ISO 12749-3:2024
(Main)Nuclear energy, nuclear technologies, and radiological protection — Vocabulary — Part 3: Nuclear installations, processes and technologies
Nuclear energy, nuclear technologies, and radiological protection — Vocabulary — Part 3: Nuclear installations, processes and technologies
This document deals with the terminological data used in the standards regarding the standardization and promotion of good practices associated with the planning, design, construction, operation and decommissioning of installations, processes and technologies involving radioactive materials. The vocabulary of nuclear installations, processes and technologies includes fuel cycle, ex-reactor nuclear criticality safety, analytical methodologies, transport of radioactive materials, materials characterization, radioactive waste management and decommissioning. NOTE See Annex A for the methodology used to develop the vocabulary.
Énergie nucléaire, technologies nucléaires et protection radiologique — Vocabulaire — Partie 3: Installations nucléaires, procédés et technologies
General Information
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Standards Content (Sample)
International
Standard
ISO 12749-3
Second edition
Nuclear energy, nuclear
2024-05
technologies, and radiological
protection — Vocabulary —
Part 3:
Nuclear installations, processes and
technologies
Énergie nucléaire, technologies nucléaires et protection
radiologique — Vocabulaire —
Partie 3: Installations nucléaires, procédés et technologies
Reference number
© ISO 2024
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ii
Contents Page
Foreword .iv
Introduction .v
1 Scope . 1
2 Normative references . 1
3 Terms and definitions . 1
3.1 Terms related to nuclear materials .1
3.2 Terms related to nuclear fuels .4
3.3 Terms related to nuclear fuel cycle .6
3.4 Terms related to nuclear criticality safety .8
3.5 Terms related to transport of radioactive material .9
3.6 Terms related to radioactive waste . 12
Annex A (informative) Methodology used in the development of the vocabulary . 17
Bibliography .24
Alphabetical index .25
iii
Foreword
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The procedures used to develop this document and those intended for its further maintenance are described
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This document was prepared by Technical Committee ISO/TC 85, Nuclear energy, nuclear technologies, and
radiological protection.
This second edition cancels and replaces the first edition (ISO 12749-3:2015), which has been technically
revised.
The main changes are as follows:
— addition of new concepts;
— modification of definitions;
— change of sources.
A list of all parts in the ISO 12749 series can be found on the ISO website.
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
This document will provide terms and definitions for concepts associated with nuclear installations,
processes, and technologies. These include specific subjects such as the nuclear fuel cycle; ex-reactor nuclear
criticality safety, analytical methodologies, transport of radioactive materials, characterization of materials,
radioactive waste management, and decommissioning of nuclear installations. Excluded topics are specific
enabling technologies and techniques for non-peaceful applications, sealed sources, radiation processing,
nuclear power plants and research reactors (with regard to nuclear criticality safety while fuel is loaded
in the reactor core). Terminological data are taken from ISO standards developed by ISO/TC 85/SC 5 and
other technically validated documents issued by the International Atomic Energy Agency (IAEA) or other
international organizations.
Unambiguous communication of concepts associated with nuclear installations, processes, and technologies
is crucial to prevent misunderstandings or misinterpretations of terms used in documents developed by
ISO/TC 85/SC 5. In line with the international demand for harmonization of terminology regarding nuclear
and radiological activities, this document will contribute by providing terms and definitions to meet the
requirements of users and industry. It will also improve promotion, knowledge and use of international
standards dealing with nuclear installations, processes and technologies and will help experts developing
technical standards to avoid overlapping and contradiction.
Nuclear fuels for different power reactors are produced according to different designs. However, several
concepts are present in all of them and need to be designated by common terms and described by harmonized
definitions in order to avoid misunderstandings. Difficulties can also arise due to the wide variety of units
of measure employed. Thus, to enhance comprehension as well as comparability, it is advisable to adopt
unified units of measure.
