ISO 12749-5:2026

International
Standard
ISO 12749-5
Second edition
Nuclear energy, nuclear
2026-08
technologies, and radiological
protection — Vocabulary —
Part 5:
Nuclear reactors
Énergie nucléaire, technologies nucléaires, et radioprotection —
Vocabulaire —
Partie 5: Réacteurs nucléaires
Reference number
© ISO 2026
All rights reserved. Unless otherwise specified, or required in the context of its implementation, 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
CP 401 • Ch. de Blandonnet 8
CH-1214 Vernier, Geneva
Phone: +41 22 749 01 11
Email: copyright@iso.org
Website: www.iso.org
Published in Switzerland
ii
Contents Page
Foreword .iv
Introduction .v
1 Scope . 1
2 Normative references . 1
3 Terms and definitions . 1
3.1 General terms related to nuclear reactors .1
3.2 Terms related to nuclear fission .3
3.3 Terms related to nuclear reactor types .6
3.4 Terms related to nuclear power plants .8
3.5 Terms related to research reactors.10
3.6 Terms related to special-purpose reactors . 13
3.7 Terms related to nuclear projects .14
3.8 Terms related to structures, systems and components .17
3.9 Terms related to nuclear power plant operations and support . 22
3.10 Terms related to nuclear safety in reactor design and operation . 26
Annex A (informative) Methodology used in the development of the vocabulary .32
Bibliography .44
Alphabetical index .45

iii
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).
ISO draws attention to the possibility that the implementation of this document may involve the use of (a)
patent(s). ISO takes no position concerning the evidence, validity or applicability of any claimed patent
rights in respect thereof. As of the date of publication of this document, ISO had not received notice of (a)
patent(s) which may be required to implement this document. However, implementers are cautioned that
this may not represent the latest information, which may be obtained from the patent database available at
www.iso.org/patents. ISO shall not be held responsible for identifying any or all such patent rights.
Any trade name used in this document is information given for the convenience of users and does not
constitute an endorsement.
For an explanation of 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 www.iso.org/iso/foreword.html.
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 12749-5:2018), which has been technically
revised.
The main changes are as follows:
— the addition of the following sentence in the Scope: “The scope of this document does not cover nuclear
fusion reactors.”
— the updated of all the sources of the terminological entries.
— the addition of the heading “Terms related to nuclear fission”.
— the deletion of the heading “Terms related to multiple-purpose reactors”.
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 provides terms and definitions for main concepts in the whole field of nuclear reactor
science, technology, engineering, projects and operations, excluding quantitative data. Terminological data
are taken from ISO standards developed by TC 85/SC 6, from other technically validated documents issued
by international organizations, especially IAEA and IEC, while several definitions have been drafted by
WG 1 experts based on their experience and after detailed discussions on concept characteristics, the best
wording for their designations and definitions, as well as the most important links between concepts.
In most cases, international consensus exists among the communities of nuclear reactor specialists world-
wide, on the most relevant concepts in the nuclear reactor field. Nevertheless, clear, and unambiguous terms
for these concepts are also needed.
The foregoing should also be considered together with the fact that many people are involved in the broad
nuclear reactor field, having different scopes and levels of scientific and technical knowledge, and frequently
having very specific activities within that broad field. Thus, there can be different understandings and
assumptions about concepts. Hence, the result could be a poor communication that might lead into
unexpected, different risky situations or consequences, if a conceptual difference is behind.

v
International Standard ISO 12749-5:2026(en)
Nuclear energy, nuclear technologies, and radiological
protection — Vocabulary —
Part 5:
Nuclear reactors
1 Scope
This document encompasses the collection of terms, definitions, notes to entry and examples corresponding
to nuclear reactors, excluding quantitative data. It provides the minimum essential information for
each nuclear reactor concept represented by a single term. Full understanding of concepts requires
background knowledge of the nuclear field. It is intended to facilitate communication and promote common
understanding.
The scope of this document does not cover nuclear fusion reactors.
