Description
Under the terms of Article III of its Statute, the IAEA is authorized to establish or adopt standards of safety for protection of health and minimization of danger to life and property, and to provide for the application of these standards.
INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA
ISBN 978–92 –0–103010–8
ISSN 1020–525X
“Governments, regulatory bodies and operators everywhere must
ensure that nuclear material and radiation sources are used
beneficially, safely and ethically. The IAEA safety standards are
designed to facilitate this, and I encourage all Member States to
make use of them.”
Yukiya Amano
Director General
Safety through international standards
IAEA Safety Standards
Disposal of
Radioactive Waste
for protecting people and the environment
No. SSR-5
Specific Safety Requirements
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P1449_cover.indd 1 2011-05-05 08:39:50
IAEA SAFETY RELATED PUBLICATIONS
IAEA SAFETY STANDARDS
Under the terms of Article III of its Statute, the IAEA is authorized to establish
or adopt standards of safety for protection of health and minimization of danger to life
and property, and to provide for the application of these standards.
The publications by means of which the IAEA establishes standards are issued in
the IAEA Safety Standards Series. This series covers nuclear safety, radiation safety,
transport safety and waste safety. The publication categories in the series are Safety
Fundamentals, Safety Requirements and Safety Guides.
Information on the IAEA’s safety standards programme is available at the IAEA
Internet sitehttp://www-ns.iaea.org/standards/
The site provides the texts in English of published and draft safety standards. The
texts of safety standards issued in Arabic, Chinese, French, Russian and Spanish, the
IAEA Safety Glossary and a status report for safety standards under development are
also available. For further information, please contact the IAEA at PO Box 100,
1400 Vienna, Austria.
All users of IAEA safety standards are invited to inform the IAEA of experience
in their use (e.g. as a basis for national regulations, for safety reviews and for training
courses) for the purpose of ensuring that they continue to meet users’ needs.
Information may be provided via the IAEA Internet site or by post, as above, or by
email to [email protected].
OTHER SAFETY RELATED PUBLICATIONS
The IAEA provides for the application of the standards and, under the terms of
Articles III and VIII.C of its Statute, makes available and fosters the exchange of
information relating to peaceful nuclear activities and serves as an intermediary among
its Member States for this purpose.
Reports on safety and protection in nuclear activities are issued as Safety
Reports, which provide practical examples and detailed methods that can be used in
support of the safety standards.
Other safety related IAEA publications are issued as Radiological Assessment
Reports, the International Nuclear Safety Group’s INSAG Reports, Technical Reports
and TECDOCs. The IAEA also issues reports on radiological accidents, training
manuals and practical manuals, and other special safety related publications. Security
related publications are issued in the IAEA Nuclear Security Series.
P1449_cover.indd 2 2011-05-05 08:39:50
DISPOSAL OF RADIOACTIVE WASTE
The following States are Members of the International Atomic Energy Agency:
The Agency’s Statute was approved on 23 October 1956 by the Conference on the Statute of the
IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1957. The
Headquarters of the Agency are situated in Vienna. Its principal objective is “to accelerate and enlarge the
contribution of atomic energy to peace, health and prosperity throughout the world’’.
AFGHANISTAN
ALBANIA
ALGERIA
ANGOLA
ARGENTINA
ARMENIA
AUSTRALIA
AUSTRIA
AZERBAIJAN
BAHRAIN
BANGLADESH
BELARUS
BELGIUM
BELIZE
BENIN
BOLIVIA
BOSNIA AND HERZEGOVINA
BOTSWANA
BRAZIL
BULGARIA
BURKINA FASO
BURUNDI
CAMBODIA
CAMEROON
CANADA
CENTRAL AFRICAN
REPUBLIC
CHAD
CHILE
CHINA
COLOMBIA
CONGO
COSTA RICA
CÔTE D’IVOIRE
CROATIA
CUBA
CYPRUS
CZECH REPUBLIC
DEMOCRATIC REPUBLIC
OF THE CONGO
DENMARK
DOMINICAN REPUBLIC
ECUADOR
EGYPT
EL SALVADOR
ERITREA
ESTONIA
ETHIOPIA
FINLAND
FRANCE
GABON
GEORGIA
GERMANY
GHANA
GREECE
GUATEMALA
HAITI
HOLY SEE
HONDURAS
HUNGARY
ICELAND
INDIA
INDONESIA
IRAN, ISLAMIC REPUBLIC OF
IRAQ
IRELAND
ISRAEL
ITALY
JAMAICA
JAPAN
JORDAN
KAZAKHSTAN
KENYA
KOREA, REPUBLIC OF
KUWAIT
KYRGYZSTAN
LATVIA
LEBANON
LESOTHO
LIBERIA
LIBYAN ARAB JAMAHIRIYA
LIECHTENSTEIN
LITHUANIA
LUXEMBOURG
MADAGASCAR
MALAWI
MALAYSIA
MALI
MALTA
MARSHALL ISLANDS
MAURITANIA
MAURITIUS
MEXICO
MONACO
MONGOLIA
MONTENEGRO
MOROCCO
MOZAMBIQUE
MYANMAR
NAMIBIA
NEPAL
NETHERLANDS
NEW ZEALAND
NICARAGUA
NIGER
NIGERIA
NORWAY
OMAN
PAKISTAN
PALAU
PANAMA
PARAGUAY
PERU
PHILIPPINES
POLAND
PORTUGAL
QATAR
REPUBLIC OF MOLDOVA
ROMANIA
RUSSIAN FEDERATION
SAUDI ARABIA
SENEGAL
SERBIA
SEYCHELLES
SIERRA LEONE
SINGAPORE
SLOVAKIA
SLOVENIA
SOUTH AFRICA
SPAIN
SRI LANKA
SUDAN
SWEDEN
SWITZERLAND
SYRIAN ARAB REPUBLIC
TAJIKISTAN
THAILAND
THE FORMER YUGOSLAV
REPUBLIC OF MACEDONIA
TUNISIA
TURKEY
UGANDA
UKRAINE
UNITED ARAB EMIRATES
UNITED KINGDOM OF
GREAT BRITAIN AND
NORTHERN IRELAND
UNITED REPUBLIC
OF TANZANIA
UNITED STATES OF AMERICA
URUGUAY
UZBEKISTAN
VENEZUELA
VIETNAM
YEMEN
ZAMBIA
ZIMBABWE
DISPOSAL OF RADIOACTIVE WASTE
SPECIFIC SAFETY REQUIREMENTS
This publication includes a CD-ROM containing the IAEA Safety Glossary:
2007 Edition (2007) and the Fundamental Safety Principles (2006),
each in Arabic, Chinese, English, French, Russian and Spanish versions.
The CD-ROM is also available for purchase separately.
See:http://www-pub.iaea.org/MTCD/publications/publications.asp
INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA, 2011
IAEA SAFETY STANDARDS SERIES No. SSR-5
IAEA Library Cataloguing in Publication Data
Disposal of radioactive waste. — Vienna : International Atomic Energy Agency,
2011.
p. ; 24 cm. — (IAEA safety standards series, ISSN 1020–525X ;
no. SSR-5)
STI/PUB/1449
ISBN 978–92–0–103010–8
Includes bibliographical references.
1. Radioactive waste disposal — Safety measures. — 2. Radioactive waste
disposal — Environmental aspects. — 3. Radioactive waste sites —
Management. 4. Safety standards. I. International Atomic Energy Agency.
II. Series.
IAEAL 10–00661
COPYRIGHT NOTICE
All IAEA scientific and technical publications are protected by the terms of
the Universal Copyright Convention as adopted in 1952 (Berne) and as revised in
1972 (Paris). The copyright has since been extended by the World Intellectual
Property Organization (Geneva) to include electronic and virtual intellectual
property. Permission to use whole or parts of texts contained in IAEA
publications in printed or electronic form must be obtained and is usually subject
to royalty agreements. Proposals for non-commercial reproductions and
translations are welcomed and considered on a case-by-case basis. Enquiries
should be addressed to the IAEA Publishing Section at:
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International Atomic Energy Agency
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PO Box 100
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fax: +43 1 2600 29302
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email: [email protected]http://www.iaea.org/books
© IAEA, 2011
Printed by the IAEA in Austria
April 2011
STI/PUB/1449
FOREWORD
by Yukiya Amano
Director General
The IAEA’s Statute authorizes the Agency to “establish or adopt…
standards of safety for protection of health and minimization of danger to life and
property” — standards that the IAEA must use in its own operations, and which
States can apply by means of their regulatory provisions for nuclear and radiation
safety. The IAEA does this in consultation with the competent organs of the
United Nations and with the specialized agencies concerned. A comprehensive
set of high quality standards under regular review is a key element of a stable and
sustainable global safety regime, as is the IAEA’s assistance in their application.
The IAEA commenced its safety standards programme in 1958. The
emphasis placed on quality, fitness for purpose and continuous improvement has
led to the widespread use of the IAEA standards throughout the world. The Safety
Standards Series now includes unified Fundamental Safety Principles, which
represent an international consensus on what must constitute a high level of
protection and safety. With the strong support of the Commission on Safety
Standards, the IAEA is working to promote the global acceptance and use of its
standards.
Standards are only effective if they are properly applied in practice. The
IAEA’s safety services encompass design, siting and engineering safety,
operational safety, radiation safety, safe transport of radioactive material and safe
management of radioactive waste, as well as governmental organization,
regulatory matters and safety culture in organizations. These safety services assist
Member States in the application of the standards and enable valuable experience
and insights to be shared.
Regulating safety is a national responsibility, and many States have decided
to adopt the IAEA’s standards for use in their national regulations. For parties to
the various international safety conventions, IAEA standards provide a
consistent, reliable means of ensuring the effective fulfilment of obligations
under the conventions. The standards are also applied by regulatory bodies and
operators around the world to enhance safety in nuclear power generation and in
nuclear applications in medicine, industry, agriculture and research.
Safety is not an end in itself but a prerequisite for the purpose of the
protection of people in all States and of the environment — now and in the future.
