ISO 18075:2026

International
Standard
ISO 18075
Second edition
Reactor technology — Power
2026-08
reactor analysis — Steady-state
neutronics methods
Technologie du réacteur — Analyse des réacteurs de puissance —
Méthodes stationnaires en neutronique
Reference number
© ISO 2026
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ii
Contents Page
Foreword .iv
Introduction .v
1 Scope . 1
2 Normative references . 1
3 Terms, definitions and abbreviations . 1
3.1 Terms and definitions .1
3.2 Abbreviations .5
4 Relation to other standards . 5
5 Methods of calculation . 5
5.1 General .5
5.2 Conditions to be considered .6
5.3 Multigroup cross sections .7
5.3.1 Basic data .7
5.3.2 Preparation of multigroup constants .7
5.3.3 System dependent spectrum calculations .7
5.3.4 Choice of cell and supercell .7
5.3.5 Cell environment .8
5.3.6 Calculation model .8
5.4 Collapse to few-groups and spatial homogenization .8
5.5 Calculation of reactivity, reaction rates and neutron flux distributions .9
5.5.1 Models .9
5.5.2 Uncertainties and assumptions .9
5.6 Calculation of reaction rates in reactor components .10
5.7 Depletion calculations .10
5.8 Common practices . 12
5.8.1 General . 12
5.8.2 PWR core physics methods . 12
5.8.3 BWR core physics methods. 13
5.8.4 LMR core physics methods . 15
5.8.5 PHWR core physics methods .17
5.8.6 HTGR core physics methods .19
6 Verification and validation of the calculation system .20
6.1 Overview . 20
6.2 Verification . . 20
6.2.1 General . 20
6.2.2 Unit testing . 20
6.2.3 Integral testing .21
6.3 Validation .21
6.3.1 Overview .21
6.3.2 Unit testing .21
6.3.3 Integral testing . 22
6.3.4 Code to code comparisons . 22
6.4 Biases and uncertainties . 23
7 Documentation .23
8 Summary of requirements .24
Annex A (informative) Computer codes in common use .25
Bibliography .28

iii
Foreword
ISO (the International Organization for Standardization) is a worldwide federation of national standards
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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 document should be noted. This document was drafted in accordance with the editorial rules of the
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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. This document is based on a standard
[2]
developed by the American Nuclear Society (ANS) of which the current version is ANSI/ANS-19.3-2022 .
This second edition cancels and replaces the first (ISO 18075:2018), which has been technically revised.
The main changes are as follows:
— addition of new terms in Clause 3;
— figures slightly revised;
— 5.8.5 was amended so that CANDU is only an example;
— correction of internal references;
— Table A.1 corrected;
— editorial amendments in the whole document.
Any feedback or questions on this document should be directed to the user’s national standards body. A
complete listing of these bodies can be found at www.iso.org/members.html.

iv
Introduction
The design and operation of nuclear reactors require knowledge of the conditions under which a reactor
will be critical, as well as the degree of subcriticality or supercriticality when these conditions change. In
addition, knowledge is required of the spatial distribution of neutron reaction rates in reactor components
as a prerequisite, for example, for inferring proper power and temperature distributions to ensure the
satisfaction of thermal-limit and safety-limit requirements. Both reaction-rate spatial distributions and
reactivity have been measured by suitable experimental techniques, either in mock-ups or in the operating
reactors themselves. These quantities can also be calculated by various techniques. Available reactor
experimental data have been used to validate the steady-state neutronic calculations within reasonable
margins. As more accurate nuclear cross sections and improved calculation methods have become available,
steady-state neutronic calculations have been utilized extensively for nuclear fuel and core designs and
analyses and thus have become increasingly important.

v
International Standard ISO 18075:2026(en)
Reactor technology — Power reactor analysis — Steady-state
neutronics methods
1 Scope
This document provides guidance for performing and validating the sequence of steady-state calculations
leading to prediction in all types of operating commercial nuclear reactors, of the following:
— reaction-rate spatial distributions;
— reactivity;
— change of nuclide compositions with time.
The document provides
a) guidance for the selection of computational methods,
b) criteria for verification and validation of calculation methods used by reactor core analysts,
c) criteria for evaluation of accuracy and range of applicability of data and methods, and
d) requirements for documentation of the preceding.
2 Normative references
There are no normative references in this document.
3 Terms, definitions and abbreviations
For the purposes of this document, the following terms and definitions apply.
ISO and IEC maintain terminology 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 Terms and definitions
3.1.1
application-dependent multigroup
discrete energy-group structure that is intermediate between the application-independent multigroup
structure and a few-group structure
Note 1 to entry: The application-dependent multigroup structure can be such that the group constants are dependent
on reactor composition through an estimated neutron energy spectrum. An application dependent multigroup data
set is one type of averaged data set.