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 General terms related to nuclear reactors
3.1.1
nuclear installation
nuclear facility subject to authorization that is part of the nuclear fuel cycle, except facilities for the mining
or processing of uranium ores or thorium ores and disposal facilities for radioactive 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]
3.1.2
nuclear reactor
special device having an inventory of nuclear fuel material containing fissionable nuclides (3.2.16) and,
depending on the technology, can have neutron moderating, neutron absorbing and cooling materials, all of
them geometrically arranged in a particular neutron multiplicative configuration (3.2.10) designed and built
for having the capability of initiating, maintaining and extinguishing a controlled, self-sustaining nuclear
fission (3.2.1) chain reaction, under adequate safety conditions

3.1.3
research reactor
nuclear reactor (3.1.2) used mainly for the generation and utilization of neutron flux and ionizing radiation
for research and other purposes
Note 1 to entry: Experimental facilities associated with the reactor and storage, handling and treatment facilities for
radioactive material on the same site, which are directly related to safe operation of the research reactor, are included.
Note 2 to entry: Facilities commonly known as critical assemblies are included.
[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. By splitting the definition into a definition and two notes to entry.]
3.1.4
nuclear power plant
NPP
nuclear reactor (3.1.2) installation with the main purpose of producing heat energy, converting it to
electricity, or both
Note 1 to entry: Nuclear power plant is a nuclear reactor (3.1.2) or reactors together with all structures (3.8.2), systems
and components (3.8.13) necessary for ensuring safe and secure operation in compliance with safeguards agreements
and for the production of power, i.e. heat or electricity.
3.1.5
nuclear island
part of the nuclear power plant (3.1.4) that consists of the containment (3.8.3), auxiliary and fuel buildings
3.1.5.1
reactor core
part of a nuclear reactor (3.1.2), where the fuel assemblies sustaining the fission chain are located
Note 1 to entry: In most cases, moderator (3.1.5.1.2) and coolant are also included in the reactor core.
3.1.5.1.1
core reflector
material placed around the reactor core (3.1.5.1), totally or partially enveloping it, in order to scatter most of
leaking neutrons back into the core, improving neutron economy
Note 1 to entry: Since most of nuclear reactors (3.1.2), have a preferred vertical cylinder geometry, they have three
different reflectors: upper axial, lower axial and radial.
Note 2 to entry: The most used materials as reflectors for slow or thermal neutrons (3.2.18), are H O (light or ordinary
water), D O (heavy water), Be (beryllium), C (carbon, graphite). The most used material for fast neutrons (3.2.19) is
stainless steel.
3.1.5.1.2
moderator
material that reduces neutron energy by scattering without appreciable capture
Note 1 to entry: Materials of prime concern are those containing light nuclei with large scattering cross sections and
relatively low absorption cross sections (e.g. water, hydrocarbons, polyethylene, oils and graphite).
3.1.5.2
reactor internals
any of the different structural parts inside the nuclear reactor (3.1.2), covering various functions, either as
part of the nuclear reactor itself or as part of reactor-associated systems
Note 1 to entry: Some of these functions are:
a) to direct inlet and outlet primary coolant (3.1.5.4) flow and its distribution among all reactor heat sources, similar
structure (3.8.2) for other fluids inside the reactor (like liquid moderator (3.1.5.1.2) and reflectors or both);

b) to provide in-core locations and protection for in-core instrumentation (3.8.5.1.1) and for elements and components
(3.8.13) related to reactivity (3.2.12);
c) to provide power control and safety shutdown.
3.1.5.3
reactor vessel
enveloping structure (3.8.2) for housing all parts of a reactor core (3.1.5.1) and supporting reactor internals
(3.1.5.2)
Note 1 to entry: Main parts usually allocated inside the reactor vessel, are as follows:
a) the reactor core (3.1.5.1) and its reflectors;
b) all structures (3.8.2) and tubes of reactor internals (3.1.5.2).
Note 2 to entry: The reactor vessel additionally serves as confining structure (3.8.2) for the primary coolant (3.1.5.4).
Moreover, depending on the reactor technology, this vessel is also part of a high-pressure boundary for the primary
coolant, mainly encompassing the reactor core (3.1.5.1).