The risks associated with ionizing radiation must be assessed and controlled
without unduly limiting the contribution of nuclear energy to equitable and
sustainable development. Governments, regulatory bodies and operators
everywhere must ensure that nuclear material and radiation sources are used
beneficially, safely and ethically. The IAEA safety standards are designed to
facilitate this, and I encourage all Member States to make use of them.
DISCLAIMER
The IAEA safety standards reflect an international consensus on what
constitutes a high level of safety for protecting people and the environment from
harmful effects of ionizing radiation. The process of developing, reviewing and
establishing the IAEA standards involves the IAEA Secretariat and all Member
States, many of which are represented on the four IAEA safety standards
committees and the IAEA Commission on Safety Standards.
The IAEA standards, as a key element of the global safety regime, are kept
under regular review by the Secretariat, the safety standards committees and the
Commission on Safety Standards. The Secretariat gathers information on
experience in the application of the IAEA standards and information gained from
the follow-up of events for the purpose of ensuring that the standards continue to
meet users’ needs. The present publication reflects feedback and experience
accumulated until 2010 and it has been subject to the rigorous review process for
standards.
The accident at the Fukushima Daiichi nuclear power plant in Japan caused
by the disastrous earthquake and tsunami of 11 March 2011 and the consequences
of the emergency for people and the environment have to be fully investigated.
They are already under study in Japan, at the IAEA and elsewhere. Lessons to be
learned for nuclear safety and radiation protection and for emergency
preparedness and response will be reflected in IAEA safety standards as they are
revised and issued in the future.
THE IAEA SAFETY STANDARDS
BACKGROUND
Radioactivity is a natural phenomenon and natural sources of radiation
are features of the environment. Radiation and radioactive substances have
many beneficial applications, ranging from power generation to uses in
medicine, industry and agriculture. The radiation risks to workers and the
public and to the environment that may arise from these applications have to
be assessed and, if necessary, controlled.
Activities such as the medical uses of radiation, the operation of nuclear
installations, the production, transport and use of radioactive material, and the
management of radioactive waste must therefore be subject to standards of
safety.
Regulating safety is a national responsibility. However, radiation risks
may transcend national borders, and international cooperation serves to
promote and enhance safety globally by exchanging experience and by
improving capabilities to control hazards, to prevent accidents, to respond to
emergencies and to mitigate any harmful consequences.
States have an obligation of diligence and duty of care, and are expected
to fulfil their national and international undertakings and obligations.
International safety standards provide support for States in meeting their
obligations under general principles of international law, such as those relating
to environmental protection. International safety standards also promote and
assure confidence in safety and facilitate international commerce and trade.
A global nuclear safety regime is in place and is being continuously
improved. IAEA safety standards, which support the implementation of
binding international instruments and national safety infrastructures, are a
cornerstone of this global regime. The IAEA safety standards constitute
a useful tool for contracting parties to assess their performance under these
international conventions.
THE IAEA SAFETY STANDARDS
The status of the IAEA safety standards derives from the IAEA’s Statute,
which authorizes the IAEA to establish or adopt, in consultation and, where
appropriate, in collaboration with the competent organs of the United Nations
and with the specialized agencies concerned, standards of safety for protection
of health and minimization of danger to life and property, and to provide for
their application.
With a view to ensuring the protection of people and the environment
from harmful effects of ionizing radiation, the IAEA safety standards establish
fundamental safety principles, requirements and measures to control the
radiation exposure of people and the release of radioactive material to the
environment, to restrict the likelihood of events that might lead to a loss of
control over a nuclear reactor core, nuclear chain reaction, radioactive source
or any other source of radiation, and to mitigate the consequences of such
events if they were to occur. The standards apply to facilities and activities that
give rise to radiation risks, including nuclear installations, the use of radiation
and radioactive sources, the transport of radioactive material and the
management of radioactive waste.
Safety measures and security measures
1
have in common the aim of
protecting human life and health and the environment. Safety measures and
security measures must be designed and implemented in an integrated manner
so that security measures do not compromise safety and safety measures do not
compromise security.
The IAEA safety standards reflect an international consensus on what
constitutes a high level of safety for protecting people and the environment
from harmful effects of ionizing radiation. They are issued in the IAEA Safety
Standards Series, which has three categories (see Fig. 1).
Safety Fundamentals
Safety Fundamentals present the fundamental safety objective and
principles of protection and safety, and provide the basis for the safety
requirements.
Safety Requirements
An integrated and consistent set of Safety Requirements establishes the
requirements that must be met to ensure the protection of people and the
environment, both now and in the future. The requirements are governed by
the objective and principles of the Safety Fundamentals. If the requirements
are not met, measures must be taken to reach or restore the required level of
safety. The format and style of the requirements facilitate their use for the
establishment, in a harmonized manner, of a national regulatory framework.
Requirements, including numbered ‘overarching’ requirements, are expressed
1
See also publications issued in the IAEA Nuclear Security Series.
as ‘shall’ statements. Many requirements are not addressed to a specific party,
the implication being that the appropriate parties are responsible for fulfilling
them.
Safety Guides
Safety Guides provide recommendations and guidance on how to comply
with the safety requirements, indicating an international consensus that it is
necessary to take the measures recommended (or equivalent alternative
measures). The Safety Guides present international good practices, and
increasingly they reflect best practices, to help users striving to achieve high
levels of safety. The recommendations provided in Safety Guides are expressed
as ‘should’ statements.
APPLICATION OF THE IAEA SAFETY STANDARDS
The principal users of safety standards in IAEA Member States are
regulatory bodies and other relevant national authorities. The IAEA safety
Part 1. Governmental, Legal and
Regulatory Framework for Safety
Part 2. Leadership and Management
for Safety
Part 3. Radiation Protection and the
Safety of Radiation Sources
Part 4. Safety Assessment for
Facilities and Activities
Part 5. Predisposal Management
of Radioactive Waste
Part 6. Decommissioning and
Termination of Activities
Part 7. Emergency Preparedness
and Response
1. Site Evaluation for
Nuclear Installations
2. Safety of Nuclear Power Plants
2.1. Design and Construction
2.2. Commissioning and Operation
3. Safety of Research Reactors
4. Safety of Nuclear Fuel
Cycle Facilities
5. Safety of Radioactive Waste
Disposal Facilities
6. Safe Transport of
Radioactive Material
General Safety Requirements Specific Safety Requirements
Safety Fundamentals
Fundamental Safety Principles
Collection of Safety Guides
FIG. 1. The long term structure of the IAEA Safety Standards Series.
standards are also used by co-sponsoring organizations and by many
organizations that design, construct and operate nuclear facilities, as well as
organizations involved in the use of radiation and radioactive sources.
The IAEA safety standards are applicable, as relevant, throughout the
entire lifetime of all facilities and activities — existing and new — utilized for
peaceful purposes and to protective actions to reduce existing radiation risks.
They can be used by States as a reference for their national regulations in
respect of facilities and activities.
The IAEA’s Statute makes the safety standards binding on the IAEA in
relation to its own operations and also on States in relation to IAEA assisted
operations.
The IAEA safety standards also form the basis for the IAEA’s safety
review services, and they are used by the IAEA in support of competence
building, including the development of educational curricula and training
courses.
International conventions contain requirements similar to those in the
IAEA safety standards and make them binding on contracting parties.
The IAEA safety standards, supplemented by international conventions,
industry standards and detailed national requirements, establish a consistent
basis for protecting people and the environment. There will also be some
special aspects of safety that need to be assessed at the national level. For
example, many of the IAEA safety standards, in particular those addressing
aspects of safety in planning or design, are intended to apply primarily to new
facilities and activities. The requirements established in the IAEA safety
standards might not be fully met at some existing facilities that were built to
earlier standards. The way in which IAEA safety standards are to be applied
to such facilities is a decision for individual States.
The scientific considerations underlying the IAEA safety standards
provide an objective basis for decisions concerning safety; however, decision
makers must also make informed judgements and must determine how best to
balance the benefits of an action or an activity against the associated radiation
risks and any other detrimental impacts to which it gives rise.
DEVELOPMENT PROCESS FOR THE IAEA SAFETY STANDARDS
The preparation and review of the safety standards involves the IAEA
Secretariat and four safety standards committees, for nuclear safety (NUSSC),
radiation safety (RASSC), the safety of radioactive waste (WASSC) and the
safe transport of radioactive material (TRANSSC), and a Commission on
Safety Standards (CSS) which oversees the IAEA safety standards programme
(see Fig. 2).
All IAEA Member States may nominate experts for the safety standards
committees and may provide comments on draft standards. The membership of
the Commission on Safety Standards is appointed by the Director General and
includes senior governmental officials having responsibility for establishing
national standards.
A management system has been established for the processes of planning,
developing, reviewing, revising and establishing the IAEA safety standards.
It articulates the mandate of the IAEA, the vision for the future application of
the safety standards, policies and strategies, and corresponding functions and
responsibilities.
INTERACTION WITH OTHER INTERNATIONAL ORGANIZATIONS
The findings of the United Nations Scientific Committee on the Effects of
Atomic Radiation (UNSCEAR) and the recommendations of international
Secretariat and
consultants:
drafting of new or revision
of existing safety standard
Draft
Endorsement
by the CSS
Final draft
Review by
safety standards
committee(s)
Member States
Comments
Draft
Outline and work plan
prepared by the Secretariat;
review by the safety standards
committees and the CSS
FIG. 2. The process for developing a new safety standard or revising an existing standard.
expert bodies, notably the International Commission on Radiological
Protection (ICRP), are taken into account in developing the IAEA safety
standards. Some safety standards are developed in cooperation with other
bodies in the United Nations system or other specialized agencies, including
the Food and Agriculture Organization of the United Nations, the United
Nations Environment Programme, the International Labour Organization, the
OECD Nuclear Energy Agency, the Pan American Health Organization and
the World Health Organization.
INTERPRETATION OF THE TEXT
Safety related terms are to be understood as defined in the IAEA Safety
Glossary (seehttp://www-ns.iaea.org/standards/safety-glossary.htm). Otherwise,
words are used with the spellings and meanings assigned to them in the latest
edition of The Concise Oxford Dictionary. For Safety Guides, the English
version of the text is the authoritative version.
