ISO 12749-3:2015

INTERNATIONAL ISO
STANDARD 12749-3
First edition
2015-08-15
Nuclear energy, nuclear technologies,
and radiological protection —
Vocabulary —
Part 3:
Nuclear fuel cycle
Énergie nucléaire, technologies nucléaires et protection
radiologique — Vocabulaire —
Partie 3: Cycle de combustibles nucléaires
Reference number
©
ISO 2015
© ISO 2015, Published in Switzerland
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ii © ISO 2015 – All rights reserved

Contents Page
Foreword .iv
Introduction .v
1 Scope . 1
2 Structure of the vocabulary . 1
3 Terms and definitions . 2
3.1 General terms related to nuclear fuel cycle . 2
3.2 Terms related to conversion and enrichment . 6
3.3 Terms related to fuel fabrication . 7
3.4 Terms related to fuel characteristics . 8
3.5 Terms related to transport of radioactive material .10
3.6 Terms related to reprocessing .12
3.7 Terms related to radioactive waste .12
3.8 Terms related to decommissioning .17
3.9 Terms related to nuclear criticality safety.19
Annex A (informative) Methodology used in the development of the vocabulary .21
Annex B (informative) Alphabetical index .33
Bibliography .36
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
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ISO collaborates closely with the International Electrotechnical Commission (IEC) on all matters of
electrotechnical standardization.
The procedures used to develop this document and those intended for its further maintenance are
described in the ISO/IEC Directives, Part 1. In particular the different approval criteria needed for the
different types of ISO documents should be noted. This document was drafted in accordance with the
editorial rules of the ISO/IEC Directives, Part 2 (see www.iso.org/directives).
Attention is drawn to the possibility that some of the elements of this document may be the subject of
patent rights. ISO shall not be held responsible for identifying any or all such patent rights. Details of
any patent rights identified during the development of the document will be in the Introduction and/or
on the ISO list of patent declarations received (see www.iso.org/patents).
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assessment, as well as information about ISO’s adherence to the WTO principles in the Technical
Barriers to Trade (TBT) see the following URL: Foreword - Supplementary Information
The committee responsible for this document is ISO/TC 85, Nuclear energy, nuclear technologies, and
radiological protection.
This first edition cancels and replaces ISO 921:1997, of which it forms the subject of a technical revision.
ISO 12749 consists of the following parts, under the general title Nuclear energy, nuclear technologies,
and radiological protection:
— Part 2: Radiological protection
— Part 3: Nuclear fuel cycle
— Part 4: Dosimetry for radiation processing
The following parts are under preparation:
— Part 5: Reactors
iv © ISO 2015 – All rights reserved

Introduction
This part of ISO 12749 will provide terms and definitions for nuclear fuel cycle concepts dealing with
specific subjects such as fuel fabrication, fuel characteristics, and nuclear criticality safety and with
transport and radioactive waste related topics, excluding reactors operations. Terminological data are
taken from ISO standards developed by TC 85/SC 5 and other technically validated documents issued
by international organizations.
Unambiguous communication of nuclear energy concepts is crucial taking into account the relevant
implications that may arise from misunderstandings with regard to equipment and materials involved
in the standards dealing with any subject regarding nuclear energy activities. 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. In another nuclear fuel technology subfield, difficulties arise due to
the wide variety of units employed to measure the fuel burnout level. Thus, to enhance comprehension,
it is advisable to adopt unified measure units.
Conceptual arrangement of terms and definitions is based on concepts systems that show corresponding
relationships among nuclear energy concepts. Such arrangement provides users with a structured view
of the nuclear energy sector and will facilitate common understanding of all related concepts. Besides,
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.
INTERNATIONAL STANDARD ISO 12749-3:2015(E)
Nuclear energy, nuclear technologies, and radiological
protection — Vocabulary —
Part 3:
Nuclear fuel cycle
1 Scope
This part of ISO 12749 lists unambiguous terms and definitions related to nuclear fuel cycle concepts
in the subject field of nuclear energy, excluding reactor operations. It is intended to facilitate
communication and promote common understanding.
2 Structure of the vocabulary
The terminology entries are presented in the conceptual order of the English preferred terms. The
structure of each entry is in accordance with ISO 10241-1:2011.
All the terms included in this part of ISO 12749 deal exclusively with nuclear fuel cycle. When selecting
terms and definitions, special care has been taken to include the terms that need to be defined, that is
to say, either because the definitions are essential to the correct understanding of the corresponding
concepts or because some specific ambiguities need to be addressed.
The notes appended to certain definitions offer clarification or examples to facilitate understanding of
the concepts described. In certain cases, miscellaneous information is also included, for example, the
units in which a quantity is normally measured, recommended parameter values, references, etc.