3.1.2
application-independent multigroup
discrete energy-group structure that is sufficiently detailed that the group constants may be considered as
being independent of reactor composition, geometry, or spectrum for a wide range of reactor analysis
Note 1 to entry: The application-independent multigroup structure can be employed directly in reactor-design
spectrum calculations, or it can be employed to generate group constants in an application-dependent multigroup
structure. An application-independent multigroup data set is one type of averaged data set.
3.1.3
averaged data set
data set prepared by averaging an evaluated data set or a processed continuous-energy data set with a
specified weighting function over a specified energy group structure
Note 1 to entry: The group structure and weighting functions may be selected to be application dependent or
[1]
application independent, e.g, light water reactors (LWRs), are dealt with in ANSI/ANS 19.1-2019 .
3.1.4
cell
supercell
one or more reactor components with associated coolant (and possibly additional moderator and structural
material) that, for computational purposes, are assumed to form a spatially repeating array in the reactor
Note 1 to entry: The simplest example of a cell is the “pin cell” in which a single fuel rod or pin is surrounded by coolant
(e.g. light water, heavy water, or sodium). Another example is a bundle of fuel rods cooled by heavy water within a
housing, surrounded by a heavy water moderator space.
Note 2 to entry: More complex geometric configurations are also used for some applications. These are often referred
to as “supercells”, or sometimes “(fuel) assembly cells”, although the exact definition of the term varies greatly
between reactor types and is even somewhat subjectively defined for a particular reactor type. Supercells, in the
context of this document, represent more complex “cell” configurations that involve a collection of contiguous cells
forming an assumed repeating array within the reactor, or augmented cells incorporating additional regions to serve
as a computational artifice, e.g. to account for significant spectrum effects due to compositions outside the cell, or cell
configurations including a reactivity device in addition to fuel, coolant, moderator and poison.
3.1.5
collapse
method by which a many energy group set of cross sections is reduced to fewer energy groups
3.1.6
data set
collection of microscopic cross sections and nuclear constants encompassing the range of materials and
reaction processes needed for the application area of interest
3.1.7
dilution cross section
cross section, which refers to a background cross section that represents the effects of other isotopes in a
material, particularly in the context of self-shielding in nuclear reactions
3.1.8
evaluated data set
data set (3.1.6) that is completely and uniquely specified over the ranges of energy and angles important to
reactor calculations
Note 1 to entry: Such a data set is based upon available information (experimental measurement results and nuclear
theories) and employs a judgment as to the best physical description of the interaction processes. An evaluated data
set is intended to be independent of reactor composition, geometries, energy group structures, and spectra.

3.1.9
experimental data
any experimentally measured quantity or quantities
Note 1 to entry: As such it is applied herein to both differential cross-section measurements and integral measurements
(e.g. control-rod worth) obtained from reactor experiments or operations.
3.1.10
few-group
energy-group (typically two-group) structure that is adopted for a particular application
Note 1 to entry: The few-group constants for a region are dependent on a specific reactor composition and geometry
through a calculated energy spectrum, and are also dependent on other conditions such as pressure, temperature,
xenon concentration, void fraction.
3.1.11
fine group
energy group structure (typically hundreds of groups for LWR, thousands for LFMR) that is adopted for a
particular application.
Note 1 to entry: The fine-group constants are dependent on temperature.
3.1.12
lattice
array of cells (or a fuel-assembly cell) with its associated immediate environment, such as the volume of
moderator associated with it
3.1.13
lumped pseudo fission product
fictitious fission product that is used in depletion calculations to represent a combination of certain minor
fission products that are not tracked in an explicit fission product model
Note 1 to entry: A lumped pseudo fission product is characterized by effective neutron cross sections and decay
constants.
3.1.14
methods of calculation
mathematical equations, approximations, assumptions, associated numerical parameters, and calculational
procedures 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
“calculation method”.
3.1.15
multigroup
more than a few-groups (3.1.10), can be hundreds or thousands of energy groups
3.1.16
probability table parameters
parameters for a probability table that is used to process cross sections
3.1.17
processed continuous-energy data set
data set (3.1.6) prepared by expansion or compaction of an evaluated data set (3.1.8) using specified
algorithms
Note 1 to entry: Such a data set is intended to be independent of reactor composition, geometries and spectra.
3.1.18
quasi steady-state condition
evolution of reactivity and flux is sufficiently low between two time intervals such that we can consider that
the core is under a succession of static states