The background and context of each standard in the IAEA Safety
Standards Series and its objective, scope and structure are explained in
Section 1, Introduction, of each publication.
Material for which there is no appropriate place in the body text
(e.g. material that is subsidiary to or separate from the body text, is included in
support of statements in the body text, or describes methods of calculation,
procedures or limits and conditions) may be presented in appendices or
annexes.
An appendix, if included, is considered to form an integral part of the
safety standard. Material in an appendix has the same status as the body text,
and the IAEA assumes authorship of it. Annexes and footnotes to the main
text, if included, are used to provide practical examples or additional
information or explanation. Annexes and footnotes are not integral parts of the
main text. Annex material published by the IAEA is not necessarily issued
under its authorship; material under other authorship may be presented in
annexes to the safety standards. Extraneous material presented in annexes is
excerpted and adapted as necessary to be generally useful.
CONTENTS
1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Background (1.1–1.26). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Objective (1.27–1.28). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
Scope (1.29–1.32) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
Structure (1.33) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
2. PROTECTION OF PEOPLE AND THE ENVIRONMENT . . . . . . . 10
Application of the fundamental safety principles (2.1–2.6). . . . . . . . . 10
Radiation protection in the operational period (2.7–2.14) . . . . . . . . . . 11
Radiation protection in the post-closure period (2.15–2.19) . . . . . . . . 13
Environmental and non-radiological concerns (2.20–2.24). . . . . . . . . 15
3. SAFETY REQUIREMENTS FOR PLANNING
FOR THE DISPOSAL OF RADIOACTIVE WASTE (3.1–3.5). . . . . 16
Governmental, legal and regulatory framework . . . . . . . . . . . . . . . . . 17
Requirement 1: Government responsibilities (3.6–3.7) . . . . . . . . . . 17
Requirement 2: Responsibilities of the
regulatory body (3.8–3.11). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
Requirement 3: Responsibilities of the operator (3.12–3.16) . . . . . 19
Safety approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
Requirement 4: Importance of safety in the process of
development and operation of a disposal facility (3.17–3.20). . . 20
Requirement 5: Passive means for the safety of the disposal
facility (3.21–3.25). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
Requirement 6: Understanding of a disposal facility and
confidence in safety (3.26–3.31) . . . . . . . . . . . . . . . . . . . . . . . . . 22
Design concepts for safety (3.32–3.34) . . . . . . . . . . . . . . . . . . . . . . . . 24
Requirement 7: Multiple safety functions (3.35–3.38) . . . . . . . . . . 24
Requirement 8: Containment of radioactive waste (3.39–3.42). . . . 26
Requirement 9: Isolation of radioactive waste (3.43–3.47) . . . . . . . 27
Requirement 10: Surveillance and control of passive
safety features (3.48) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
4. REQUIREMENTS FOR THE DEVELOPMENT, OPERATION
AND CLOSURE OF A DISPOSAL FACILITY (4.1) . . . . . . . . . . . . 29
Framework for disposal of radioactive waste . . . . . . . . . . . . . . . . . . . 29
Requirement 11: Step by step development and evaluation of
disposal facilities (4.2–4.5) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
The safety case and safety assessment (4.6–4.11) . . . . . . . . . . . . . . . . 30
Requirement 12: Preparation, approval and use of the safety
case and safety assessment for a disposal facility
(4.12–4.14) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
Requirement 13: Scope of the safety case and safety
assessment (4.15–4.22) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
Requirement 14: Documentation of the safety case and safety
assessment (4.23–4.25) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
Steps in the development, operation and closure of
a disposal facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
Requirement 15: Site characterization for a disposal facility
(4.26–4.29) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
Requirement 16: Design of a disposal facility (4.30–4.32) . . . . . . . 36
Requirement 17: Construction of a disposal facility
(4.33–4.34) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37
Requirement 18: Operation of a disposal facility (4.35–4.37) . . . . . 38
Requirement 19: Closure of a disposal facility (4.38–4.41). . . . . . . 38
5. ASSURANCE OF SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39
Requirement 20: Waste acceptance in a disposal facility
(5.1–5.3) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39
Requirement 21: Monitoring programmes at a disposal facility
(5.4–5.5) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40
Requirement 22: The period after closure and institutional
controls (5.6–5.14) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41
Requirement 23: Consideration of the State system of
accounting for, and control of, nuclear material (5.15–5.19) . . . 43
Requirement 24: Requirements in respect of nuclear security
measures (5.20–5.21) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44
Requirement 25: Management systems (5.22–5.26) . . . . . . . . . . . . 45
6. EXISTING DISPOSAL FACILITIES (6.1) . . . . . . . . . . . . . . . . . . . . 46
Requirement 26: Existing disposal facilities (6.2–6.3) . . . . . . . . . . 46
APPENDIX: ASSURANCE OF COMPLIANCE WITH THE
SAFETY OBJECTIVE AND CRITERIA . . . . . . . . . . . . . . . 49
REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53
ANNEX: RADIOACTIVE WASTE CLASSIFICATION . . . . . . . . . . 55
CONTRIBUTORS TO DRAFTING AND REVIEW . . . . . . . . . . . . . . . . . 57
BODIES FOR THE ENDORSEMENT OF
IAEA SAFETY STANDARDS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59
1
1. INTRODUCTION
BACKGROUND
General
1.1. Radioactive waste arises from the generation of electricity in nuclear power
plants, from nuclear fuel cycle operations and from activities in which radioactive
material is used. It also arises from activities and processes in which radioactive
material of natural origin becomes concentrated in waste material and safety
needs to be considered in its management. Radioactive waste can be generated by
a wide range of activities varying from activities in hospitals to nuclear power
plants to mines and mineral processing facilities.
1.2. The properties of radioactive waste are likewise varied, not only in terms of
radioactive content and activity concentration but also in terms of physical and
chemical properties. Its rate of generation is also varied. A common characteristic
of all radioactive waste is its potential to present a hazard to people and to the
environment, and it must, therefore, be managed so as to reduce any associated
risks to acceptable levels. The potential hazard can range from large to trivial:
a variation reflected in the management and disposal options necessary for
various types of waste.
1.3. The safety principles to be applied in all activities for radioactive waste
management are set out in the IAEA Safety Fundamentals [1]. These principles
also form the ethical and conceptual basis for the Joint Convention on the Safety
of Spent Fuel Management and on the Safety of Radioactive Waste Management
[2]. The requirements for radiation protection are set out in the International
Basic Safety Standards for Protection against Ionizing Radiation and for the
Safety of Radiation Sources (the International Basic Safety Standards) [3]. Many
of the safety requirements and concepts of protection adopted in the standards
and in the Joint Convention [2] derive from the recommendations of the
International Commission on Radiological Protection (ICRP) [4–7].
1.4. This Safety Requirements publication establishes safety requirements
relating to the disposal of radioactive waste of all types. It sets out the safety
objective and criteria for the protection of people and the environment against
radiation risks arising from disposal facilities for radioactive waste in operation
and after closure. In order to meet the criteria, measures may need to be taken in
site selection and evaluation and in the design, construction, operation and
2
closure of the disposal facility. The requirements are essential from a safety
perspective and failure to meet any of the requirements would require action to be
taken.
1.5. This Safety Requirements publication does not reiterate all the safety
requirements in respect of the governmental, legal and regulatory framework,
radiation protection and emergency planning that are established in other Safety
Requirements publications. It is based on the premise that, in general,
arrangements have to be in place to ensure that these related requirements are
met. This Safety Requirements publication does set out some requirements that
are closely related to these other thematic areas and which are of particular
importance to the safety of radioactive waste disposal facilities. Guidance on the
fulfilment of the safety requirements set out in this Safety Requirements
publication is provided in several Safety Guides specific to different types of
radioactive waste disposal facility.
1.6. The preferred strategy for the management of all radioactive waste is to
contain it (i.e. to confine the radionuclides to within the waste matrix, the
packaging and the disposal facility) and to isolate it from the accessible
biosphere. This strategy does not preclude the discharge (i.e. controlled release)
of effluents, arising from waste management activities, that contain residual
amounts of radionuclides, or the clearance of materials that meet the relevant
criteria. International safety standards have been established covering both of
these circumstances [8, 9].
1.7. Radioactive waste may arise initially in various gaseous, liquid and solid
forms. In waste management activities, the waste is generally processed to
produce stable and solid forms, and reduced in volume and immobilized, as far as
practicable, to facilitate their storage, transport and disposal. This Safety
Requirements publication is concerned with the stage of disposal of solid or
solidified materials, which is the last step in the process of radioactive waste
management.
3
Concepts relating to disposal (and storage) of radioactive waste
1.8. The term ‘disposal’ refers to the emplacement of radioactive waste into a
facility or a location with no intention of retrieving the waste
1
. Disposal options
are designed to contain the waste by means of passive engineered and natural
features and to isolate it from the accessible biosphere to the extent necessitated
by the associated hazard. The term disposal implies that retrieval is not intended;
it does not mean that retrieval is not possible.
1.9. By contrast, the term ‘storage’ refers to the retention of radioactive waste in
a facility or a location with the intention of retrieving the waste. Both options,
disposal and storage, are designed to contain waste and to isolate it from the
accessible biosphere to the extent necessary. The important difference is that
storage is a temporary measure following which some future action is planned.
This may include further conditioning or packaging of the waste and, ultimately,
its disposal. Guidance on the safe storage of radioactive waste is provided in
Ref. [11].
1.10. A number of design options for disposal facilities have been developed and
various types of disposal facility have been constructed in many States and are in
operation. These design options have different degrees of containment and
isolation capability appropriate to the radioactive waste that they will receive.
The specific aims of disposal are:
(a) To contain the waste;
(b) To isolate the waste from the accessible biosphere and to reduce
substantially the likelihood of, and all possible consequences of,
inadvertent human intrusion
2
into the waste;
(c) To inhibit, reduce and delay the migration of radionuclides at any time from
the waste to the accessible biosphere;
(d) To ensure that the amounts of radionuclides reaching the accessible
biosphere due to any migration from the disposal facility are such that
possible radiological consequences are acceptably low at all times.