According to the title, the vocabulary deals with concepts belonging to the general nuclear energy
subject field within which concepts in the nuclear fuel cycle sub-subject field are taken into account.
See Annex A for the methodology used to develop the vocabulary.
nuclear fuel cycle
3.1 General terms related to
nuclear fuel cycle
3.8
3.7
3.3 3.4 3.5 3.6
3.2
Terms
Terms
Terms Terms related Terms Terms related
Terms
related related to
related to to fuel related to to
related to
decommi-
to radio-
fuel characteristics transport of reprocessing
conversion
ssioning
active
fabrication radioactive
and
waste
material
enrichment
3.9
Terms related to nuclear
criticality safety
3 Terms and definitions
3.1 General terms related to nuclear fuel cycle
3.1.1
nuclear fuel
fissionable nuclear material used in a reactor core or intended for use in a reactor core
3.1.1.1
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)  manufacture of nuclear fuel (3.1.1);
e)  uses of the nuclear fuel;
f)  reprocessing (3.1.1.1.2.2) and recycling (3.1.1.1.2.3) of spent fuel;
g)  temporary radioactive material storage (3.1.1.1.2.1)of spent fuel and radioactive waste (3.7.1) from fuel
fabrication (3.1.1.1.1.3) and reprocessing (3.1.1.1.2.2) and disposal of spent nuclear fuel (3.1.1.1.5) [open fuel cycle
(3.1.1.7)] or high-level waste (closed fuel cycle (3.1.1.8)];
h)  any related research and development activities;
i)  transport of radioactive material;
j)  all waste management (3.7.7) activities [including decommissioning (3.8.1]) 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 to be addressed in ISO 12749-5.
[SOURCE: Adapted from IAEA Safety Glossary, 2007 Edition, modified — By splitting the definition into
a definition and a note.]
3.1.1.1.1
front end
steps of the nuclear fuel cycle (3.1.1.1) ending with fuel introduction into the reactor core
3.1.1.1.1.1
nuclear material conversion
modification of the chemical composition of nuclear material so as to facilitate its further use or
processing; in particular, to provide feed material for enrichment of isotopes of interest and/or reactor
fuel fabrication (3.1.1.1.1.3)
Note 1 to entry: To produce material for fuel fabrication (3.1.1.1.1.3), the following are examples of conversion
that can be carried out: U O or UF to uranium dioxide (UO ), U or Pu nitrate to oxide, or U or Pu oxides to metal.
3 8 6 2
[SOURCE: IAEA Safeguards Glossary, 2001 Edition, modified — By splitting the definition into a
definition and note 1 to entry.]
2 © ISO 2015 – All rights reserved

3.1.1.1.1.2
isotope enrichment
isotope separation process by which the fractional abundance of a specified isotope in an element is
increased such as increasing the abundance of U relative to natural uranium (3.1.1.2) or increasing
the abundance of the D O in water
Note 1 to entry: Usually, the term will be “enrichment”.
3.1.1.1.1.2.1
enriched fuel
fuel made with uranium that has been modified by increasing the abundance of the fissile isotope U
3.1.1.1.1.3
fuel fabrication
process for manufacturing fuel elements (3.3.6) or other reactor components containing nuclear material
Note 1 to entry: Manufacturing process includes nuclear material conversion (3.1.1.1.1.1), storage, and physic-
chemical analyses of materials.
[SOURCE: IAEA Safeguards Glossary, 2001 Edition]
3.1.1.1.2
back end
steps of the nuclear fuel cycle (3.1.1.1) beginning with the final removal of the fuel from the reactor core
Note 1 to entry: The processes can include radioactive material storage (3.1.1.1.2.1)at or away from reactor,
reprocessing (3.1.1.1.2.2), recycling (3.1.1.1.2.3), conditioning and disposal.
[SOURCE: IAEA-TECDOC-1613 “Nuclear fuel cycle information system”, 2009, modified — By splitting
the definition into a definition and note 1 to entry.]
3.1.1.1.2.1
radioactive material storage
holding of radioactive sources, spent nuclear fuel (3.1.1.1.5), or radioactive waste (3.7.1) in a facility that
provides for containment with the intention of retrieval
[SOURCE: IAEA Safety Glossary 2007]
3.1.1.1.2.2
reprocessing
process or operation of extracting fission products (3.1.5) from spent nuclear fuel (3.1.1.1.5) to enable
reuse of the nuclear fuel (3.1.1)in a reactor
3.1.1.1.2.3
recycling
use, for the fabrication of nuclear fuel (3.1.1), of fissionable materials (3.1.3) separated from spent nuclear
fuel (3.1.1.1.5)
3.1.1.1.2.3.1
mixed oxide fuel
MOX fuel
mixture of oxides of different fissionable elements
Note 1 to entry: In the nuclear fuel cycle (3.1.1.1), MOX is interpreted as mixed uranium and plutonium oxides
unless otherwise specified.