Arrangement of terms and definitions is based on concepts systems that show corresponding relationships
among the various concepts. Such arrangement provides users with a structured view of the nuclear
installations, processes, and technologies sector and will facilitate common understanding of all related
concepts. In addition, concepts systems and conceptual arrangement of terminological data will be helpful
to any kind of user because it will promote clear, accurate, and useful communication.
v
International Standard ISO 12749-3:2024(en)
Nuclear energy, nuclear technologies, and radiological
protection — Vocabulary —
Part 3:
Nuclear installations, processes and technologies
1 Scope
This document deals with the terminological data used in the standards regarding the standardization
and promotion of good practices associated with the planning, design, construction, operation and
decommissioning of installations, processes and technologies involving radioactive materials.
The vocabulary of nuclear installations, processes and technologies includes fuel cycle, ex-reactor nuclear
criticality safety, analytical methodologies, transport of radioactive materials, materials characterization,
radioactive waste management and decommissioning.
NOTE See Annex A for the methodology used to develop the vocabulary.
2 Normative references
There are no normative references in this document.
3 Terms and definitions
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 related to nuclear materials
3.1.1
nuclear material
material containing one or more of the following: plutonium except that with isotopic concentration
exceeding 80 % in Pu; uranium enriched in the isotope 235 or 233; uranium containing the mixture of
isotopes as occurring in nature other than in the form of ore or ore residue
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p. ISBN:
978-92-0-141822-7.]
3.1.2
critical
having an effective neutron multiplication factor equal to unity
[SOURCE: ISO 1709:2018, 3.1]
3.1.3
nuclear criticality
state of a nuclear chain reacting medium when the nuclear chain reaction (3.1.9) is just self-sustaining (or
critical (3.1.2)), i.e. when the reactivity is zero
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p. ISBN:
978-92-0-141822-7]
3.1.4
radionuclide
nuclide which is in an unstable state due to excess of internal energy and which will attain a stable state by
emitting radiation
40 235 238 232
Note 1 to entry: Radionuclides (3.1.4) are either naturally occurring, such as K, U, U, Th and their radioactive
decay products, or produced by activation or other artificial means.
[SOURCE: ISO 12749-1:2020, 3.1.8]
3.1.5
radioactivity
stochastic process whereby nuclei undergo spontaneous disintegration, usually accompanied by the
emission of subatomic particles, or photons
[SOURCE: ISO 12749-1:2020, 3.1.1]
3.1.6
nuclear installation
any nuclear facility subject to authorization that is part of the nuclear fuel cycle (3.3.1), except facilities for
the mining or processing of uranium ores or thorium ores and disposal facilities for radioactive waste (3.6.1)
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p. ISBN:
978-92-0-141822-7]
3.1.7
fissionable nuclide
nuclide capable of undergoing fission by interaction with neutrons of some energy
[SOURCE: LA-11627-MS, Glossary of Nuclear Criticality Terms. Los Alamos National Laboratory, 1989]
238 240
Note 1 to entry: Fissionable nuclides include U, Pu, and others with neutron-energy fission thresholds, in addition
to those nuclides that are fissile.
3.1.8
fissile nuclide
nuclide capable of undergoing fission by interaction with neutrons of any energy
[SOURCE: ISO 1709:2018, 3.4]
Note 1 to entry: The term is usually applied to fission predominantly with slow neutrons. The interpretation of “slow”
may vary but the properties of fissile nuclides are clearly distinct from other fissionable nuclides (3.1.7).
233 235 239 241
Note 2 to entry: Particular examples are U, U, Pu and Pu.