1
Terminology used in this publication is defined and explained in the IAEA Safety
Glossary [10] (seehttp://www-ns.iaea.org/standards/safety-glossary.htm).
2
‘Human intrusion’ refers to human actions that affect the integrity of a disposal facility
and which could potentially give rise to radiological consequences. Only those human actions
that result in direct disturbance of the disposal facility (i.e. the waste itself, the contaminated
near field or the engineered barrier materials) are considered.
4
1.11. The balance between the importance of each of the above mentioned aims
and the extent to which and the way in which they are accomplished will vary,
depending on the characteristics of the waste and the type of disposal facility.
1.12. Disposal facilities are not expected to provide complete containment and
isolation of waste over all time; this is neither practicable nor necessitated by the
hazard associated with waste, which declines with time.
Types of disposal facility for radioactive waste
1.13. As indicated in para. 1.10, a number of design options for disposal facilities
have been developed and various types of disposal facility have been constructed
and are in operation around the world.
1.14. Within any State or region, a number of disposal facilities of different
designs may be required in order to accommodate radioactive waste of various
types. The classification of radioactive waste is discussed in an IAEA Safety
Guide [12] and the different classes of radioactive waste are presented in the
Annex. The following disposal options have been adopted in one or more States,
corresponding to recognized classes of radioactive waste:
(a) Specific landfill disposal: Disposal in a facility similar to a conventional
landfill facility for industrial refuse but which may incorporate measures to
cover the waste. Such a facility may be designated as a disposal facility for
very low level radioactive waste (VLLW) with low concentrations or
quantities of radioactive content [12]. Typical waste disposed of in a facility
of this type may include soil and rubble arising from decommissioning
activities.
(b) Near surface disposal: Disposal in a facility consisting of engineered
trenches or vaults constructed on the ground surface or up to a few tens of
metres below ground level. Such a facility may be designated as a disposal
facility for low level radioactive waste (LLW) [12].
(c) Disposal of intermediate level waste: Depending on its characteristics,
intermediate level radioactive waste (ILW) can be disposed of in different
types of facility [12]. Disposal could be by emplacement in a facility
constructed in caverns, vaults or silos at least a few tens of metres below
ground level and up to a few hundred metres below ground level. It could
include purpose built facilities and facilities developed in or from existing
mines. It could also include facilities developed by drift mining into
mountainsides or hillsides, in which case the overlying cover could be more
than 100 m deep.
5
(d) Geological disposal: Disposal in a facility constructed in tunnels, vaults or
silos in a particular geological formation (e.g. in terms of its long term
stability and its hydrogeological properties) at least a few hundred metres
below ground level. Such a facility could be designed to receive high level
radioactive waste (HLW) [12], including spent fuel if it is to be treated as
waste. However, with appropriate design, a geological disposal facility
could receive all types of radioactive waste.
(e) Borehole disposal: Disposal in a facility consisting of an array of boreholes,
or a single borehole, which may be between a few tens of metres up to a few
hundreds of metres deep. Such a borehole disposal facility is designed for
the disposal of only relatively small volumes of waste, in particular disused
sealed radioactive sources. A design option for very deep boreholes, several
kilometres deep, has been examined for the disposal of solid high level
waste and spent fuel, but this option has not been adopted for a disposal
facility by any State.
(f) Disposal of mining and mineral processing waste: Disposal usually on or
near the ground surface, but the manner and the large volumes in which the
waste arises, its physicochemical form and its content of long lived
radionuclides of natural origin distinguish it from other radioactive waste.
The waste is generally stabilized in situ and covered with various layers of
rock and soil.
1.15. This Safety Requirements publication applies to all of the above mentioned
types of disposal and disposal facilities. Comprehensive guidance on meeting the
requirements established in this Safety Requirements publication is given in
IAEA Safety Guides, in each of which a particular type of disposal, as described
above, is considered.
1.16. In accordance with the graded approach, as required in the International
Basic Safety Standards and other standards [3, 13, 14], the ability of the chosen
disposal system to provide containment of the waste and to isolate it from people
and the environment will be commensurate with the hazard potential of the waste.
The requirements set out in this Safety Requirements publication apply to all
types of disposal facility. However, the extent of provisions necessary to meet the
requirements will vary in accordance with the graded approach. This is reflected
in the Safety Guides for the different types of facility mentioned in para. 1.14.
Disposal facility life cycle
1.17. The development (i.e. site selection and evaluation, and facility design and
construction) of most types of disposal facility is likely to take place over
6
extended periods of time. The period over which disposal facilities will be
operated prior to closure will, in most cases, also extend over decades. Different
activities will be conducted in this period of development, such as site selection
and evaluation, and facility design and construction, with decisions being made to
proceed to the next set of activities or the next step in the development of the
facility.
1.18. Such a step by step approach enables: the ordered accumulation and
assessment of the necessary scientific and technical data; the evaluation of
possible sites; the development of disposal concepts; iterative studies for design
development and safety assessment with progressively improving data; technical
and regulatory reviews; public consultation and political decisions. However, the
level of study and the process will depend on the facility and on national
practices.
1.19. The step by step approach, together with the consideration of a range of
options for the design and operational management of a disposal facility, is
expected to provide flexibility for responding to new technical information and
advances in waste management and material technologies. It also enables social,
economic and political aspects of the disposal facility to be addressed, to ensure
that all reasonable measures have been taken to further prevent, inhibit or delay
releases to the environment.
1.20. This approach may include options for reversing a given step or even, for
most types of facility, for retrieving waste after its emplacement, if this were
considered to be appropriate.
1.21. The developers of disposal facilities may define a number of steps relating
to their own programme needs. In this Safety Requirements publication, however,
the step by step approach refers to the steps that are imposed by the regulatory
body and by political decision making processes.
1.22. It is convenient to identify three periods associated with the development,
operation and closure of a disposal facility: (i) the pre-operational period,
(ii) the operational period and (iii) the post-closure period. Various activities will
take place in these periods and some may be undertaken to varying degrees
throughout part or all of the lifetime of the facility:
(i) The pre-operational period includes concept definition, site evaluation
(selection, verification and confirmation), safety assessment and design
studies. It also includes the development of those aspects of the safety case
7
for safety in operation and after closure that are required in order to set the
conditions of authorization, obtain the authorization and proceed with the
construction of the disposal facility and the initial operational activities.
The monitoring and testing programmes that are needed to inform
operational management decisions are put in place.
(ii) The operational period begins when waste is first received at the facility.
From this time, radiation exposures may occur as a result of waste
management activities, and these are subject to control in accordance with
the requirements for protection and safety. Monitoring, surveillance and
testing programmes continue to inform operational management decisions
and to provide the basis for decisions concerning the closure of the facility
or parts of it. Safety assessments for the period of operation and the period
after closure and the safety case are updated as necessary to reflect actual
experience and increasing knowledge. In the operational period,
construction activities may take place at the same time as waste
emplacement in, and closure of, other parts of the facility. This period may
include activities for waste retrieval, if considered necessary, prior to
closure, activities following the completion of waste emplacement and the
final closure and sealing of the facility.
(iii) The post-closure period begins at the time when all the engineered
containment and isolation features have been put in place, operational
buildings and supporting services have been decommissioned and the
facility is in its final configuration. After its closure, the safety of the
disposal facility is provided for by means of passive features inherent in the
characteristics of the site and the facility and the characteristics of the waste
packages, together with certain institutional controls, particularly for near
surface facilities. Such institutional controls are put in place to prevent
intrusion into facilities and to confirm that the disposal system is
performing as expected by means of monitoring and surveillance.
Monitoring may also be carried out to provide public assurance. The licence
will be terminated after the period of active institutional control, when all
the necessary technical, legal and financial requirements have been
fulfilled.
1.23. This Safety Requirements publication is concerned with providing for the
protection of people and the environment against the hazards associated with
waste management activities relating to waste disposal, including hazards that
could arise in the operational period and following closure. Assurance of this
protection will be provided by the application of legal and regulatory
requirements in the pre-operational and operational periods, and in some cases in
the post-closure period.
8
1.24. The disposal system (i.e. the disposal facility and the environment in which
it is sited) is developed in a series of steps in which the scientific understanding of
the disposal system and of the design of the disposal facility is progressively
advanced. Safety assessment is an important tool for guiding site selection and
evaluation and for assisting with the design of the facility. It is also used for
evaluating the prevailing level of understanding of the disposal system and for
assessing the associated uncertainties through the various steps in the
development of the facility. The extent and complexity of such an assessment will
vary with the type of facility and will be related to the hazard potential of the
waste.
1.25. Moreover, the development of disposal facilities that incorporate provisions
in design or operation to facilitate reversibility, including retrievability, is
considered in several national programmes for waste management. In some
States, post-closure retrievability is a legal requirement and constitutes a
boundary condition on the options available, which must always satisfy the safety
requirements for disposal. No relaxation of safety standards or requirements
could be allowed on the grounds that waste retrieval may be possible or may be
facilitated by a particular provision. It would have to be ensured that any such
provision would not have an unacceptable adverse effect on safety or on the
performance of the disposal system. This subject is not extensively dealt with in
this Safety Requirements publication.
1.26. The safety case (i.e. the collection of arguments and evidence to
demonstrate the safety of a facility) for a disposal facility will be developed
together with the development of the facility. This approach provides a basis for
decisions relating to the development, operation and closure of the facility. It also
allows the identification of areas of uncertainty on which attention needs to be
focused to improve further the understanding of those aspects influencing the
safety of the disposal system.
OBJECTIVE
1.27. The objective of this Safety Requirements publication is to set out the safety
objective and criteria for the disposal of all types of radioactive waste and to
establish, on the basis of the principles established in Ref. [1], the requirements
that must be satisfied in the disposal of radioactive waste.