3.1.1.1.2.4
encapsulation
encasement of radioactive contaminants in a suitable material for final disposal
3.1.1.1.3
burnup
average energy released by a defined region of the fuel during its irradiation
Note 1 to entry: This region could be a complete fuel assembly (3.3.6.1) or some part of the assembly. Burnup
is commonly expressed as energy released per mass of initial fissionable actinides (3.1.8) (uranium only for
this part of ISO 12749). Units commonly used are expressed in megawatt day per metric ton of initial uranium
(MWd/t) or gigawatt day per metric ton of initial uranium (GWd/t).
[SOURCE: ISO 27468:2011, 3.4]
3.1.1.1.4
used nuclear fuel
fuel that has been activated in the fission process of a nuclear reactor core
3.1.1.1.5
spent nuclear fuel
fuel that has been burned in the core of a nuclear reactor and is no longer efficient to maintain its
specific nuclear service
3.1.1.2
natural uranium
elemental uranium containing the naturally occurring uranium isotopes (approximately 99,28% U,
235 234
0,72% U by mass, and small amount of U)
[SOURCE: Adapted from IAEA Safety Glossary, 2007]
3.1.1.2.1
depleted uranium
uranium containing U fractional abundance less than that of natural uranium (3.1.1.2)
Note 1 to entry: Depleted uranium is the complement product to enriched uranium (3.1.1.2.2) where in the former,
U mass fraction is higher than that of natural uranium (3.1.1.2).
[SOURCE: Adapted from IAEA Safety Glossary, 2007]
3.1.1.2.2
enriched uranium
uranium containing a greater mass fraction or percentage of U than in natural uranium (3.1.1.2)
[SOURCE: IAEA Safety Glossary, 2007 Edition, modified — By adding “fraction or” before “percentage”
and replacing “0,72%” with “in natural uranium”.]
3.1.1.3
uranium 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
[SOURCE: ISO 921:1997, modified — The word “abundance” has been changed to “concentration”.]
EXAMPLE Yellowcake, concentrated crude oxide U O
3 8.
3.1.1.4
nuclear criticality
state of a nuclear chain reacting medium when the chain reaction is just self-sustaining
[SOURCE: IAEA Safety Glossary 2007, modified — The phrase “(or critical), i.e. when the reactivity is
zero” has been removed as being unnecessary to the definition and not defined in ISO 12749.]
3.1.1.5
nuclear criticality safety
protection against the consequences of a nuclear criticality accident (3.1.1.6) preferably by prevention of
the accident and responses to such accidents should they occur
4 © ISO 2015 – All rights reserved

3.1.1.6
nuclear criticality accident
nuclear criticality excursion
release of energy as a result of accidentally producing a self-sustaining or divergent fission chain reaction
[SOURCE: LA— 11627-MS DE90 000884, Glossary of Nuclear Criticality Terms]
3.1.1.7
open fuel cycle
once-through fuel cycle
nuclear fuel cycle (3.1.1.1) excluding recycling (3.1.1.1.2.3) of actinide (3.1.8) nuclides from used nuclear
fuel (3.1.1.1.4)
3.1.1.8
closed fuel cycle
nuclear fuel cycle (3.1.1.1) including recycling (3.1.1.1.2.3) of actinide (3.1.8) nuclides from used nuclear
fuel (3.1.1.1.4)
Note 1 to entry: The fuel cycle can be “closed” in various ways, for example, by the recycling of enriched uranium
(3.1.1.2.2) and plutonium through thermal reactors (thermal recycle) by the re-enrichment of the uranium
recovered as a result of spent fuel reprocessing (3.1.1.1.2.2)or by the use of plutonium in a fast breeder reactor.
3.1.2
fissile nuclide
nuclide capable of undergoing fission by interaction with any energy neutrons
3.1.3
fissionable material
material capable of undergoing fission by interaction with neutrons of some neutron energy range
3.1.4
fertile nuclide
nuclide which is not itself fissile, but can be converted into a fissile nuclide (3.1.2) by irradiation in a reactor
Note 1 to entry: There are two basic fertile nuclides, uranium-238 and thorium-232. When these fertile nuclides
capture neutrons, they are converted into fissile plutonium-239 and uranium-233, respectively.
3.1.5
fission product
nuclide produced from nuclear fission or from subsequent radioactive decay of such a nuclide
[SOURCE: ISO 27468:2011, 3.9, modified — By adding “or from subsequent radioactive decay of such a
nuclide” in the definition.]