3.1.9
nuclear chain reaction
series of nuclear reactions in which one of the agents necessary to the series is itself produced by the same
reactions
[SOURCE: ISO 1709:2018, 3.8]
3.1.10
fission product
radionuclide (3.1.4) produced by nuclear fission
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p. ISBN:
978-92-0-141822-7]
3.1.11
burnup
loss of fissile material by the fission process that is usually described in terms of the number of fissioned
atoms or on the total energy liberated in unit volume or mass
Note 1 to entry: Burnup fission product (3.1.10) as energy released by fissions is commonly used for NPP
Note 2 to entry: Burnup as number of fissions per unit volume or mass is commonly used for fuel behaviour modelling
and neutronic calculations
Note 3 to entry: Burnup as number of fissions is commonly used for experimental reactors
3.1.12
subcriticality
having or involving a chain reaction that is not self-sustaining
3.1.13
subcriticality limit
limit value of subcriticality dimension (3.1.14) which is respected in order to ensure subcriticality (3.1.12) of a unit
[SOURCE: ISO 21391:2019, 3.8]
3.1.14
subcriticality dimension
controlled geometrical dimension (item dimension or layout dimension) for which a limit shall be respected
to ensure subcriticality (3.1.12) of a unit
[SOURCE: ISO 21391:2019, 3.7, modified — The word "controlled" was moved at the beginning of the
definition.]
3.1.15
neutron leakage
neutrons leaving a fissile system boundary such that they no longer interact with that system
Note 1 to entry: For an array of fissile units, neutron leakage from one unit may or may not interact with other units.
[SOURCE: ISO 1709:2018, 3.7]
3.1.16
neutron absorber
material with which neutrons interact significantly by reactions resulting in their disappearance as free
particles
[SOURCE: ISO 1709:2018, 3.6]
3.1.17
over batching
unintended increase in the quantity of a material that is controlled for nuclear criticality safety (3.4.1) such
that one or more extra discrete quantities are present
[SOURCE: ISO 1709:2018, 3.13]
3.1.18
permeation
passage of a fluid through a solid permeable barrier (even if there are no leaks (3.5.17)) by adsorption-
diffusion-desorption mechanisms
Note 1 to entry: Permeation should not be considered as a release of activity unless the fluid itself is radioactive. In
this document, permeation is applied only to gases.
[SOURCE: ISO 12807:2018, 3.14]
3.1.19
permeation rate
quantity of gases passing through permeable walls per unit time
[SOURCE: ISO 12807:2018, 3.15]
3.1.20
attenuation
physical process based on interaction between a radiation source and matter placed in the path of the
radiation that results in a decrease in the intensity of the emitted radiation
Note 1 to entry: Attenuation experienced in non-destructive analysis (NDA) of waste packages (3.6.12) includes self-
attenuation by the radioactive material itself as well as attenuation effects in the waste matrix (3.6.14), internal
barrier(s) and external container(s).
[SOURCE: ISO 19017:2015, 2.2]
3.1.21
attenuation correction factor
used to correct (compensate) for the effect of attenuation (3.1.20) within an NDA measurement equal to the
ratio between the unattenuated and the attenuated radiation flux
Note 1 to entry: After attenuation correction the measured quantity is considered to be representative of the
unattenuated activity of the radioactive substance measured.
[SOURCE: ISO 19017:2015, 2.3]
Note 2 to entry: A subcritical dimension is a different term, usually referring to a fissile material dimension that
relies on single-parameter control to avoid making a unit critical. Examples are subcritical cylinder (3.5.6) diameter,
subcritical slab thickness and subcritical volume.
Note 3 to entry: The subcriticality (3.1.12) of a unit may be ensured by other types of controls in addition to dimensional
controls (e.g. mass control, density control).
3.2 Terms related to nuclear fuels
3.2.1
nuclear fuel
fissionable nuclear material (3.1.1) in the form of fabricated elements for loading into the reactor core of a
civil nuclear power plant or research reactor
[SOURCE: ISO 12749-1:2020, 3.2.5]
3.2.2
cladding
external layer of material that houses nuclear fuel (3.2.1) and provides the containment (means of
confinement) of radionuclides (3.1.4) produced during fission
Note 1 to entry: Material also provides structural support and protection from chemically reactive conditions (e.g.,
corrosion).
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p.