9
1.28. This Safety Requirements publication is intended for use by all persons
responsible for, and concerned with, radioactive waste management and making
decisions in relation to the development, operation and closure of disposal
facilities, especially those persons concerned with the related regulatory aspects.
Safety Guides provide comprehensive guidance on, and international best
practices for, meeting the requirements in respect of different types of disposal
facility.
SCOPE
1.29. This Safety Requirements publication applies to the disposal of radioactive
waste of all types by means of emplacement in designed disposal facilities,
subject to the necessary limitations and controls being placed on the disposal of
the waste and on the development, operation and closure of facilities. The
classification of radioactive waste is discussed in Ref. [12].
1.30. This Safety Requirements publication establishes requirements to provide
assurance of the radiation safety of the disposal of radioactive waste, in the
operation of a disposal facility and especially after its closure. The fundamental
safety objective is to protect people and the environment from harmful effects of
ionizing radiation. This is achieved by setting requirements on the site selection
and evaluation and design of a disposal facility, and on its construction, operation
and closure, including organizational and regulatory requirements.
1.31. Meeting these requirements forms a part of the wider process involved in
selecting and evaluating a site and developing a disposal facility. Broader
planning, financial, economic and social issues, as well as issues of conventional
safety and environmental impacts, will also be considered in this wider process.
This Safety Requirements publication does not address these broader issues, nor
does it address the transport of waste to the site or environmental impacts other
than radiological consequences.
1.32. Experience to date in selecting sites for disposal facilities has shown that
acceptance of a disposal facility by a broad range of interested parties depends on
a number of factors. The process of involving interested parties in decision
making processes for disposal facilities is increasingly seen to be of great
importance. The detailed consideration of such processes is, however, beyond the
scope of this Safety Requirements publication.
10
STRUCTURE
1.33. The background to, the concepts of, and the safety objective for disposal are
set out in Sections 1 and 2. The safety requirements for disposal facilities are set
out in Sections 3–6. These requirements comprise 26 numbered ‘shall’ statements
in bold type.
2. PROTECTION OF PEOPLE AND THE ENVIRONMENT
APPLICATION OF THE FUNDAMENTAL SAFETY PRINCIPLES
2.1. The IAEA Safety Fundamentals publication Fundamental Safety Principles
[1] sets out the fundamental safety objective and safety principles that apply for
all facilities and activities in radioactive waste management, including the
disposal of radioactive waste. As stated in Ref. [1], the fundamental safety
objective is to protect people and the environment from harmful effects of
ionizing radiation.
2.2. The strategy adopted at present to achieve this fundamental safety objective
in respect of the disposal of radioactive waste is to contain the waste and to
isolate it from the accessible biosphere, to the extent that this is necessary. The
biosphere is that part of the environment that is normally inhabited by living
organisms, and in this Safety Requirements publication the ‘accessible biosphere’
is taken generally to include those elements of the environment, including
groundwater, surface water and marine resources, that are used by people or
accessible to people. The accessible biosphere is, therefore, that part of the
environment that the objective, criteria and requirements set out in this Safety
Requirements publication are established to protect.
2.3. By applying the strategy of containment and isolation of waste, it is implicit
that if waste were to be disturbed after its disposal in a facility, then radiation
doses might be incurred.
2.4. According to Ref. [1], disposal facilities are to be developed in such a way
that people and the environment are protected both now and in the future
(Ref. [1], Principle 7). In this regard, the prime consideration is the radiological
hazard presented by radioactive waste. The ICRP developed the System of
11
Radiological Protection that applies to all facilities and activities, and this system
was adopted in the International Basic Safety Standards [3].
2.5. The ICRP has elaborated the application of the System of Radiological
Protection to the disposal of solid radioactive waste in its Publications 77 and 81
[5, 6], which it reconfirmed in Publication 103 [7]. This provides a starting point
for the safety considerations discussed here in relation to disposal facilities.
Environmental concerns and other non-radiological concerns are considered at
the end of Section 2.
2.6. The safety objective and criteria set out in this section apply regardless of
national boundaries. Transboundary issues are dealt with in the framework of
existing conventions, treaties and bilateral agreements. Particular specific
obligations apply to Contracting Parties to the Joint Convention on the Safety of
Spent Fuel Management and on the Safety of Radioactive Waste Management [2].
RADIATION PROTECTION IN THE OPERATIONAL PERIOD
2.7. The radiation safety requirements and the related safety criteria for the
operational period of a disposal facility are the same as those for any nuclear
facility or activity involving radioactive material and are established in the
International Basic Safety Standards [3]. Disposal facilities receiving waste from
nuclear fuel cycle facilities will generally be licensed nuclear facilities and have
to operate under the terms of a facility licence. Disposal facilities for small
quantities of waste (e.g. borehole facilities) may not be regarded as nuclear
facilities in some States but have to be subject to an appropriate regulatory
process and have to be licensed accordingly.
2.8. In radiation safety terms, the disposal facility is considered to be a source of
radiation that is under regulatory control in a planned exposure situation. In the
operational period, any radioactive release can be verified, exposures can be
controlled and actions can be taken if necessary. The engineering means and
practical means of achieving safety are well known, although their use in a
disposal facility involves specific considerations. The primary goal is to ensure
that radiation doses are as low as reasonably achievable and within the applicable
system of dose limitation.
2.9. The optimization of protection (that is, the process of determining measures
for protection and safety to make exposures, and the probability and magnitude of
potential exposures, “as low as reasonably achievable, economic and social
12
factors being taken into account”) is considered in the design of the disposal
facility and in the planning of all operations [3].
2.10. Relevant considerations in the optimization of measures for protection and
safety include: the separation of mining and construction activities from waste
emplacement activities; the use of remote handling equipment and shielded
equipment for waste emplacement, where necessary; the control of the working
environment so as to reduce the potential for accidents and their potential
consequences; and the minimization of the need for maintenance in supervised
areas and controlled areas. Contamination is required to be controlled and
prevented to the extent possible [3].
2.11. No releases of radionuclides, or only very minor releases (such as small
amounts of gaseous radionuclides), may be expected during the normal operation
of a radioactive waste disposal facility and hence there will not be any significant
doses to members of the public. Even in the event of an accident involving the
breach of a waste package on the site of a disposal facility, releases are unlikely to
have any radiological consequences outside the facility.
2.12. The absence of radiological consequences of any significance outside the
facility would be confirmed by means of safety assessment (see the requirements
concerning the safety case and safety assessment, Requirements 12–14).
Relevant considerations include the waste form (i.e. the packaging and the
radionuclide content of the waste), the control of contamination on waste
packages and equipment, and the monitoring and control of drainage water from
the disposal facility, where applicable, and of the ventilation exhaust air from
underground disposal facilities.
2.13. For a disposal facility, as for any other operational nuclear facility or
facility where radioactive material is handled, used, stored or processed, an
operational radiation protection programme, commensurate with the radiological
hazards, is required to be put in place to ensure that doses to workers during
normal operations are controlled and that the requirements for the limitation of
radiation doses are met (see Ref. [3], paras 2.24–2.26, and Ref. [15]). In addition,
emergency plans are required to be put in place for dealing with accidents and
other incidents, and for ensuring that any consequent radiation doses are
controlled to the extent possible, with due regard for the relevant emergency
action levels [16].
2.14. The doses and risks associated with the transport of radioactive waste
through public areas to a disposal facility are required to be managed in the same
13
way as the doses and risks associated with the transport of other radioactive
material. The transport of radioactive waste is subject to the requirements of the
IAEA’s Regulations for the Safe Transport of Radioactive Material [17].
RADIATION PROTECTION IN THE POST-CLOSURE PERIOD
2.15. The safety objective and criteria for the protection of people and the
environment after closure of a disposal facility are as follows:
Safety objective
The safety objective is to site, design, construct, operate and close a disposal
facility so that protection after its closure is optimized, social and economic
factors being taken into account. A reasonable assurance also has to be provided
that doses and risks to members of the public in the long term will not exceed the
dose constraints or risk constraints that were used as design criteria.
Criteria
(a) The dose limit for members of the public for doses from all planned
exposure situations is an effective dose of 1 mSv in a year [3]. This and its
risk equivalent are considered criteria that are not to be exceeded in the
future.
(b) To comply with this dose limit, a disposal facility (considered as a single
source) is so designed that the calculated dose or risk to the representative
person who might be exposed in the future as a result of possible natural
processes
3
affecting the disposal facility does not exceed a dose constraint
of 0.3 mSv in a year or a risk constraint of the order of 10
–5
per year
4
.
(c) In relation to the effects of inadvertent human intrusion after closure, if
such intrusion is expected to lead to an annual dose of less than 1 mSv to
those living around the site, then efforts to reduce the probability of
intrusion or to limit its consequences are not warranted.
3
Natural processes include the range of conditions anticipated over the lifetime of the
facility and events that could occur with a lesser likelihood. However, extremely low
probability events would be outside the scope of consideration.
4
Risk due to the disposal facility in this context is to be understood as the probability of
fatal cancer or serious hereditary effects.
14
(d) If human intrusion were expected to lead to a possible annual dose of more
than 20 mSv (see Ref. [7], Table 8) to those living around the site, then
alternative options for waste disposal are to be considered, for example,
disposal of the waste below the surface, or separation of the radionuclide
content giving rise to the higher dose.
(e) If annual doses in the range 1–20 mSv (see Ref. [7], Table 8) are indicated,
then reasonable efforts are warranted at the stage of development of the
facility to reduce the probability of intrusion or to limit its consequences by
means of optimization of the facility’s design.
(f) Similar considerations apply where the relevant thresholds for deterministic
effects in organs may be exceeded.
2.16. It is recognized that radiation doses to people in the future can only be
estimated and that uncertainties associated with these estimates will increase for
periods farther into the future. Caution needs to be exercised in applying criteria
for periods far into the future. Beyond such timescales, the uncertainties
associated with dose estimates become so large that the criteria might no longer
serve as a reasonable basis for decision making.