3.1.6
moderator
material that has high potential for significantly reducing the energy of a free neutron
Note 1 to entry: A moderator may be important for different reasons, e.g. increasing the fission probability [of
fissile nuclide (3.1.2)], increasing the neutron absorption probability (non-fissile actinide (3.1.8) nuclides and
many other nuclides), and for obtaining a specific neutron energy spectrum for irradiation.
3.1.7
nuclear grade
material of a quality adequate for use in nuclear application
3.1.8
actinide
element with atomic number in the range from 89 to 103
Note 1 to entry: Many actinides are produced during the irradiation due to neutron capture and/or decay of
other actinides. The corresponding nuclides are all neutron producers and some are net (considering neutron
production and absorption) neutron producers in a slow neutron energy spectrum.
3.2 Terms related to conversion and enrichment
3.2.1
nuclear material conversion
see 3.1.1.1.1.1
3.2.2
isotope enrichment
see 3.1.1.1.1.2
3.2.3
empty UF cylinder
UF cylinder containing a heel (3.2.8) in quantities equal to or less than those specified in the
documents in force
[SOURCE: Adapted from ISO 7195:2005, 3.3]
3.2.4
maximum allowable working pressure
MAWP
maximum value of UF cylinder design gauge pressure (rounded up to two significant figures) at the
maximum value of UF cylinder design temperature
[SOURCE: ISO 7195:2005, 3.5]
3.2.5
minimum design metal temperature
minimum value of design metal temperature at the maximum value of UF cylinder design pressure to
meet ASME Code requirements
[SOURCE: ISO 7195:2005, 3.6]
3.2.6
tare mass
mass of the cleaned UF cylinder including its service equipment and its permanently attached
structural features
Note 1 to entry: The standard value of the mass tolerance is ±0,1 %.
[SOURCE: Adapted from IAEA TECDOC 608 (1991)]
3.2.7
effective threads
threads that are capable of providing reasonable engagement in mating threads; the first effective
thread at a run out begins one thread length below the run out scratch
[SOURCE: ISO 7195:2005, 3.2]
3.2.8
heel
residual amount of UF and non-volatile reaction products of uranium, uranium daughters (if the UF
6 6
cylinder has contained irradiated uranium) fission products (3.1.5), and transuranic elements
[SOURCE: ISO 7195:2005, 3.4]
6 © ISO 2015 – All rights reserved

3.3 Terms related to fuel fabrication
3.3.1
fuel fabrication
see 3.1.1.1.1.3
3.3.2
presintering
heating of a compact (3.3.3.1) at a temperature below the normal final sintering
temperature, for example, to increase the ease of handling or shaping the compact or to remove a
lubricant or binder (3.3.3.2) before sintering (3.3.3)
3.3.2.1
sinterable powder
powder in which the bonding of adjacent surfaces of particles can be accomplished by heating
3.3.3
sintering
process to form a metallic bond among particles and characterization of sintered
compacts (3.3.3.1)
[SOURCE: ASTM B243-11]
Note 1 to entry: The objective is to increase the density, the grain size, and the mechanical strength of the fuel
pellets (3.3.4).
3.3.3.1
compact
briquet
object produced by the compression of a powder, generally while confined in a die, eventually with the
addition of a binder (3.3.3.2)
3.3.3.2
binder
cementing medium
Note 1 to entry: The binder is either a material added to the powder to increase the strength of the compact
(3.3.3.1) and that is expelled during sintering (3.3.3) or a material (usually of relatively low melting point) added
to a powder mixture for the specific purpose of cementing together powder particles that alone would not sinter
into a strong body.
[SOURCE: ASTM B243–04a, modified — By splitting the description into a definition and a note.]
3.3.4
fuel pellet
small body of fuel, often cylindrical, formed by powder metallurgy processes
Note 1 to entry: The pellet may or may not have been sintered following compaction.
3.3.5
cladding
external layer of material applied to nuclear fuel (3.1.1) or other material to contain radioactive products
Note 1 to entry: Material also provides protection from a chemically reactive environment.
3.3.6
fuel element
nuclear fuel (3.1.1), its cladding (3.3.5), and any associated components necessary to form a
structural entity
Note 1 to entry: Commonly referred to as “fuel rod” in light water reactors.
3.3.6.1
fuel assembly
set of fuel elements (3.3.6) and associated components which are loaded into and subsequently removed
from a reactor core as a single unit
[SOURCE: IAEA Safety Glossary, 2007]
3.3.7
scrap
residues that contain sufficient quantities of nuclear material to be worthy of recovery
3.3.8
burnable absorber
burnable poison
neutron absorbing nuclides added to the fuel assembly (3.3.6.1)
Note 1 to entry: Burnable absorbers are used as an additive with the purpose of reducing the reactivity of the
fresh nuclear fuel (3.1.1).