3.1.5.4
primary coolant
fluid circulating through the reactor core (3.1.5.1), in order to extract the heat produced by the nuclear
fission reactions
Note 1 to entry: The primary coolant normally removes instantaneous fission and decay heat generated in the reactor
core (3.1.5.1) and can remove abnormal excessive heat. It circulates within the primary heat transport system (3.8.9)
and transfers all reactor thermal power to a final heat sink. Such thermal power transfer can be either “direct”, or
more usually “indirect”, by means of chained circuits circulating appropriate fluids, between the nuclear reactor
(3.1.2) and that final heat sink.
Note 2 to entry: The primary or main coolant is always a fluid and then it can be, either a liquid in single phase, or a
boiling two-phase fluid, or a gas. The most widely employed are:
a) liquid-state ordinary or light water (H O) at atmospheric or high pressure;
b) liquid-state heavy water (D O) at atmospheric or high pressure;
c) boiling H O at high pressure;
d) CO or He gases.
Note 3 to entry: The final heat sink usually can be either: the atmosphere, a river, a lake, a sea or an ocean. In a nuclear
power plant (3.1.4) one of such sinks shares its function with a turbine-generator (3.4.7) set, where part of the total
thermal power is converted to kinetic power and immediately to electric power.
Note 4 to entry: For molten salt reactors, a similar role to primary coolant is played by a fluid sometimes defined as
intermediate coolant.
3.2 Terms related to nuclear fission
3.2.1
nuclear fission
process by which a nucleus undergoes a partition in two, infrequently in three, main fission fragments,
releasing energy
Note 1 to entry: There are two types of nuclear fission: “spontaneous” fission and “induced” fission.
Note 2 to entry: The nucleus usually has a high mass number A, together with an intermediate or low average binding-
energy-per-nucleon; hence, an inherent instability exists, and the fission fragments are usually highly unstable.
Note 3 to entry: According to their capability for undergoing fission, a nucleus and its associated nuclide can be
qualified as fissionable or eventually fissile.

3.2.2
induced nuclear fission
nuclear fission (3.2.1) initiated by a nucleus when an external colliding particle is absorbed
Note 1 to entry: The absorption of the external colliding particle, usually a neutron, generates a strong increase in the
compound nucleus internal energy and, hence, increases the compound nucleus instability, favouring a large energy
release by means of nuclear fission (3.2.1).
3.2.3
spontaneous nuclear fission
nuclear fission (3.2.1) produced in a nucleus, having an inherent instability, that develops itself in a purely
stochastic way and without intervention of any external colliding particle
3.2.4
fissile nuclide
nuclide capable of undergoing fission by interaction with neutrons of any energy
Note 1 to entry: There is no fission energy threshold to undergo fission.
Note 2 to entry: The term is usually applied to fission predominantly with slow neutrons. The interpretation of “slow”
can vary but the properties of fissile nuclides are clearly distinct from other fissionable nuclides (3.2.16).
233 235 239 241
Note 3 to entry: Particular examples are U, U, Pu and Pu.
3.2.5
fertile nuclide
nuclide that becomes a fissile nuclide (3.2.4) after absorbing a neutron
238 239 240
Note 1 to entry: In practice, the main fertile nuclides are: U (producing the fissile Pu), Pu (producing the fissile
241 232 233
Pu) and Th (producing the fissile U), in all cases after the absorption of one neutron and the fast emission of
some gamma photons.
3.2.6
fission energy
energy released in the fission process, which is primarily in the form of the kinetic energy of the fission
fragments
3.2.7
fission product
radionuclide produced by nuclear fission (3.2.1)
[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.]
3.2.8
delayed fission radiation
gamma, beta radiations or both, in certain cases also alpha radiations released in a stochastic way, from a
fission product (3.2.7)
Note 1 to entry: Every possible fission product decay has extremely diverse half-lives, covering the range around 0,1 s
up to more than a billion years.
Note 2 to entry: These “delayed” released gamma, beta and alpha radiations, after interactions with neighbouring
atoms, are mainly absorbed by the surrounding materials and then finally converted into heat: they are the source of
what is designated as decay-power or decay-heat or residual-heat.