2.17. The primary goal of the disposal of radioactive waste is the protection of
people and the environment in the long term, after the disposal facility has been
closed. In this period, migration of radionuclides to the accessible biosphere,
dispersion of radionuclides into the accessible biosphere and the consequent
exposure of people may occur. This is a consequence of the slow degradation of
engineered components and the slow transport of radionuclides from the facility
by natural processes. Discrete events may lead to an earlier or greater release.
Such events could be of either natural or human origin.
2.18. Optimization under constraints is the central approach adopted to ensure the
safety of a waste disposal facility [6]. In this context, the optimization of
protection is a judgemental process, social and economic factors being taken into
account. The optimization is conducted in a structured but essentially qualitative
manner, supported by quantitative analysis.
2.19. Different methods may be used to assess the impacts of the disposal of
radioactive waste after closure of the disposal facility and to demonstrate
compliance with national regulations expressed as constraints in terms of levels
15
of dose and/or risk. This matter is addressed in the Safety Guide on the safety
case and safety assessment for disposal
5
.
ENVIRONMENTAL AND NON-RADIOLOGICAL CONCERNS
2.20. The assessment of conventional environmental impacts such as may occur
in the construction and operational periods of a disposal facility, for example,
impacts relating to traffic, noise, visual amenity, disturbance of natural habitats,
restrictions on land use and social and economic factors, is outside the scope of
this Safety Requirements publication. This Safety Requirements publication
covers the protection of the environment against radiological hazards associated
with the radioactive material in the disposal facility. The non-radiological toxic
hazard also has to be assessed where this is significant, as discussed in the
following paragraphs.
2.21. For the purposes of the current recommendations of the ICRP [4] and the
requirements of the International Basic Safety Standards [3], it is assumed that,
subject to the appropriate definition of exposed groups, the protection of people
against the radiological hazards associated with a disposal facility will also apply
the principle of protecting the environment [4–7]. The issue of the protection of
the environment from harmful effects of ionizing radiation and the development
of standards for this purpose are under discussion internationally [7].
2.22. Estimates of possible doses and/or risks due to the future migration of
radionuclides from a disposal facility are indicators of the protection of people.
On the basis of the assumption mentioned in para. 2.21, calculations to estimate
doses in which account is taken of a range of possible environmental transfer
pathways could already be considered to be indicators of environmental
protection.
2.23. Additional indicators and comparisons, such as estimates of concentrations
and fluxes of contaminants and their comparison with concentrations and fluxes
of radionuclides of natural origin within the geosphere or biosphere, may also
prove valuable in indicating a level of overall environmental protection that is
5
A Safety Guide on the Safety Case and Safety Assessment for Disposal of Radioactive
Waste is in preparation.
16
independent of assumptions about the habits
6
of people. Other factors to be
considered may include the need for protection of groundwater resources and the
ecological sensitivity of the environment into which contaminants might be
released.
2.24. The impact of non-radioactive material present in a disposal facility has to
be assessed in accordance with national or other specific regulations and this may
be significant in some cases, for example, for some mining wastes and mixtures
of radioactive and toxic wastes. If non-radioactive material may affect the release
and migration of radioactive contaminants from the radioactive waste, then such
interactions have to be considered in the safety assessment.
3. SAFETY REQUIREMENTS FOR PLANNING
FOR THE DISPOSAL OF RADIOACTIVE WASTE
3.1. Requirements are established for ensuring that the safety objective and
criteria for disposal facilities set out in Section 2 are fulfilled. The prime
responsibility for safety rests with the operator [1], to whom the majority of the
requirements apply. However, the assurance of safety and the development of a
broader confidence in safety also require a competent regulatory process within a
specified legal and regulatory framework and the allocation of responsibilities for
pre-operational activities.
3.2. The operator
7
might be a single organization or one of a number of
organizations involved, depending on the approach taken in the State. The safety
requirements for the planning of a disposal facility apply to those elements that
have to be in place prior to the development of the disposal facility, with the
purpose of ensuring safety in the operational period and after closure.
6
INTERNATIONAL ATOMIC ENERGY AGENCY, Safety Indicators in Different
Time Frames for the Safety Assessment of Underground Radioactive Waste Repositories,
IAEA-TECDOC-767, IAEA, Vienna (1994).
7
In the IAEA safety standards, ‘operator’ means any organization or person applying
for authorization or authorized and/or responsible for nuclear, radiation, radioactive waste or
transport safety when undertaking activities or in relation to any nuclear facilities or sources of
ionizing radiation. This includes, inter alia, private individuals, governmental bodies,
consignors or carriers, licensees, hospitals, self-employed persons, etc.
17
3.3. Safety in the operation of radioactive waste disposal facilities has to be
achieved by means of a variety of engineered and operational controls similar to
those used in other facilities in which radioactive material is handled, used, stored
or processed. These include the containment and shielding for the radioactive
waste and operational control over time of exposure and proximity to the waste.
Protection of the public is provided for by preventing or controlling releases from
the facility and by controlling access to the site. Operational monitoring
programmes provide assurance of these various controls.
3.4. Safety after closure is achieved by developing a disposal system in which
the various components work together to provide and to ensure the required level
of safety. This approach offers flexibility to the designer of a disposal facility to
adapt the facility’s layout and engineered barriers so as to take advantage of the
natural characteristics of the site and the barrier potential of the host geology, if
applicable. Assurance of confidence in safety is also necessary and this may
require the consideration of a number of complex issues, including the potential
impact of operations on the performance of the disposal facility after closure.
3.5. The requirements in respect of the planning of disposal facilities for
radioactive waste are set out under three headings for the governmental, legal and
regulatory framework, the safety approach and the design concepts for safety.
GOVERNMENTAL, LEGAL AND REGULATORY FRAMEWORK
Requirement 1: Government responsibilities
The government is required to establish and maintain an appropriate
governmental, legal and regulatory framework for safety within which
responsibilities shall be clearly allocated for disposal facilities for radioactive
waste to be sited, designed, constructed, operated and closed. This shall
include: confirmation at a national level of the need for disposal facilities of
different types; specification of the steps in development and licensing of
facilities of different types; and clear allocation of responsibilities, securing
of financial and other resources, and provision of independent regulatory
functions relating to a planned disposal facility.
3.6. This requirement derives from a principle established in the Fundamental
Safety Principles (Ref. [1], Principle 2). It is also stipulated under the terms of the
Joint Convention [2]. Requirements for establishing a national system for
radioactive waste management are established in Ref. [18]. A project for the
18
disposal of radioactive waste, especially for the development of a facility for the
disposal of high level and long lived radioactive waste, has to be given special
consideration within this infrastructure because of the relatively long period of
time necessary for the development of such facilities.
3.7. Matters that have to be considered include:
(a) Defining the national policy for the long term management of radioactive
waste of different types;
(b) Setting clearly defined legal, technical and financial responsibilities for
organizations that are to be involved in the development of facilities for
radioactive waste management, including disposal facilities of all types;
(c) Ensuring the adequacy and security of financial provisions for each
disposal facility;
(d) Defining the overall process for the development, operation and closure of
disposal facilities, including the legal and regulatory requirements (e.g.
licence conditions) at each step, and the processes for decision making and
the involvement of interested parties;
(e) Ensuring that the necessary scientific and technical expertise remains
available both to the operator and for the support of independent regulatory
reviews and other national review functions;
(f) Defining legal, technical and financial responsibilities and, if necessary,
providing for any institutional arrangements that are envisaged after
closure, including monitoring and ensuring the nuclear security of different
types of waste that have been disposed of.
Requirement 2: Responsibilities of the regulatory body
The regulatory body shall establish regulatory requirements for the
development of different types of disposal facility for radioactive waste and
shall set out the procedures for meeting the requirements for the various
stages of the licensing process. It shall also set conditions for the development,
operation and closure of each individual disposal facility and shall carry out
such activities as are necessary to ensure that the conditions are met.
3.8. General standards for the protection of people and the environment are
usually set out in national policy or in legislation. The regulatory body has to
develop regulatory requirements specific to each type of disposal facility for
radioactive waste, including each type that is envisaged, on the basis of national
policy and with due regard to the safety objective and criteria set out in para. 2.15.
The regulatory body has to provide guidance on the interpretation of the national
19
legislation and regulatory requirements, as necessary, and guidance on what is
expected of the operator in respect of each individual disposal facility.
3.9. The regulatory body has to engage in dialogue with waste producers, the
operators of the disposal facility and interested parties to ensure that the
regulatory requirements are appropriate and practicable. It also has to maintain
competent staff, to acquire capabilities for independent assessment and to
undertake international cooperation, as necessary, to fulfil its regulatory
functions.
3.10. The regulatory body has to document the procedures that it uses to evaluate
the safety of each type of disposal facility, the procedures that operators are
expected to follow in the context of licensing, important decisions prior to
licensing and licence applications. It also has to document the procedures that it
follows in reviewing submissions from operators to assess compliance with
regulatory requirements.
3.11. Similarly, in respect of each individual disposal facility, the regulatory body
has to set out the procedures that an operator is expected to follow in
demonstrating compliance with the conditions for the development and operation
of the facility. The regulatory body also has to set out the procedures that it
follows to assess compliance with the conditions throughout all stages of the
development, operation and closure of the facility.
Requirement 3: Responsibilities of the operator
The operator of a disposal facility for radioactive waste shall be responsible
for its safety. The operator shall carry out safety assessment and develop and
maintain a safety case, and shall carry out all the necessary activities for site
selection and evaluation, design, construction, operation, closure and, if
necessary, surveillance after closure, in accordance with national strategy, in
compliance with the regulatory requirements and within the legal and
regulatory infrastructure.
3.12. The operator has to be responsible for developing a disposal facility that is
practicable and safe and for demonstrating its safety, consistent with the
requirements of the regulatory body. This task has to be undertaken in
consideration of: the characteristics and quantities of the radioactive waste to be
disposed of; the site or sites available; the mining, excavation, construction and
engineering techniques available; and the legal and regulatory infrastructure and
regulatory requirements. The operator also has to be responsible for developing a
20
safety case, on the basis of which decisions on the development, operation and
closure of the disposal facility have to be made (see Requirements 17–19).