3.4 Terms related to fuel characteristics
3.4.1
particle size
controlling linear dimension of an individual particle as determined by analysis with sieves or other
suitable means
[SOURCE: ASTM B243-11]
3.4.1.1
particle size distribution
percentage by weight or by number of each fraction into which a powder sample has been classified
with respect to sieve number or micrometres
[SOURCE: ASTM B243-11]
3.4.2
theoretical density
density of a material calculated from the number of atoms per unit cell and measurement of the
lattice parameters
3.4.3
bulk density
mass of a quantity of a bulk solid divided by its total volume
[SOURCE: ASTM D653-11]
3.4.4
tap density
density of a powder in a container that has been tapped under specified conditions
[SOURCE: ISO 9161:2004, 3.2]
3.4.5
apparent density
density of a powder obtained by free pouring under specified conditions
[SOURCE: ISO 9161:2004, 3.1]
3.4.6
equivalent boron content
EBC
concentration of natural boron affording the same neutron absorption as the specific impurity element
8 © ISO 2015 – All rights reserved

3.4.6.1
equivalent boron content factor
EBC factor
product of ratio of the atomic mass of natural boron to that of a specified impurity element and ratio of
the thermal neutron absorption cross section of the impurity to that of boron
 
A  σ 
B i
EBCfactor =  
 
  
A σ
 t  B 
where
A is the atomic mass of boron;
B
A is the atomic mass of impurity;
i
σ is the thermal neutron absorption cross section of boron;
B
σ is the thermal neutron absorption cross section of impurity.
i
[SOURCE: ASTM C1233-09]
3.4.7
total equivalent boron content
TEBC
sum of the individual equivalent boron content (3.4.6) (EBC) values
3.4.8
working reference material
WRM
reference material with documented traceability used routinely as a calibration standard, as a
measurement control standard, or for the qualification of a measurement method
3.4.9
certified reference material
CRM
reference material accompanied by documentation issued by an authoritative body and providing one or
more specified property values with associated uncertainties and traceabilities using valid procedure
[SOURCE: JCGM 200:2012, International vocabulary of metrology – Basic and general concepts and
associated terms (VIM)]
3.4.10
porosity
ratio of the volume of voids to the total volume of a material
Note 1 to entry: Porosity is usually expressed as a percentage.
3.4.11
specific surface
surface area of one gram of powder usually expressed in square centimetres
[SOURCE: ASTM B243–11]
3.4.12
autoradiography
image that is produced by the radiation emitted from radioactive material and recorded on a
photographic film, plate, emulsion, or solid-state detector
3.4.13
pyrohydrolysis
decomposition of material by the combined action of heat and water vapour
3.5 Terms related to transport of radioactive material
3.5.1
packaging
one or more receptacles and any other components or materials necessary for the
receptacles to perform the containment and other safety functions
[SOURCE: UN Recommendations on the transport of dangerous good, 17th Revised edition, 2011]
3.5.1.1
protective packaging
outer packaging (3.1.8)or device used to provide additional protection to an inner
container during transport
3.5.2
package
complete product of the packing operation consisting of the packaging (3.5.1) and its contents prepared
for transport
[SOURCE: Adapted from IAEA Transport Regulations SSR-6 – 2012 Edition and UN Recommendations
on the transport of dangerous good, 18th Revised edition, 2013]
3.5.3
trunnion
cylindrically shaped projection on a packaging (3.5.1) attached by various means and used for lifting,
tie-down (3.5.8), supporting, or tilting packages (3.5.2) from horizontal and vertical modes
[SOURCE: ISO 10276:2010, 3.1.26]
3.5.3.1
welded trunnion
cylindrically shaped projection on a packaging (3.5.1) directly secured to the packaging by welding
[SOURCE: ISO 10276:2010, 3.1.30]
3.5.4
trunnion system
assembly of trunnion (3.5.3) and components to the packaging (3.5.1) including the trunnion attachment
components (3.5.5.1) to the packaging and the internal threads in the packaging (3.5.1) body as appropriate
[SOURCE: ISO 10276:2010, 3.1.29, modified]
3.5.4.1
primary trunnion system
trunnion system (3.5.4) provided as a primary means for the tie-down (3.5.8), supporting, or lifting of
packages (3.5.2)
[SOURCE: ISO 10276:2010, 3.1.12]
3.5.4.2
secondary trunnion system
trunnion system (3.5.4) provided as an additional or alternative means for the tie-down (3.5.8),
supporting, or lifting of packages (3.5.2)
[SOURCE: ISO 10276:2010, 3.1.17]
3.5.5
trunnion attachment
method of attaching the trunnion (3.5.3) (e.g. welding, bolting, interference fitting and bolting, or any
combination of these methods)
[SOURCE: ISO 10276:2010, 3.1.27]
10 © ISO 2015 – All rights reserved

3.5.5.1
trunnion attachment components
trunnion attachment (3.5.3) components excluding the trunnion (3.5.3), e.g. bolts, threads in the
packaging body, baseplates, etc.