3.1.19
reaction rate
rate at which neutrons interact with nuclei at a point in the reactor, a volume of the phase space or as an
integral over the reactor, according to the context
EXAMPLE The reaction rate for absorption, scattering and fission.
3.1.20
reactivity
the quantity (1 minus the eigenvalue, λ, of the steady-state neutron balance equation), written as:
MΦ = λ·FΦ
where
Φ is the neutron flux;
F is the neutron fission yield operator;
M is the scattering, absorption, and leakage operator.
Note 1 to entry: The effective multiplication factor, k , is the inverse of λ. Reactivity is a unitless, pure number. It
eff
−5
is, however, often written in terms of smaller “units”, such as milli-k = 0,001, pcm = 0,000 01 = 10 or “dollars” (and
cents), where 1 dollar is taken as the value of the delayed neutron fraction in the system of interest.
3.1.21
time-average model (CANDU)
time-average cell-homogenized few-group (3.1.10) macroscopic cross sections at each bundle location in the
core are calculated as flux-weighted quantities over the fluence range experienced by the fuel during its
residence at that specific location
Note 1 to entry: Consequently, the flux, fluence range and macroscopic cross sections at each location are linked
together and depend on the individual refuelling rate of the channel in which the fuel bundle resides. The resulting
self-consistency problem is solved iteratively to determine all the results of the time-average model.
Note 2 to entry: The CANada Deuterium Uranium (CANDU) reactor time-average model is also known as the
equilibrium-core model.
3.1.22
ultrafine group
energy group structure (typically ten thousand energy groups) that can allow to explicitly describe the
resolved resonances
Note 1 to entry: The ultrafine-group constants can be dependent on temperature.
3.1.23
experimental validation
process of determining the degree to which a model is an accurate representation of the real world from the
perspective of the intended uses of the model
3.1.24
numerical validation
process of determining the degree to which a model is accurate by comparison to a previously validated
code system
3.1.25
verification
process of determining that a model implementation accurately represents the developer’s conceptual
description of the model and the solution to the model

3.2 Abbreviations
BWR boiling water reactor
CANDU CANada Deuterium Uranium (reactor)
HTGR high temperature gas cooled reactor
LMR liquid metal reactor
LWR light water reactor
PHWR pressurized heavy water reactor
PWR pressurized water reactor
4 Relation to other standards
The following American National Standards are related to this document:
[1]
— ANSI/ANS-19.1-2019 defines the criteria to be employed in the preparation of application-independent
cross-section data files from experimental data and theoretical models. This document covers
subsequent space and energy averaging processes that may be employed to prepare cross sections for
use in the representation of the core and its environment, and the subsequent calculation of the spatial
distribution of neutron reaction rates in the core and of the core reactivity. There may be many ways
of carrying out the space and energy averaging to obtain few-group cross sections, and no unique path
for the preparation or use of cross sections employed in design calculations is defined, required, or
recommended by this standard.
[3] [4]
— ANSI/ANS-19.4-2017 (R2022) and ANSI/ANS 19.5-1995 (W2005) .
Validation of calculation systems requires comparison with available experimental results. The preceding
standards contain criteria for performing and documenting such experiments, in order to be most useful for
this purpose.
[5]
— ANSI/ANS 19.3.4-2002 (R2017) provides criteria for the establishment of the thermal energy
deposition rate distribution within a nuclear reactor core. Since the accuracy with which this can be
done is dominated by the accuracy with which neutron reaction rates can be calculated, ANSI/ANS-
19.3.4-2002 (R2017) is closely related to ANSI/ANS−19.3-2022.
[6]
— ANSI/ASME-NQA 1-2022 deals with quality assurance, including that for computer programs.
[9]
— ANSI/ANS-10.4−2008 (R2021) deals with requirements for verifying and validating computer codes,
such as those used for neutronics calculations.
[10]
— ANSI/ANS-10.5−2006 (R2021) deals with methods to respond to users’ requirements in computer
programs.
5 Methods of calculation
5.1 General
Calculations within the scope of this document are typically performed in a sequence of steps. A typical
sequence can be
a) utilization of averaged-data set cross sections, nuclide number densities, and geometrical information
(usually repeating cells or supercells) to calculate an application-dependent many-group neutron
spectrum and within-cell spatial distribution for each different reactor region and composition,