3.13. The operator has to conduct or commission the research and development
work necessary to ensure that the planned technical operations can be practically
and safely accomplished, and to demonstrate this. The operator likewise has to
conduct or commission the research work necessary to investigate, to understand
and to support the understanding of the processes on which the safety of the
disposal facility depends. The operator also has to carry out all the necessary
investigations of sites and of materials and has to assess their suitability and
obtain all the data necessary for the purposes of safety assessment.
3.14. The operator has to establish technical specifications that are justified by
safety assessment, to ensure that the disposal facility is developed in accordance
with the safety case. This has to include waste acceptance criteria (see
Requirement 20) and other controls and limits to be applied during construction,
operation and closure.
3.15. The operator has to retain all the information relevant to the safety case and
the supporting safety assessment for the disposal facility and has to retain the
inspection records that demonstrate compliance with regulatory requirements and
with the operator’s own specification. Such information and records have to be
retained, at least up until the time when the information is shown to be
superseded, or until responsibility for the disposal facility is passed on to another
organization. This occurs, for example, at closure of the facility, when all relevant
information and records have to be transferred to the organization assuming
responsibility for the facility and its safety.
3.16. The operator has to cooperate with the regulatory body and has to supply all
the information that the regulatory body may request. The need to preserve the
records for long periods of time has to be taken into account in selecting the
format and media to be used for records.
SAFETY APPROACH
Requirement 4: Importance of safety in the process of development and
operation of a disposal facility
Throughout the process of development and operation of a disposal facility
for radioactive waste, an understanding of the relevance and the
21
implications for safety of the available options for the facility shall be
developed by the operator. This is for the purpose of providing an optimized
level of safety in the operational stage and after closure.
3.17. Disposal facilities for radioactive waste may be developed and operated
over a period of several years or several decades. Key decisions, such as
decisions on site selection and evaluation, and on the design, construction,
operation and closure of the disposal facility, are expected to be made as the
project develops. In this process, decisions are made on the basis of the
information available at the time, which may be either quantitative or qualitative,
and the confidence that can be placed in that information.
3.18. Decisions on the development, operation and closure of the facility are
constrained by external factors, which include: national policy and preferences,
the capacity and capability of existing storage and disposal facilities to
accommodate waste, and the availability of suitable sites and geological
formations to host planned new disposal facilities. An adequate level of
confidence in the safety of each disposal facility has to be developed before
decisions are taken.
3.19. At each major decision point, the implications for the safety of the available
design options and operational options for the disposal facility have to be
considered and taken into account. Ensuring safety, both in the operational stage
and after closure, is the overriding concern at each decision point. If more than
one option is capable of providing the required level of safety, then other factors
also have to be considered. These factors could include public acceptability, cost,
site ownership, existing infrastructure and transport routes.
3.20. Consideration has to be given to locating the facility away from significant
known mineral resources, geothermal water and other valuable subsurface
resources. This is to reduce the risk of human intrusion into the site and to reduce
the potential for use of the surrounding area to be in conflict with the facility. The
safety of the facility has to be considered at every step in the decision making
process to ensure that safety is optimized in the sense discussed in the Appendix.
Requirement 5: Passive means for the safety of the disposal facility
The operator shall evaluate the site and shall design, construct, operate and
close the disposal facility in such a way that safety is ensured by passive
means to the fullest extent possible and the need for actions to be taken after
closure of the facility is minimized.
22
3.21. In the operational stage of a disposal facility for radioactive waste, certain
active control measures have to be applied. However, where passive features such
as the shielding and containment provided by the packaging material can provide
safety, then safety has to be ensured by such passive means.
3.22. To some extent, the safety of a disposal facility can depend on some future
actions such as maintenance work or surveillance. However, this dependence has
to be minimized to the extent possible. This is necessary because of the
possibility that safety measures that depend on future actions, such as
maintenance work or surveillance, will not be taken or will not be continued. The
cumulative probability of the failure of such safety measures will gradually
increase. Furthermore, and consistent with the Fundamental Safety Principles [1],
disposal of radioactive waste is intended to discharge the responsibility for safety
of the waste producers and the operator to the fullest extent possible, thereby
minimizing the responsibilities that are retained or are passed on to successor
organizations.
3.23. For a geological disposal facility, it is possible to provide for safety after
closure by means of passive features. It is likewise possible to provide for the
safety of a borehole disposal facility after closure by means of passive features,
owing to the host geology. In the case of a near surface disposal facility, actions
such as maintenance, monitoring or surveillance may be necessary for a period of
time after closure to ensure safety.
3.24. Providing for the safety of a disposal facility after closure by means of
passive features will entail proper closure of the facility and ending the need for
its active management. The cessation of management means that the disposal
facility, with its associated radiological hazard, is no longer under active control.
It is the performance of the natural and engineered barriers that provides safety after
closure, together, for a near surface disposal facility, with institutional controls.
3.25. In practice, even in those cases in which passive features are the primary
means for providing a reasonable assurance of safety, institutional controls,
including restrictions on land use, and a programme for monitoring may be
necessary in the post-closure period. Institutional controls and monitoring are the
subject of Requirements 21 and 22.
Requirement 6: Understanding of a disposal facility and confidence in safety
The operator of a disposal facility shall develop an adequate understanding
of the features of the facility and its host environment and of the factors that
23
influence its safety after closure over suitably long time periods, so that a
sufficient level of confidence in safety can be achieved.
3.26. Confidence has to be assured by the results of safety assessment for a
disposal facility. The features of the facility and its host environment that provide
for safety have to be identified, in addition to those factors that might be
detrimental. It has to be demonstrated that these features and factors are
sufficiently well characterized and understood. Any uncertainties have to be
taken into consideration in the assessment of safety.
3.27. The purpose of this demonstration is to establish, with a high level of
confidence, that the disposal facility and its host environment can be relied on to
provide the necessary containment and isolation over the timescales envisaged.
Certain features of the disposal facility and its environment may contribute to
safety, but may be less quantifiable, such as the remoteness of the site. The
reasoning with regard to such factors has to be based on more qualitative
arguments, and such factors provide a safety margin.
3.28. An understanding of the features of a disposal facility and how it will
perform over time is necessary in order to be able to demonstrate the
dependability of certain design features. This demonstration is assisted if such
design features are robust (i.e. their performance is of low sensitivity to possible
events and processes causing disturbances). Sufficient evidence has to be
obtained of their feasibility and effectiveness before construction activities are
commenced.
3.29. In this regard, the range of possible events and processes causing
disturbances that it is reasonable to include in such considerations has to be
subject to agreement by the regulatory body and subsequent approval by
inclusion in the safety case. These considerations permit the development of an
understanding of whether or not such events and processes cause disturbances
that could lead to the widespread loss of safety functions.
3.30. Understanding of the performance of the disposal system and its safety
features and processes evolves as more data are accumulated and scientific
knowledge is developed. Early in the development of the concept, the data
obtained and the level of understanding gained have to assure sufficient
confidence to be able to commit resources for further investigations. Before the
start of construction, during emplacement of waste and at closure of the facility,
the level of understanding has to be sufficient to support the safety case for
24
fulfilling the regulatory requirements applicable for the particular stage of the
project.
3.31. In establishing these regulatory requirements, it has to be recognized that
there are various types and components of uncertainty inherent in modelling
complex environmental systems. It also has to be recognized that there are,
inevitably, significant uncertainties associated with projecting the performance of
a disposal system over time.
DESIGN CONCEPTS FOR SAFETY
3.32. A disposal facility is designed to contain the radionuclides associated with
the radioactive waste and to isolate them from the accessible biosphere. The
disposal facility is also designed to retard the dispersion of radionuclides in the
geosphere and biosphere and to provide isolation of the waste from aggressive
phenomena that could degrade the integrity of the facility. The various elements
of the disposal system, including physical components and control procedures,
contribute to performing safety functions in different ways over different
timescales.
3.33. Requirements are established in this section for ensuring that there is
adequate defence in depth, so that safety is not unduly dependent on a single
element of the disposal facility, such as the waste package; or a single control
measure, such as verification of the inventory of waste packages; or the
fulfilment of a single safety function, such as by containment of radionuclides or
retardation of migration; or a single administrative procedure, such as a
procedure for site access control or for maintenance of the facility.
3.34. Adequate defence in depth has to be ensured by demonstrating that there are
multiple safety functions, that the fulfilment of individual safety functions is
robust and that the performance of the various physical components of the
disposal system and the safety functions they fulfil can be relied upon, as
assumed in the safety case and supporting safety assessment. It is the
responsibility of the operator to demonstrate fulfilment of the following design
requirements to the satisfaction of the regulatory body.
Requirement 7: Multiple safety functions
The host environment shall be selected, the engineered barriers of the
disposal facility shall be designed and the facility shall be operated to ensure
25
that safety is provided by means of multiple safety functions. Containment
and isolation of the waste shall be provided by means of a number of
physical barriers of the disposal system. The performance of these physical
barriers shall be achieved by means of diverse physical and chemical
processes together with various operational controls. The capability of the
individual barriers and controls together with that of the overall disposal
system to perform as assumed in the safety case shall be demonstrated. The
overall performance of the disposal system shall not be unduly dependent on
a single safety function.
3.35. The engineered and physical barriers that make up the disposal system are
physical entities, such as the waste form, the packaging, the backfill, and the host
environment and geological formation. A safety function may be provided by
means of a physical or chemical property or process that contributes to
containment and isolation, such as: impermeability to water; limited corrosion,
dissolution, leach rate and solubility; retention of radionuclides; and retardation
of radionuclide migration.
3.36. Active controls can also fulfil safety functions or contribute to confidence
in natural and engineered barriers and safety functions. The presence of a number
of physical and other elements performing safety functions gives assurance that
even if any of them do not perform fully as expected (e.g. owing to an unexpected
process or an unlikely event), a sufficient margin of safety will remain.
3.37. The physical elements and their safety functions can be complementary and
can work in combination. The performance of a disposal system is thus
dependent on different physical elements and on other elements that perform
safety functions, which act over different time periods. For example, the roles of
the waste package and the host geological formation for a geological disposal
facility may vary in different time periods.