[SOURCE: ISO 10276:2010, 3.1.28]
3.5.6
removable trunnion
cylindrically shaped projection on a package (3.5.2) secured by non-permanent methods, e.g. bolting
[SOURCE: ISO 10276:2010, 3.1.14]
3.5.7
transport cycle
complete round-trip journey of a package (3.5.2) between two successive loadings
[SOURCE: ISO 10276:2010, 3.1.24]
3.5.8
tie-down
securing of the package (3.5.2) to the transport conveyance
[SOURCE: IAEA Safety Glossary, 2007.]
3.5.9
trans-shipment
change of conveyance at any time during transport
3.5.10
gross mass
maximum mass of a package (3.5.2) fitted including the ancillaries (shock absorbers,
neutron shields, covers, transport frame as appropriate, etc.), as presented fully laden for transport
[SOURCE: IAEA Transport Regulations SSR-6 – 2012 Edition and UN Recommendations on the transport
of dangerous good, 18th Revised edition, 2013]
3.5.11
maintenance schedule
maintenance document that gives, in appropriate detail, the applicable frequency/periodicity of
maintenance items and details of methods to be employed
[SOURCE: ISO 10276:2010, 3.1.15]
3.5.12
periodic inspection
inspection of the trunnion system (3.5.4) at predetermined intervals during the “in-
service” life of the packaging (3.5.1)
[SOURCE: ISO 10276:2010, 3.1.9]
3.5.13
periodic testing
testing at predetermined intervals of the trunnion system (3.5.4) provided as a primary
means for the lifting, tie-down (3.5.8), supporting, or lifting of packages (3.5.2)
[SOURCE: ISO 10276:2010, 3.1.10]
3.5.14
independent competent organization
organization administratively and managerially separate from the designers, manufacturers,
or users of the subject package (3.5.2) constituted of specialized experts or an insurance organization
used to verify, oversee, witness, or check
[SOURCE: ISO 10276:2010, 3.1.3]
3.6 Terms related to reprocessing
3.6.1
reprocessing
see 3.1.1.1.2.2
3.6.2
reprocessing plant
installation for the chemical separation of nuclear material from fission products (3.1.5) following
dissolution of spent fuel
3.6.3
PUREX process
chemical process used in a reprocessing plant (3.6.2) to separate plutonium and uranium from fission
products (3.1.5) and from each other by means of solvent extraction (3.6.4) with tributylphosphate (TBP)
[SOURCE: ISO 921:1997]
3.6.4
solvent extraction
process used to selectively extract actinide (3.1.8)elements from aqueous medium
3.7 Terms related to radioactive waste
3.7.1
radioactive waste
material for which no further use is foreseen that contains or is contaminated with radionuclides
[SOURCE: Adapted from IAEA Safety Glossary, 2007 Edition]
Note 1 to entry: For legal and regulatory purposes, waste is considered to be radioactive if the concentrations or
activities are greater than clearance levels as established by the regulatory body.
3.7.2
waste generator
operating organization of a facility or activity that generates waste
[SOURCE: Adapted from IAEA Safety Glossary 2007, modified — By simplifying the definition.]
Note 1 to entry: For convenience, the scope of the term waste generator is sometimes extended to include
whoever currently has the responsibilities for the waste.
3.7.3
waste acceptance criteria
quantitative or qualitative criteria specified by the regulatory body or specified by an operator and
approved by the regulatory body for radioactive waste (3.7.1) to be accepted by the operator of a
repository (3.7.9.1) for disposal or by the operator of a storage facility for radioactive material storage
(3.1.1.1.2.1)
Note 1 to entry: Waste acceptance criteria might include, for example, restrictions on the activity concentration
or total activity of particular radionuclides (or types of radionuclide) in the waste or criteria concerning the
waste form (3.7.6) or packaging (3.5.1) of the waste.
12 © ISO 2015 – All rights reserved

[SOURCE: Adapted from IAEA Safety Glossary 2007, modified — By splitting the wording into a
definition and a note.]