ISBN: 978-92-0-141822-7, modified — Change “tube of pellets” with “external layer of” and delete “tube of”
and “material” and add Note 1 to entry.]
3.2.3
nuclear fuel pellet
nuclear fuel (3.2.1) in ceramic form and with a cylindrical shape
3.2.4
fuel element
nuclear fuel (3.2.1), its cladding (3.2.2) and any associated components necessary to form a structural entity
Note 1 to entry: Commonly referred to as “fuel rod” in light water reactors.
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation protection and EMERGENCY Preparedness and Response. Vienna: IAEA, 2022. 246 p.
ISBN: 978-92-0-141822-7 modified — Delete the term “rod”.]
Note 2 to entry: In some countries “fuel element” is used as a synonym for “ fuel assembly” (3.2.5)
3.2.5
fuel assembly
set of fuel elements (3.2.4) and associated components which are loaded into and subsequently removed
from a reactor core as a single unit
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p. ISBN:
978-92-0-141822-7.]
Note 1 to entry: In light water reactors “fuel elements” in the definition should be replaced with “fuel rods”.
Note 2 to entry: In some countries “fuel element” is used as a synonym for “fuel assembly”.
Note 3 to entry: In some countries “fuel bundle” is used as a synonym for “fuel assembly”.
3.2.6
sinter
increase the bonding in a mass of powder or a compact by heating below the melting point of the main
constituent
[SOURCE: ASTM B243-22]
3.2.7
specific surface
surface area of one gram of powder, usually expressed in square centimetres
[SOURCE: ASTM B243–22]
3.2.8
theoretical density
density of a material calculated from the number of atoms per unit cell and from the measurement of the
lattice parameters
3.2.9
apparent density
loose bulk density
dry mass per unit volume of a powder obtained by free pouring under specified conditions
[SOURCE: ISO 9161:2019, 3.1]
3.2.10
tap density
dry mass per unit volume of a powder in a container that has been tapped under specified conditions
[SOURCE: ISO 9161:2019, 3.2]
3.2.11
normal stress
component of stress normal to the plane of reference
[SOURCE: ISO 10276:2019, 3.1.10]
3.2.12
bending stress
variable component of normal stress (3.2.11), which might not be linear across the thickness
[SOURCE: ISO 10276:2019, 3.1.1]
3.2.13
membrane stress
component of normal stress (3.2.11) that is uniformly distributed and equal to the average stress across the
thickness of the section under consideration
[SOURCE: ISO 10276:2019, 3.1.9]
3.2.14
linearized stress
sum of the membrane stress (3.2.13) and of the linear component of the bending stress (3.2.12)
[SOURCE: ISO 10276:2019, 3.1.5]
3.2.15
peak stress
maximum stress that occurs in a component by reason of geometry, local discontinuities or local thermal
stress, including the effects, if any, of stress concentration
[SOURCE: ISO 10276:2019, 3.1.12]
3.2.16
scrap
residue that contains sufficient quantities of nuclear material (3.1.1) to be worthy of recovery
3.3 Terms related to nuclear fuel cycle
3.3.1
nuclear fuel cycle
operations associated with the production of nuclear energy
Note 1 to entry: The nuclear fuel cycle includes the following stages:
a) mining and processing of uranium or thorium ores;
b) conversion;
c) enrichment of uranium;
d) manufacturing of nuclear fuel (3.2.1);
e) uses of the nuclear fuel (3.2.1);
f) reprocessing (3.3.2) and recycling (3.3.6) of spent nuclear fuel (3.3.9);
g) temporary radioactive material storage (3.6.15) of spent nuclear fuel and radioactive waste (3.6.1) from fuel
fabrication and reprocessing (3.3.2) and disposal of spent nuclear fuel (open fuel cycle) (3.3.13) or high-level waste
(closed fuel cycle (3.3.14));
h) any related research and development activities;
i) transport (3.5.1) of radioactive material;
j) all waste management activities (including decommissioning relating to operations associated with the
production of nuclear energy).