3.2.9
delayed fission neutron
neutron emitted in few particular fission product (3.2.7) decays, typically with half-lives roughly in the range
0,1 s to 1 min, following initiation of a nuclear fission (3.2.1) process
Note 1 to entry: Such decay occurs between two energy levels of a fission product (3.2.7) namely precursor-favouring
a neutron release, hence, the emitted neutron will have a quite defined kinetic energy at its release, typically below
1 MeV.
Note 2 to entry: In a fission neutron population, since delayed neutrons have kinetic evolutions dictated by those rather
long periods, as compared to the extremely fast evolutions of prompt neutrons, the first ones provide an important
contribution to the kinetic control of that neutron population.
3.2.10
neutron multiplicative configuration
geometrical disposition of materials, one or more containing fissionable nuclides (3.2.16), the whole
configuration being capable of maintaining a multiplicative chain reaction of neutron-induced nuclear
fissions (3.2.2)
3.2.11
nuclear chain reaction
successive induced nuclear fissions (3.2.2) initiated by neutrons released, for the most part, from previous
fissions in fissionable nuclides (3.2.16)
Note 1 to entry: The free neutron population in a system is multiplied by fissions releasing several neutrons per fission,
compensating partially or totally, or exceeding the total neutron losses by capture and leakage from the system.
Note 2 to entry: A nuclear chain reaction can be initiated by a pre-existent small neutron population (like that resulting
from photoneutrons in particular materials for this purpose), or by neutrons released by spontaneous fissions, or by
a specific “neutron source” emitting neutrons that initiates its own repetition chain reaction, a fissionable nucleus
absorbs a neutron and fissions spontaneously, releasing additional neutrons.
3.2.12
reactivity
ρ
for a nuclear chain reacting medium:
ρ = (k – 1)/k
eff eff
where k (3.2.15) is the ratio between the number of fissions in two succeeding generations (later to earlier)
eff
of the chain reaction
Note 1 to entry: A measure of the deviation from criticality (3.2.15) of a nuclear chain reacting medium, such that
positive values correspond to a supercritical state and negative values correspond to a subcritical state.
[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 (Retrieved: 4 July 2024)]
3.2.13
poison
substance used to reduce reactivity (3.2.12), typically in a reactor core (3.1.5.1), by virtue of its high neutron
absorption cross-section
[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. (Retrieved: 4 July 2024)]

3.2.14
effective neutron multiplication factor
k
eff
ratio between current to previous generation of neutron population in multiplicative medium
Note 1 to entry: The total number of produced neutrons per unit time includes all prompt and delayed neutrons
released in fissions; while, the total number of lost neutrons are the sum of all absorbed neutrons (in fission and
capture reactions), plus all leaking neutrons escaping.
Note 2 to entry: There are three possible circumstances: a) k > 1: fission power increases in time; b) k = 1: fission
eff eff
power remains constant in time; c) k < 1: fission power decreases in time. When k > 1, system is called supercritical;
eff eff
when k = 1, system is called critical; and finally, when k < 1, then system is called subcritical
eff eff
3.2.15
criticality
state of a nuclear chain reacting medium when the nuclear chain reaction (3.2.11) is just self-sustaining or
critical, i.e. when the reactivity (3.2.12) 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 (Retrieved: 4 July 2024)]
3.2.16
fissionable nuclide
nuclides that are capable of supporting a self-sustaining nuclear chain reaction (3.2.11) with neutrons of any
speed
[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 (Retrieved: 4 July 2024)]
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.
[SOURCE: LA-11627-MS, Glossary of Nuclear Criticality Terms. Los Alamos National Laboratory, 1989.]
3.2.17
prompt fission neutron
neutron emitted from a fission fragment in a stochastic way, with high kinetic-energy just following
initiation of a nuclear fission (3.2.1) process
Note 1 to entry: The number of prompt neutrons released per fission, is stochastic as indicated, with an average value
in the range 2,5 to 3 for most of concerned nuclides.
3.2.18
thermal neutron
neutron with a kinetic energy of about 0,025 eV, which is the energy corresponding to the most probable
speed of 2 000 m/s at 293 K
3.2.19
fast neutron
neutron with kinetic energy greater than 1 keV
Note 1 to entry: Neutrons with energy levels greater than 0,025 eV and up to 1 keV are termed epithermal or resonance
neutrons.