3.7.4
radioactive waste characterization
determination of the physical, chemical, and radiological properties of the waste to establish the need
for further adjustment, treatment or conditioning, or its suitability for further handling, processing,
radioactive material storage (3.1.1.1.2.1), or disposal
[SOURCE: IAEA Safety Glossary 2007]
3.7.4.1
representative sample
sample which reflects the characteristics of the population from which it was drawn
3.7.4.2
composite sample
mixture of samples from different containers such that the mass ratio of the samples is equal to the
ratio of the radioactive waste (3.7.1)masses contained in the containers
[SOURCE: ISO 921:1997, modified]
EXAMPLE Series of samples taken over a given period of time and weighted by collection rate or a combined
sample consisting of a series of discrete samples taken over a given period of time and mixed according to a
specified weighting factor such as stream flow or collection rate.
[SOURCE: ISO 21238:2007, 2.6]
3.7.4.3
reference source
radionuclide sealed in a suitable containment of which the radioactive characteristics are determined
by comparison with a reference material
[SOURCE: ISO 14850-1:2004, 2.1]
3.7.4.4
scaling factor
factor to calculate the radioactivity of a difficult-to-measure nuclides (3.7.4.6) through measurement of a
key nuclide (3.7.4.5) serving as a surrogate
Note 1 to entry: The factor is derived through independent experimental determinations and/or through
mathematical construct relating the radioactive properties of the nuclides.
3.7.4.5
key nuclide
gamma-emitting nuclide whose radioactivity is correlated with that of difficult-to-measure nuclides
(3.7.4.6) and can be readily measured directly by non-destructive assay means
Note 1 to entry: Also called “easy-to-measure nuclide” or “marker nuclide”.
60 137
EXAMPLE Co and/or Cs.
[SOURCE: ISO 21238:2007, 2.2]
3.7.4.6
difficult-to-measure nuclide
nuclide whose radioactivity is difficult to measure directly from the outside of the waste packages
(3.7.5) by non-destructive assay means
EXAMPLE Alpha emitting nuclides, beta emitting nuclides, and characteristic X-ray emitting nuclides.
[SOURCE: ISO 21238:2007, 2.1]
3.7.4.7
source volume
volume in m taken up by the matrix (or by the waste) in which the radioactivity is distributed
[SOURCE: ISO 14850-1:2004, 2.2]
3.7.5
waste package
product of conditioning that includes the waste form (3.7.6) and any container(s) and internal barriers
(3.7.6) (e.g. absorbing materials and liner) as prepared for handling, transport, storage, and/or disposal
[SOURCE: Adapted from IAEA Safety Glossary 2007]
3.7.5.1
overpackaging
secondary (or additional) outer container for one or more waste packages (3.7.5) used for handling,
transport, radioactive material storage (3.1.1.1.2.1), and/or disposal
[SOURCE: IAEA Safety Glossary 2007]
3.7.5.2
outermost container
outer shell of a waste package (3.7.5)
[SOURCE: ISO 6962:2004, 3.4]
Note 1 to entry: The vessel into which the waste form (3.7.6) is placed for handling, transport, radioactive material
storage (3.1.1.1.2.1), and/or eventual disposal; also, the outer barrier (3.7.9.2) protecting the waste from external
intrusions. The waste container is a component of the waste package (3.7.5). For example, molten high-level waste
glass would be poured into a specially designed container (canister) where it would cool and solidify. Note that the
term waste canister is considered to be a specific term for a container for spent fuel or vitrified high-level waste.
[SOURCE: IAEA Safety Glossary 2007]
3.7.5.3
radioactive waste mockup
package (3.5.2) consisting of a container and of well-characterized materials representative of a matrix
[SOURCE: ISO 14850-1:2004, 2.5]
3.7.5.4
reference package
mockup containing reference sources (3.7.4.3) in a well-defined configuration
[SOURCE: ISO 14850-1:2004, 2.6]
3.7.6
waste form
waste in its physical and chemical form after treatment or conditioning prior to packaging (3.5.1) and
which is a component of the waste package (3.7.5)
[SOURCE: ISO 6962:2004, 3.3]
3.7.6.1
dry active waste
solid waste generated in various waste streams in nuclear facilities including protective clothing,
replaced equipment, parts, plastics, polyvinyl chloride sheets, and high-efficiency particulate air filters
removed during plant operation and maintenance
[SOURCE: ISO 21238:2007, 2.11]
14 © ISO 2015 – All rights reserved

3.7.6.2
homogeneous waste
radioactive waste (3.7.1) that shows an essentially uniform distribution of activity and physical contents
[SOURCE: ISO 21238:2007, 2.12]
3.7.6.3
heterogeneous waste
radioactive waste (3.7.1) that does not meet the definition of homogeneous waste (3.7.6.2) including solid
components and mixtures of solid components such as dry active waste (3.7.6.1) and cartridge filters
[SOURCE: ISO 21238:2007, 2.13]
3.7.6.4
waste waters
liquids from a nuclear plant or other nuclear installation the activity of which is lower than the levels
permitted by the regulatory body
3.7.7
waste management
all administrative and operational activities involved in the handling, pretreatment, treatment,
conditioning, transport, radioactive material storage (3.1.1.1.2.1), and disposal of radioactive waste (3.7.1)
[SOURCE: IAEA Safety Glossary, 2007]
3.7.8
waste predisposal
any waste management (3.7.7) steps carried out prior to disposal such as pretreatment, treatment,
conditioning, radioactive material storage (3.1.1.1.2.1), and transport activities
Note 1 to entry: Predisposal is used as a contraction of “pre-disposal radioactive waste management [(3.7.7)]” not
a form of disposal.