b) utilization of the spectra and spatial distributions from step a) to collapse and homogenize averaged-
data set cross sections to few-group form,
c) utilization of the few-group cross sections from step b) and geometrical information about the reactor
to calculate reactivity and few-group flux spatial distributions in the reactor,
d) utilization of the preceding information to compute reaction rates in physical reactor components
(including the reactor vessel if needed), and
e) utilization of the preceding information to calculate changes in nuclide composition of fuel and possibly
other reactor components with exposure.
In some applications, the collapse to few-group form is not used, and the multigroup cross sections
and geometrical information about the reactor are used to calculate reactivity and multigroup spatial
distributions in the reactor.
Not all steps in the sequence would normally be executed for a given problem. It is not a requirement of this
document that a particular sequence of calculations, such as the one previously listed, be used. Similarly,
the use of the preceding sequence does not, in itself, demonstrate compliance with this document. The use
of a specific calculation procedure shall be justified by the procedure presented in Clause 6. However, the
preceding sequence does provide an adequate framework within which most of the problems in steady-state
reactor physics calculations can be discussed. Therefore, each of the aforementioned steps will be discussed
in later passages of this subclause.
A summary of the requirements of this document is given in Clause 8.
5.2 Conditions to be considered
Consideration shall be given to all conditions that significantly affect the calculated quantities. The method
of calculation shall be capable of treating the reactor composition or configuration under the conditions
being studied.
Important conditions that may be significant include, but are not limited to the following:
a) presence of control elements (rods, cruciforms or other forms), and degradation (or depletion) of the
effectiveness of control elements;
b) presence and spatial distribution of burnable or soluble absorbers;
c) presence of adjacent, unlike fuel assemblies;
d) composition and geometric layout of fuel in an assembly;
e) dependence of coolant or moderator density upon conditions, or their spatial dependence;
f) depletion dependent conditions, including previous power history, coolant-density history, control-
element history, and soluble-absorber history of fuel assemblies;
g) presence of materials or conditions, or both, outside the core, such as the core shroud in a boiling water
reactor (BWR);
h) presence of sources, detectors, structural materials, and experimental devices;
i) spatial variations in temperatures (e.g. fuel temperature, coolant temperature, and moderator
temperature);
k) spatial and temporal variations of important nuclides (e.g. xenon, samarium, and actinides).

5.3 Multigroup cross sections
5.3.1 Basic data
The primary sources of basic nuclear data that are used for the generation of multigroup constants
are evaluated data sets. Examples of these are the evaluated nuclear data file Version B (ENDF, /B, see
References [11], [12] and [17]), Japanese evaluated nuclear data library (JENDL, see References [15] and [16]),
Russian library of evaluated neutron reaction data (BROND, see References [14] and [18]), joint evaluated
fission and fusion library (JEFF, and see Reference [13]), and Chinese evaluated nuclear data library (CENDL,
see Reference [19]) evaluated data sets. The properties and criteria for selecting these sources of basic
[1]
nuclear data are specified in ANSI/ANS 19.1-2019 .
5.3.2 Preparation of multigroup constants
5.3.2.1 Processing evaluated data sets
When preparing multigroup constants directly from evaluated data sets or from processed continuous-
energy data sets, the procedures for the preparation of averaged data sets described in Reference [1] should
be followed. The multigroup constants can be sensitive to the selection of an energy-dependent weighting
spectrum and to the choice of group structure. The smaller the number of energy groups, the greater the
sensitivity will be. Therefore, an estimate of the reactor spectrum is needed and should be obtained from
measurements in identical or similar reactors or from analytical models of neutron slowing-down or source
spectra. Results may be sensitive to the modelling of the spectra.
5.3.2.2 Collapsing application-independent averaged (or multigroup) data sets
The preparation of application-dependent multigroup constants from existing application-independent
multigroup constants or from processed continuous-energy data sets shall entail use of an application-
dependent energy spectrum estimate (see 5.3.3 and 5.4). This procedure employs a weighting spectrum that
is selected to preserve important system-dependent characteristics during the averaging process. These
characteristics usually include reaction rates, and may include other quantities.
5.3.3 System dependent spectrum calculations
The multigroup cross-section set (see 5.3) should be used in the calculation of the neutron energy spectra in
the system under investigation. The energy spectra are established by the geometry, material composition,
and operating conditions of the reactor in an interplay of neutron leakage with reactions such as absorptions
and scattering. The neutron energy spectrum may vary from one region of the core to another and it may be
necessary to compute the spectra for several representative regions of the reactor core.
5.3.4 Choice of cell and supercell
Many reactor cores can be thought of as composed of repeating units called cells, such as a single fuel pin cell
or a fuel assembly cell (this formalism can be extended to absorber pins or water holes), with its associated
structures, coolant, and moderator (where this is distinct from the coolant).
Once a cell is selected, one approach is to compute the spectrum representative of this cell. It is necessary
to inspect the cell and its surroundings to determine if the spectrum in the cell is generated by the cell and
its similar surroundings alone, or if the spectrum in the cell is influenced by parts of the reactor not made of
similar cells. When the spectrum is influenced by regions of the core external to the cell, a supercell may be
defined, and the spectrum characteristic of the supercell is computed. The supercell may be a repeating unit
of the reactor containing noncell materials such as water channels, control-rods, and structural materials.
Other noncell regions such as absorber pins, when present, should be included in the supercell if they
significantly influence the spectrum. For either a cell or a supercell, outer boundary conditions are specified
to be consistent with symmetry assumptions.