3.3 Terms related to nuclear reactor types
3.3.1
nuclear reactor
(See 3.1.2)
3.3.1.1
power reactor
nuclear reactor (3.1.2) designed to produce electrical power or propulsion
3.3.1.2
research reactor
(See 3.1.3)
3.3.1.3
special-purpose reactor
nuclear reactor (3.1.2) designed with a particular target
EXAMPLE Prototype reactors (3.6.1.3), demonstration reactors (3.6.1.2), naval propulsion reactors, desalination
reactors (3.6.1.5), material testing reactors (3.5.1.4), hydrogen production reactors.
3.3.1.4
breeder reactor
nuclear reactor (3.1.2) designed to produce more fissile nuclides (3.2.4) than it uses, having the conversion
ratio greater than one
Note 1 to entry: The conversion ratio (CR) or conversion coefficient measures the ability of a nuclear reactor (3.1.2) to
238 239
transfer a fertile nuclide (3.2.5), e.g., U, into a fissile (3.2.4), e.g., Pu.
[SOURCE: Koelzer, W. Glossary of Nuclear Terms. Germany, 2013]
3.3.1.5
converter reactor
nuclear reactor (3.1.2) designed to produce less fissile nuclides (3.2.4) than it uses, having a conversion ratio
smaller than one
3.3.1.6
transmutation reactor
nuclear reactor (3.1.2) designed for the purpose of eliminating partially the radioactive wastes contained in
other reactors spent nuclear fuel
3.3.1.7
fast reactor
nuclear reactor (3.1.2) designed and operated mainly using a predominantly fast neutron (3.2.19) energy
spectrum
Note 1 to entry: The main contribution to fission power typically from neutrons with energies above 100 keV.
3.3.1.8
thermal reactor
nuclear reactor (3.1.2) designed and operated using predominantly thermal neutron (3.2.18) reactions
Note 1 to entry: The main contribution to fission power typically from neutrons with energies below 1 eV.
3.3.1.9
gas cooled reactor
GCR
nuclear reactor (3.1.2) that uses gas as primary coolant (3.1.5.4)
Note 1 to entry: A gas cooled reactor can be either a gas-cooled thermal reactor (3.3.1.8) or a gas cooled fast reactor
(3.3.1.7).
Note 2 to entry: The gas is usually helium (He) or carbon dioxide (CO ).
3.3.1.10
light water reactor
LWR
nuclear thermal reactor (3.3.1.8) cooled and moderated by light water

3.3.1.11
heavy water reactor
nuclear thermal reactor (3.3.1.8) that uses heavy water as its moderator (3.1.5.1.2)
Note 1 to entry: Heavy water is a moderator and thus permits the use of unenriched uranium as fuel.
[SOURCE: NRC. Nuclear Regulatory Commission Glossary.http:// www .nrc .gov/ readig -rm/ basic -ref/ glossary
.html. (Retrieved 16 April 2025). modified. By adding “nuclear thermal” at the beginning of the definition
and splitting the definition into a definition and one note to entry.]
3.3.1.12
liquid metal reactor
liquid metal fast reactor
LMFR
nuclear reactor (3.1.2) using as coolant a liquid metal, like sodium, lead, or some alloy
3.3.1.13
boiling water reactor
BWR
nuclear reactor (3.1.2) with water as a coolant and as a moderator (3.1.5.1.2), boiling in the core
Note 1 to entry: In a boiling water reactor the generated heat is removed from the core by evaporation.
[SOURCE: Koelzer, W. & FORSCHUNGSZENTRUM KARLSRUHE GMBH TECHNIK UND UMWELT. Dictionary of
nuclear power. upd. ed.; Lexikon zur Kernenergie. Germany: N. p., 2011.]
3.3.1.14
pressurized water reactor
PWR
nuclear power reactor (3.3.1.1) with water as a coolant and as a moderator (3.1.5.1.2), sufficiently high
pressurized to remain liquid in the core
Note 1 to entry: In a boiling water reactor (3.3.1.13), the generated heat is removed from the core by evaporation.
[SOURCE: Koelzer, W. & FORSCHUNGSZENTRUM KARLSRUHE GMBH TECHNIK UND UMWELT. Dictionary of
nuclear power. upd. ed.; Lexikon zur Kernenergie. Germany: N. p., 2011.]