[SOURCE: IAEA Safety Glossary 2007]
3.7.8.1
durability
ability of a material to exist for a long period of time while retaining its original
qualities and properties
[SOURCE: ISO 16797:2004, 2.3]
3.7.8.2
alteration
superficial chemical modification of a material due to surrounding agents
[SOURCE: ISO 16797:2004, 2.1]
3.7.9
waste disposal
emplacement of waste in an appropriate facility without the intention of retrieval
Note 1 to entry: Some countries use the term disposal to include discharges of effluents to the environment.
[SOURCE: IAEA Safety Glossary 2007]
3.7.9.1
repository
nuclear facility where radioactive waste (3.7.1) is emplaced for disposal
[SOURCE: IAEA Radioactive Waste Management Glossary 2003 Edition]
3.7.9.2
barrier
physical obstruction that prevents or delays the movement of radionuclides or other material between
components in a system, for example, a waste repository (3.7.9.1)
Note 1 to entry: In general, a barrier can be an engineered barrier which is constructed or a natural (or
geological) barrier.
Note 2 to entry: Barriers can also comprise spent nuclear fuel (3.1.1.1.5)containers and other qualified packaging
(3.5.1), stocks, and radioactive material storage (3.1.1.1.2.1) facilities used to confine nuclear material and
nuclear waste.
3.7.9.3
confinement
barrier (3.7.9.2) which surrounds the main parts of a facility containing radioactive materials and which
is designed to prevent or mitigate the uncontrolled release of radioactive material to the environment
Note 1 to entry: Confinement is similar in meaning to containment, but confinement is typically used to refer
to the barriers (3.7.9.2) immediately surrounding the radioactive material, whereas, containment refers to
the additional layers of defence intended to prevent the radioactive materials reaching the environment if the
confinement is breached.
3.7.10
waste processing
any operation that changes the characteristics of radioactive waste (3.7.1) including pretreatment,
treatment, and conditioning
[SOURCE: IAEA Safety Glossary 2007]
3.7.11
waste pretreatment
any or all of the operations prior to radioactive waste treatment (3.7.1.2) such as collection, segregation
(3.7.11.1), chemical adjustment, and decontamination (3.7.11.2)
[SOURCE: IAEA Safety Glossary 2007]
3.7.11.1
segregation
activity where types of radioactive waste (3.7.1) or material (radioactive or exempt) are separated or
are kept separate on the basis of radiological, chemical, and/or physical properties to facilitate waste
handling and/or processing
[SOURCE: IAEA Safety Glossary 2007]
3.7.11.2
decontamination
complete or partial removal of radioactive contamination by a deliberate physical, chemical, or
biological process
[SOURCE: IAEA Safety Glossary 2007]
3.7.12
waste treatment
chemical or physical processing, or both, of radioactive waste (3.7.1) for interim or ultimate disposal
3.7.13
waste conditioning
operations that produce a radioactive waste (3.7.1) package suitable for handling, transport, radioactive
material storage (3.1.1.1.2.1), and/or disposal
Note 1 to entry: Conditioning may include the conversion of the radioactive waste (3.7.1) to a solid waste form
(3.7.6), enclosure of the waste in containers and, if necessary, provision of an overpack.
16 © ISO 2015 – All rights reserved

[SOURCE: IAEA Safety Glossary, 2007]
3.7.13.1
immobilization
prevention of the potential for migration or dispersion of radionuclides by conversion into a stable solid
form
Note 1 to entry: The aim of immobilization is to reduce the potential for migration or dispersion of radionuclides
during handling, transport, radioactive material storage (3.1.1.1.2.1), and/or disposal.
3.7.13.2
solidification
immobilization (3.7.13.1) of gaseous, liquid, or liquid-like materials by conversion into a solid waste form
usually with the intent of producing a physically stable material that is easier to handle and less dispersible
Note 1 to entry: Calcination, drying, cementation, bituminization, and vitrification (3.7.13.3) are some of the
typical ways of solidifying liquid waste.
3.7.13.3
vitrification
process of incorporating radioactive waste (3.7.1) into a glass or glass-like form
Note 1 to entry: Vitrification is commonly applied to the solidificati
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