5.3.5 Cell environment
The assumption that a reactor is made of an array of similar cells or supercells is, at best, an approximation,
and if the spectrum in the cell is influenced by external regions, these effects should be included in the
spectrum calculations. These effects may be caused by leakage across the cell or supercell boundaries
and thus may be energy and direction dependent. Temperature effects in fuel (e.g. Doppler broadening),
temperature effects in the moderator and/or coolant, and variations in density or composition of coolant
and/or moderator shall be included in the calculation. Corrections for a nonuniform temperature distribution
within the cell should be made, or the temperature distribution should be included in the calculation.
The environment can be explicitly modelled in the case of lattice calculations.
5.3.6 Calculation model
5.3.6.1 General
The calculation model of the cell or supercell often can be considered to have two aspects - the geometric
model and the neutronic transport model, - though the two aspects may not be clearly separable.
5.3.6.2 Geometric model
The geometric model refers to the manner in which the physical configuration of the cell is represented in
the mathematical solution. Geometric approximations may be employed when all aspects of the physical
configuration are not of comparable importance, the primary objective being to reduce the number of
dimensions employed in the solution of the problem or to transform to a more convenient or simplified
geometry. Different geometric approximations may be made concerning the same physical configuration
for different purposes. The choice of geometric models appropriate to the analyses shall be justified and
documented.
In some calculations, one geometric dimension of the model may be dropped, as long as the leakage in the
missing direction is taken into account by the judicious inclusion of a buckling or leakage term that stands as
the surrogate of the missing leakage.
5.3.6.3 Neutronic transport model
Various calculation procedures may be utilized to describe neutron transport phenomena in cell studies.
Different degrees of approximation may be made depending on the nature of the problem and the objectives
of the calculation. A very detailed type of calculation is continuous-energy Monte Carlo. This statistical
procedure follows “histories” of large numbers of individual neutrons. Initially, this technique has served
primarily as a guide to the accuracy of other procedures but may be used in mainstream applications as well.
Other transport models generate numerical solutions (by collision probability methods, for instance) of
the transport equation. Approximations also are introduced in representing energy-transfer kernels. The
analyst shall demonstrate and document that the transport model used is appropriate to the problem under
consideration. For example, the analyst shall demonstrate that the spatial mesh, the order of scattering (P1,
P3, etc.), and the order of quadrature (in Sn methods) are adequate to achieve stated accuracy levels for the
calculated reactivity and reaction rates.
5.4 Collapse to few-groups and spatial homogenization
When performing full reactor calculations, it is usually adequate and desirable to collapse the cross sections
from the multigroup structure into a few-groups. The actual group structure chosen should depend on
the type of calculation that is to use the few-group data and the sensitivity of that calculation to the group
structure.
It is usually desirable to also convert the heterogeneous unit cells into equivalent spatially homogeneous
cells for use in full reactor calculations.