Note 2 to entry: Where water is uses as moderator (3.1.5.1.2) and as a coolant to remove heat from the core.
3.3.1.15
pressurized heavy water reactor
PHWR
nuclear thermal reactor (3.3.1.8) cooled and moderated by heavy water (D O), having a pressurized D O
2 2
coolant to be kept permanently in the liquid state
3.4 Terms related to nuclear power plants
3.4.1
nuclear power plant
NPP
(See 3.1.4)
3.4.1.1
single-unit nuclear power plant
single-unit NPP
nuclear power plant (3.1.4) having a single electric production line

3.4.1.2
multiple-unit nuclear power plant
multiple-unit NPP
nuclear power plant (3.1.4) having on a common site, two or more electric production lines
Note 1 to entry: The electricity generation can be either almost identical or different between them, though usually
sharing various support or auxiliary installations, systems, supplies or both.
3.4.2
reference unit power
maximum (electrical) power that could be maintained continuously throughout a prolonged period of
operation (3.9.1) under reference ambient conditions
Note 1 to entry: The power value is expressed in units of megawatt (electrical).
[SOURCE: IAEA. Glossary of terms in PRIS reports – PRIS: Power reactor information system. (Retrieved: 22
July, 2024), modified. By splitting the definition into a definition and a note to entry. https:// www .iaea .org/
PRIS/ Glossary .aspx. ]
3.4.3
energy availability factor
ratio of the energy that the available capacity could have produced during a specified period, to the energy
that the reference unit power (3.4.2) could have produced during the same period
[SOURCE: IAEA. Glossary of terms in PRIS reports – PRIS: Power reactor information system. (Retrieved: 22
July, 2024) https:// www .iaea .org/ PRIS/ Glossary .aspx.]
3.4.4
NPP unit
each unit represents a separate reactor (nuclear island (3.1.5) and BOP (3.4.5)) capable of being operated
Note 1 to entry: In the case of dual- or multi-unit plants, a unit can operate independently of the state of completion
or operating condition of any other units co-located on the same site but in different containment (3.8.3)/confinement
buildings, even though the units can have some shared or common systems.
[SOURCE: IAEA. Terms for describing advance nuclear power plants. IAEA Nuclear Energy Series No. NR-T-
1.19. Vienna: IAEA, 2023. 20p. ISBN 978-92-0-145923-7, modified by splitting the definition into a definition
and a note to entry.]
3.4.5
balance of plant
BOP
part of the nuclear power plant (3.1.4) that consists of a set of a main turbine generator (3.4.7) unit and all
associated systems, structures (3.8.2) and components (3.8.13) necessary to produce electric power
3.4.6
turbine island
part of the nuclear power plant (3.1.4) that consists of the turbine building
[SOURCE: IAEA. Construction technologies for nuclear power plants. IAEA Nuclear Energy Series No. NP-T-2.5.
Vienna: IAEA, 2011. 207 p. ISBN 978-92-0-119510-4]
3.4.7
turbine generator
equipment that produces electrical energy and which type is defined by the turbine-driving fluid
3.4.8
steam turbine balance of plant
steam turbine BOP
balance of plant (3.4.5) type that produces electric energy associated to a nuclear steam supply system (3.4.9)

3.4.9
nuclear steam supply system
NSSS
nuclear reactor (3.1.2) and all its associated systems, structures (3.8.2) and components (3.8.13), necessary
to reliably and safely produce steam at a suitable pressure to drive a steam turbine generator (3.4.10) unit
Note 1 to entry: The nuclear steam supply system can have different thermal power sources, such as PWR (3.3.1.14),
PHWR (3.3.1.15), BWR (3.3.1.13), GCR (3.3.1.9) and LMFR (3.3.1.12).