Note 2 to entry: Reactor operation and other activities at a reactor site are not addressed in this part of ISO 12749, but
are addressed in ISO 12749-5.
[SOURCE: ISO 12749-1:2020, 3.2.6]
3.3.2
reprocessing
process or operation, the purpose of which is to extract radioactive isotopes from spent nuclear fuel (3.3.9)
for further use
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p. ISBN:
978-92-0-141822-7]
3.3.3
reprocessing plant
installation for the chemical separation of nuclear material (3.1.1) from fission products (3.1.10) following
dissolution of spent nuclear fuel (3.3.9)
3.3.4
solvent extraction
process used to selectively extract actinide elements from aqueous medium
3.3.5
PUREX process
chemical process used to separate plutonium and uranium from fission products (3.1.10) and from each other
by means of solvent extraction (3.3.4) with tributyl phosphate (TBP)
3.3.6
recycling
use, for the fabrication of nuclear fuel (3.2.1) of fissionable materials separated from spent nuclear fuel (3.3.9)
3.3.7
depleted uranium
uranium containing a lesser mass percentage of U than natural uranium
[11]
Note 1 to entry: This usage is specific to the Transport Regulations .
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p. ISBN:
978-92-0-141822-7]
Note 2 to entry: Bulk of the by-product (~96 %) from uranium enrichment.
3.3.8
used nuclear fuel
nuclear fuel (3.2.1) that has been activated in the fission process of a nuclear reactor core
3.3.9
spent nuclear fuel
nuclear fuel (3.2.1) that has been burned in the core of a nuclear reactor and is no longer efficient to maintain
its specific nuclear service
3.3.10
uranium concentrate
uranium ore concentrate
product with a high concentration in uranium obtained by physical and chemical treatments of the ores,
requiring further refinement before it is suitable for nuclear use
EXAMPLE Yellowcake, concentrated crude oxide U O .
3 8
3.3.11
remediation
measures taken for contaminant removal, containment or monitored non-intervention at a contaminated
site to reduce exposure to radiation, and for improvement in the environmental and/or economic value of
the contaminated site
Note 1 to entry: Remediation of a site does not necessarily imply a restoration of the site to pristine condition.
[SOURCE: ISO 18557:2017, 3.22]
3.3.12
remediation objectives
objectives, including those related to technical, administrative and legal requirements
EXAMPLE Technical objectives: residual contamination concentrations, engineering performance.
[SOURCE: ISO 18557:2017, 3.23]
Note 1 to entry: Remediation (3.3.11) of a site does not necessarily imply a restoration of the site to pristine condition.
[SOURCE: ISO 18557:2017, 3.22]
3.3.13
open fuel cycle
mining, processing, conversion, enrichment of uranium, nuclear fuel (3.2.1) fabrication, reactor operation,
electrical generation or other energy products, storage of spent nuclear fuel (3.3.9), disposal and final end
states for all waste
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p. ISBN:
978-92-0-141822-7]
3.3.14
closed fuel cycle
mining, processing, conversion, enrichment of uranium, nuclear fuel (3.2.1) fabrication, reactor operation,
electrical generation or other energy products, reprocessing (3.3.2) to recover fissile material, storage of
reprocessed fissile material, disposal (for highly radioactive fission products (3.1.10)) and final end states for
all waste
[SOURCE: IAEA. IAEA Nuclear Safety and Security Glossary. Terminology Used in Nuclear Safety, Nuclear
Security, Radiation Protection and Emergency Preparedness and Response. Vienna: IAEA, 2022. 246 p. ISBN:
978-92-0-141822-7]
3.4 Terms related to nuclear criticality safety
3.4.1
nuclear criticality safety
protection against the consequences of a nuclear criticality accident (3.4.5), preferably by prevention of the
accident and responses to such accidents should they occur
[SOURCE: ISO 1709:2018, 3.10]
3.4.2
nuclear criticality safety programme
arrangements and procedures implemented in order to ensu
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