron fluence rate (i.e. the time-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
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SIST EN ISO 19226:2020
ISO 19226:2017(E)
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
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

[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.
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4.2 Transport calculation
4.2.1 Input data
The four major types of input required are.
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) are required.
b) Geometric model:

The geometric model should represent the physical configuration as closely as practical. “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 accuracy of the data evaluation (ENDF/B, JEFF, JENDL…);
2) the energy group structure;
3) the order of the scattering anisotropy (i.e. P expansion);
4) the method used for group-collapsing.
d) Core neutron source:

The determination of the neutron source should include the temporal, spatial, 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 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 envelope (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. 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 solution 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. Typically, Monte Carlo codes allow use of either multigroup or continuous-energy

cross sections. The continuous-energy cross sections are recommended. Variance-reduction techniques

4 © ISO 2017 – All rights reserved
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ISO 19226:2017(E)

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 to

improve the statistics at locations far from the core include the following, provided that preliminary

checking has been done:
a) source biasing;
b) geometry splitting with Russian roulette;
c) increasing importances;
d) surface restarts;
e) weight windows.
4.2.4 Adjoint fluence 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 source biasing in the direct mode calculation;
— 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 to track the fluence during plant operation. However, when the operating

conditions that affect the transport calculation (e.g. downcomer and core bypass coolant densities,

core mechanical design) remain the same, the multiple transport calculations may be replaced by a

[6]
single adjoint calculation .

The adjoint is calculated for an adjoint source located at the vessel or other location of interest that

is taken to be proportional to the energy-dependent response cross section. Typically, in the case of

flux and/or fluence (E > 1 MeV), the source is taken to be unity above 1,0 MeV and zero below 1,0 MeV.

When a dosimeter reaction rate is required, the source typically is taken to be equal to an energy-

dependent dosimeter cross section. The fluence (or reaction rate response) at the location of interest is

then determined for each cycle by integrating the cycle-specific core neutron source over the calculated

adjoint function.

If Monte-Carlo method is used, and if adjoint mode is not available in the code, there may exist options

in direct mode that identify the originated sources (spatially and in energy).
4.3 Validation of neutron fluence calculational values

Prior to performing transport calculations for a particular facility, the calculational methodology shall

be validated by
a) comparing results with benchmarked calculations and measurements, and
b) demonstrating that it accurately determines appropriate benchmark results.
4.4 Determination of calculational uncertainties

Calculational uncertainties associated with the methodology for predicting neutron fluence typically

include the following:

a) nuclear data (e.g. transport cross sections, dosimeter reaction cross sections, and fission spectra);

b) geometry (e.g. locations of internals and deviations from the nominal dimensions);

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c) isotopic composition of material (e.g. density and composition of coolant water, vessel internals,

the core barrel, thermal shielding, the pressure vessel with cladding, and concrete shielding);

d) neutron sources (e.g. space and energy distribution depending on fuel burnup);

e) m
...

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