3.4.10
steam turbine generator
equipment in a BOP (3.4.5) that produces electrical energy, where the generator is driven by a steam turbine
supplied by a NSSS (3.4.9)
3.4.11
steam condenser
equipment in a steam turbine BOP (3.4.8), aimed at receiving the exhausting low-pressure turbine steam,
bringing this fluid into the liquid state by cooling
3.4.12
gas turbine balance of plant
gas turbine BOP
balance of plant (3.4.5) type that produces electric energy associated with a nuclear high temperature gas
supply system (3.4.13)
3.4.13
nuclear high temperature gas supply system
NHTGSS
nuclear reactor (3.1.2) and all its associated systems, structures (3.8.2) and components (3.8.13), necessary
to reliably and safely produce gas at high temperature and pressure to drive a gas turbine generator (3.4.14)
unit
Note 1 to entry: The nuclear high temperature gas supply system can have different thermal power sources, such as
GCR (3.3.1.9), LMFR (3.3.1.12) and GCFR.
3.4.14
gas turbine generator
equipment in a BOP (3.4.5) that produces electrical energy, where the generator is driven by a gas turbine
supplied by a NHTGSS (3.4.13)
3.4.15
gas cooler
equipment in a gas turbine BOP (3.4.12), aimed at lowering the gas temperature
Note 1 to entry: In a gas turbine BOP (3.4.12) the NHTGSS (3.4.13) feeds a gas turbine generator (3.4.14) that produces
electrical power. To make this possible, the heat that is not converted to electrical or mechanical power, is transferred
to the secondary cooling fluid.
3.4.16
nuclear island
(See 3.1.5)
3.5 Terms related to research reactors
3.5.1
research reactor
(See 3.1.3)
3.5.1.1
training and education reactor
type of nuclear reactor (3.1.2) dedicated to training students to operate, maintain, regulate, and improve
nuclear reactors and other facilities, aimed from training in very specific reactor concepts up to broad
education in the nuclear reactor field observing the nuclear fission (3.2.1) process and the interaction of
radiation with matter
Note 1 to entry: The application fields can be medical, biological, electronic, engineering, industrial as well as related
to specific reactor concepts or nuclear fuel cycle strategies services, technological developments, industrial production
purposes or both.
3.5.1.2
multiple-purpose reactor
nuclear reactor (3.1.2) designed to fulfil several main purposes together, providing different services, except
electric energy supply
EXAMPLE Reactors that produce radioisotopes, provide irradiation boxes and positions, irradiated material
studies, neutron beams for research and development work, and personnel training.
Note 1 to entry: The application fields can be medical, biological, electronic, engineering, industrial as well as related
to specific reactor concepts or nuclear fuel cycle strategies services, technological developments, industrial production
purposes or both.
3.5.1.3
radioisotope production reactor
nuclear reactor (3.1.2) aimed at producing radioisotopes for medical and industrial applications, by neutron
capture or fission reactions
Note 1 to entry: Usually, the “neutron transmutation doping” (3.5.3) is an application included in radioisotope
production reactors.
3.5.1.4
material testing reactor
nuclear reactor (3.1.2) aimed at irradiating materials, with the purpose of testing and qualifying the
behaviour of the irradiated materials
Note 1 to entry: The materials tested usually include nuclear fuels.
Note 2 to entry: Material testing reactor (MTR) is a generic name given to those reactors that use MTR fuel.
3.5.2
radioisotope production
activity to produce radioactive nuclides from neutron reactions occurring in the material contained in a
target that is being exposed to a neutron radiation field
3.5.3
neutron transmutation doping
technique that consists in the irradiation of a target material with a neutron flux, mostly thermal, in order to
produce dispersed impurities in the target material
Note 1 to entry: Main application is on silicon target, where the impurities produced are the dopants of the
semiconductor.
Note 2 to entry: Special care shall be taken regarding the uniformity of the distribution of the introduced dopants.

3.5.4
boron neutron capture therapy
BNCT
technique for radiation therapy based on B (n, alpha) reaction in cancer tissue or cell which, before the
neutron radiation, has absorbed a chemical compound containing B isotope
Note 1 to entry: Non-radioactive boron is added to pharmaceutical agents that are selective to be located in specific
tissues (tumour localizing drugs). Boron, more precisely the B isotope, has high cross section (that means high
probability) to absorb neutrons; therefore, a nuclear reaction is produced due to neutron capture by boron with its
resulting rupture into a lithium ion and an alpha particle. Both resulting particles have high energy and produce
ionization very close to the reaction point, destroying the tissue cells.