When collapsing cross sections to few-groups and performing spatial homogenization, important system
characteristics – such as reaction rates in a unit-cell, reactivity of the cell and core, or reaction-rate ratios
– should be preserved to the extent practical. This preservation is an attempt to maintain an equivalence
between the many group heterogeneous calculation and the coarser few-group homogeneous calculation.
The actual quantity or quantities preserved and the method of doing this should depend on the intended use
of the few-group data.
The calculation used in the collapse shall include or approximately account for all important effects of space
and energy that cannot be adequately modelled in the calculations to follow, such as self-shielding and
spectrum dependence on surrounding materials.
The cross sections of each nuclide present, to a significant degree, shall be retained individually whenever
calculations of individual reaction rates are to be carried out. These cross sections should also be the
starting points for depletion calculations (e.g. calculations of changes in nuclide composition with time).
5.5 Calculation of reactivity, reaction rates and neutron flux distributions
5.5.1 Models
The calculations being considered in this subclause have as their objective the computation of a measure
of closeness to criticality of a specified reactor-core configuration, and the reaction rates as a function of
position in the core under a steady-state or quasi steady-state condition. A number of models may be used
for this purpose.
A frequently used measure of closeness to criticality is the effective multiplication factor, k . This is
eff
appropriate, for example, in describing the closeness to criticality of a reactor in its shut-down condition.
However, most steady-state reactor calculations are intended to represent conditions at critical or an
artificial steady-state for the purpose of calculating reactivity and safety margins or reactivity coefficients.
In addition, code-system bias and uncertainties may lead to a non-unity k as a reference point. The
eff
definition of reactivity, ρ, in this document is then (1-1/k ).
eff
For the purpose of discussion in this subclause, it is assumed that cross sections for all regions of the reactor
have been generated in multigroup or few-group homogenized form by the techniques described previously.
A number of models are in common use for performing neutron-flux calculations. Some of these are:
a) solving the diffusion equations by finite-difference, finite-element, or other methods;
b) solving the transport or simplified transport equations by discrete-ordinates or collision-probability
methods or by the method of characteristics;
c) solving the reactor neutron balance equations and reaction rate distributions by nodal or other methods.
Thermohydraulic and thermal models linked with neutronic simulation are needed for core calculation.
The preceding examples of models are by no means exhaustive of models that may be used. However, they
are sufficient to illustrate the variety of methods being used, each of which may have characteristic types of
uncertainties and assumptions.
5.5.2 Uncertainties and assumptions
Usually, the model used to describe neutron transport in the reactor calculation is an approximation to a
more accurate model. For example, diffusion theory is an approximation to transport theory. Thus, there
will be some error because of the model per se. In the implementation of a model via a computer program, it
is common for additional approximations to be made. For example, it may be assumed that the neutron flux
or current remains constant over small areas or along small line segments. Thus, the solution produced by
the computer program will be an approximation to the solution of the model equations.

The following are examples of many modelling assumptions or approximations that are commonly made,
and that may contribute to a calculation bias and/or to uncertainties:
a) the assumption that neutron flux in the core as a function of all three spatial dimensions may be
represented as the product of functions that separately are a function of only one or two dimensions
(spatial separability) although this assumption is used less extensively today;
b) geometrical transformations used to model the physical situation;
c) the use of artificial boundary conditions within the core (e.g. at the boundaries of heavily-absorbing
control slabs or cruciforms);
d) assumptions of symmetry for configurations that are not precisely symmetric;
e) the choice of a small number of energy groups to represent the neutron energy variation in the core;
f) the assumption of linearity or simplified variation of flux between the spatial points within a spatial
mesh structure, the dimensions of which may be specified somewhat arbitrarily;
g) the choice of a limited number of directions or spherical harmonics moments to represent the angular
variation of the flux in transport or approximate transport solvers;
h) the use of bucklings to simulate leakage effects in the directions not explicitly represented;
i) the assumption that dissimilar media may be homogenized;
j) the use of pre-calculated region-homogenized (typically lattice-homogenized) cross sections at a
predefined power history;
k) the use of interpolation or curve fitting techniques for the calculation of cross sections at local
conditions.
All of the preceding assumptions or approximations are, in principle, amenable to numerical studies aimed
at establishing the deviation of the normally used procedures from more precise solutions of the model
equations. Numerical methods should normally be used only within the range of parameters for which the
biases or uncertainties of the methods are known.
5.6 Calculation of reaction rates in reactor components
When a model that simplifies the physical description is used, means shall be provided to convert the results
of the model calculation into reaction rates in the physical components as required by the application. For the
reaction rate ca
...