3.5.5
neutron activation analysis
NAA
technique of elemental analysis based on the identification and measurement of intrinsic radiation of nuclide
formed in sample material by neutron bombarding
3.5.5.1
prompt gamma-ray neutron activation analysis
PGAA
technique of elemental analysis based on the identification and measurement of intrinsic radiation emitting
from sample material by neutron continuously bombarding
Note 1 to entry: As a consequence of absorbing neutrons, a gamma ray is emitted by the sample in a very short time
(in comparison with the times for the gamma rays obtained in the neutron activation analysis (3.5.5)). The energy and
intensity of the measured peaks lead to determination of the isotopic composition.
Note 2 to entry: This is a technique useful for trace elements that cannot be well detected by neutron activation analysis
(3.5.5).
3.5.6
neutron diffraction technique
technique by which the study of the scattering of neutrons upon a target reveals information about the
structure of the target material
Note 1 to entry: In order that it is possible to study materials structure, the energy of the neutrons shall be in an
energy range so that their wavelength is comparable to the typical dimensions of the materials being studied. The cold
sources serve this purpose of taking the neutrons to some desired energy range for a variety of materials of interest.
Note 2 to entry: This technique is also available not only in research reactors (3.1.3) but in ad hoc research facilities
like neutron spallation sources.
Note 3 to entry: The technique has a variety of fields of application like, material science, molecular biology and
archaeology.
3.5.7
rod type fuel
pin type fuel
rod of nuclear fuel, comprising a cylindrical tube (cladding) containing an assembly of fuel pellets (typically
UO ) sealed both ends to form a structural unit
Note 1 to entry: In light water reactors “fuel rod (FR)” is used as “rod type fuel”.
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.]
3.5.8
fuel assembly
FA
fuel bundle
set of fuel elements and associated components (3.8.13) which are loaded into and subsequently removed
from a reactor core (3.1.5.1) 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 (3.3.1.10) “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”.
3.5.8.1
annular fuel assembly
fuel assembly (3.5.8) that consists of annular fuel rods, which are clad in their inner and outer surfaces, and
associated components (3.8.13) necessary to form a structural unit
3.6 Terms related to special-purpose reactors
3.6.1
special-purpose reactor
(See 3.3.1.3)
3.6.1.1
zero-power reactor
special-purpose reactor (3.3.1.3) with a negligible thermal power, behaving as an almost identical copy, in its
neutron parameters, of some reference reactor core (3.1.5.1) design
Note 1 to entry: “Critical facility”, “critical mock-up” and “neutronic maquette” are all possible alternative designations
of the same concept.
Note 2 to entry: Zero-power reactors are generally built to assess a novel core design for a power reactor (3.3.1.1),
adequately reproducing the materials and geometry of the designed core.
3.6.1.2
demonstration reactor
special-purpose reactor (3.3.1.3) aimed at characterizing a novel nuclear reactor (3.1.2) design, reproducing
its most relevant and distinctive features and its main nuclear, process system (3.8.10)and safety systems
(3.8.7)
Note 1 to entry: Demonstration reactors should be able to operate in a range of parameters allowing the validation of
the design in the novel features. Usually, their reactor power is scaled down to a fraction of the foreseen power output,
optimizing the investment needs.
Note 2 to entry: The design to be assessed by a demonstration reactor is generally an innovative power reactor (3.3.1.1)
design that is planned for future commercial deployment.
3.6.1.3
prototype reactor
special-purpose reactor (3.3.1.3) aimed at characterizing and qualifying a novel nuclear reactor (3.1.2) design,
reproducing the design of a future series of nuclear reactors, though usually with additional capabilities,
margins and instrumentation and control devices
Note 1 to entry: The design to be qualified by a prototype reactor is generally an innovative power reactor (3.3.1.1)
design that is planned for commercial deployment.
Note 2 to entry: Additional instrumentation allows performance of additional engineering and qualification tests
during commissioning, and during operation (3.9.1).

Note 3 to entry: Prototype reactors can be scaled to a fraction of the foreseen power output through a detailed
similarity analysis and can lack auxiliary systems that are not related to reactor perf
...