3.38. The safety case has to explain and justify the functions performed by each
physical element and other features. It also has to identify the time periods over
which physical components and other features are expected to perform their
various safety functions, and also the alternative or additional safety functions
that are available if a physical element does not fully perform or another safety
function is not fulfilled.
26
Requirement 8: Containment of radioactive waste
The engineered barriers, including the waste form and packaging, shall be
designed, and the host environment shall be selected, so as to provide
containment of the radionuclides associated with the waste. Containment
shall be provided until radioactive decay has significantly reduced the
hazard posed by the waste. In addition, in the case of heat generating waste,
containment shall be provided while the waste is still producing heat energy
in amounts that could adversely affect the performance of the disposal
system.
3.39. The containment of radioactive waste implies designing the disposal
facility to avoid or minimize the release of radionuclides. Releases of small
amounts of gaseous radionuclides and of small fractions of other highly mobile
species from some types of radioactive waste may be inevitable. Such releases,
nevertheless, have to be demonstrated to be acceptable by means of safety
assessment. The containment may be provided by the characteristics of the waste
form and the packaging and by the characteristics of other engineered
components of the disposal system and the host environment and geological
formation.
3.40. The containment of the radionuclides in the waste form and the packaging
over a defined period has to ensure that the majority of shorter lived radionuclides
decay in situ. For low level waste, such periods would be of the order of several
hundred years; for high level waste the period would be several thousands of
years. For high level waste, it also has to be ensured that any migration of
radionuclides outside the disposal system would occur only after the heat
produced by radioactive decay has substantially decreased.
3.41. Radioactive waste from mining and mineral processing may include
radionuclides with very long half-lives. Providing assurance of the integrity of
the containment features of disposal facilities for such waste over the
corresponding timescales requires particular consideration. If the waste has
activity levels for which the dose and/or risk criteria for human intrusion into
such facilities (see para. 2.15) might be exceeded, alternative disposal options
will have to be considered. Possible alternative options include, for example,
disposal of the waste below the surface, or separation of the radionuclide content
giving rise to the higher dose, as determined by the safety case for the disposal
facility.
27
3.42. Containment is most important for more highly concentrated radioactive
waste, such as intermediate level waste and vitrified waste from fuel
reprocessing, or for spent nuclear fuel. Attention also has to be given to the
durability of the waste form. The most highly concentrated waste has to be
emplaced in a containment configuration that is designed to retain its integrity for
a long enough period of time to enable most of the shorter lived radionuclides to
decay and for the associated generation of heat to decrease substantially. Such
containment may not be practicable or necessary for low level waste. The
containment capability of the waste package has to be demonstrated by means of
safety assessment to be appropriate for the waste type and the overall disposal
system.
Requirement 9: Isolation of radioactive waste
The disposal facility shall be sited, designed and operated to provide features
that are aimed at isolation of the radioactive waste from people and from the
accessible biosphere. The features shall aim to provide isolation for several
hundreds of years for short lived waste and at least several thousand years
for intermediate and high level waste. In so doing, consideration shall be
given to both the natural evolution of the disposal system and events causing
disturbance of the facility.
3.43. For near surface facilities, isolation has to be provided by the location and
the design of the disposal facility and by operational and institutional controls.
For geological disposal of radioactive waste, isolation is provided primarily by
the host geological formation as a consequence of the depth of disposal.
3.44. Isolation means design to keep the waste and its associated hazard apart
from the accessible biosphere. It also means design to minimize the influence of
factors that could reduce the integrity of the disposal facility. Sites and locations
with higher hydraulic conductivities have to be avoided. Access to waste has to
be made difficult to gain without, for example, violation of institutional controls
for near surface disposal. Isolation also means providing for a very slow mobility
of radionuclides to impede migration from disposal facilities.
3.45. Location of a disposal facility in a stable geological formation provides
protection of the facility from the effects of geomorphological processes, such as
erosion and glaciation. The disposal facility has to be located away from known
areas of significant underground mineral resources or other valuable resources.
This will reduce the likelihood of inadvertent disturbance of the facility and will
avoid resources being made unavailable for exploitation.
28
3.46. In some cases, it may not be possible to provide sufficient assurance of
separation from the accessible biosphere, owing to phenomena such as uplift,
erosion and glaciation. In such cases, and if the remaining activity in the waste is
still significant at the time such phenomena occur, the possibility of human
intrusion has to be evaluated in determining the degree of isolation provided.
3.47. Over time periods of several thousand years or more, the migration of a
fraction of the longer lived and more mobile radionuclides from the waste in a
geological disposal facility (or in other facilities that may include longer lived
radionuclides, such as borehole facilities) may be inevitable. The safety criteria to
apply in assessing such possible releases are set out in para. 2.15. Caution needs
to be exercised in applying criteria for periods far into the future. Beyond such
timescales, the uncertainties associated with dose estimates become so large that
the criteria might no longer serve as a reasonable basis for decision making. For
such long time periods after closure, indicators of safety other than estimates of
dose or individual risk may be appropriate, and their use has to be considered.
Requirement 10: Surveillance and control of passive safety features
An appropriate level of surveillance and control shall be applied to protect
and preserve the passive safety features, to the extent that this is necessary,
so that they can fulfil the functions that they are assigned in the safety case
for safety after closure.
3.48. For geological disposal and for the disposal of intermediate level
radioactive waste, the passive safety features (barriers) have to be sufficiently
robust so as not to require repair or upgrading. The long term safety of a disposal
facility for radioactive waste is required not to be dependent on active
institutional control (see Requirement 22). For near surface disposal facilities,
including those for radioactive waste from the mining and processing of minerals,
measures for surveillance and control of the disposal facility might be instituted.
These measures might include restrictions on access by people and animals,
inspection of physical conditions, retention of appropriate maintenance
capabilities, and surveillance and monitoring as a method of checking whether
performance is as specified (i.e. checking for degradation). The intent of
surveillance and monitoring is not to measure radiological parameters but to
ensure the continuing fulfilment of safety functions.
29
4. REQUIREMENTS FOR THE DEVELOPMENT,
OPERATION AND CLOSURE OF A DISPOSAL FACILITY
4.1. Section 4 establishes safety requirements relating to the step by step
implementation of the planning measures mentioned previously that are
necessary for safety and to assist in developing confidence in the safety of
disposal facilities. The requirements are set out under three headings:
(i) framework for disposal of radioactive waste; (ii) the safety case and safety
assessment; and (iii) steps in the development, operation and closure of disposal
facilities.
FRAMEWORK FOR DISPOSAL OF RADIOACTIVE WASTE
Requirement 11: Step by step development and evaluation of disposal
facilities
Disposal facilities for radioactive waste shall be developed, operated and
closed in a series of steps. Each of these steps shall be supported, as
necessary, by iterative evaluations of the site, of the options for design,
construction, operation and management, and of the performance and safety
of the disposal system.
4.2. A step by step approach to the development of a disposal facility for
radioactive waste refers to the steps that are imposed by the regulatory body and
by political decision making processes (see para. 1.18). This approach is taken to
provide an opportunity to ensure the quality of the technical programme and the
associated decision making. For the operator, it provides a framework in which
sufficient confidence in the technical feasibility and safety of the disposal facility
can be built at each step in its development.
4.3. Confidence has to be developed and refined by means of iterative design
and safety studies as the project progresses [19]. The process has to provide for:
the collection, analysis and interpretation of the relevant scientific and technical
data; the development of designs and operational plans; and the development of
the safety case for safety in the operational stage and after closure. The step by
step process provides access for all interested parties to the safety basis for the
disposal facility. This facilitates the relevant decision making processes that
enable the operator to proceed to the next significant step in the development of
the facility, and on to its operation and, finally, its closure.
30
4.4. The step by step approach to the development of a disposal facility also
allows opportunities for independent technical review, regulatory review, and
political and public involvement in the process. The nature of the reviews and
involvement will depend on national practices and on the facility in question.
Technical reviews by, or on behalf of, the operator and the regulatory body may
focus on site selection and evaluation and design options, the adequacy of the
scientific basis and analyses, and whether safety standards and requirements have
been met.
4.5. Alternative waste management options, the site selection and evaluation
process and aspects of public acceptability, for example, may be considered in
farther reaching reviews. Technical reviews have to be undertaken prior to
selection of a disposal option, prior to selection of a site, prior to construction and
prior to operation. Periodic reviews also have to be undertaken during the
operation of the facility and following closure, up to termination of the facility
licence.
THE SAFETY CASE AND SAFETY ASSESSMENT
4.6. The development of a safety case and supporting safety assessment for
review by the regulatory body and interested parties is central to the
development, operation and closure of a disposal facility for radioactive waste.
The safety case substantiates the safety of the disposal facility and contributes to
confidence in its safety. The safety case is an essential input to all important
decisions concerning the disposal facility. It has to provide the basis for
understanding the disposal system and how it will behave over time. It has to
address site aspects and engineering aspects, providing the logic and rationale for
the design, and has to be supported by safety assessment. It also has to address the
management system put in place to ensure quality for all aspects important to
safety.
4.7. At any step in the development of a disposal facility, the safety case also has
to identify and acknowledge the unresolved uncertainties that exist at that stage
and their safety significance, and approaches for their management.
4.8. The safety case has to include the output of the safety assessment (see
paras 4.9–4.11), together with additional information, including supporting
evidence and reasoning on the robustness and reliability of the facility, its design,
the logic of the design, and the quality of safety assessment and underlying
assumptions.
31
4.9. The safety case may also include more general arguments relating to the
disposal of radioactive waste and information to put the results of safety
assessment into perspective. Any unresolved issues at any step in the
development or in the operation or closure of the facility have to be
acknowledged in the safety case and guidance has to be provided for work to
resolve these issues.
4.10. Safety assessment is the process of systematically analysing the hazards
associated with a disposal facility and assessing the ability of the site and the
design of the facility to provide for the fulfilment of safety functions and to meet
technical requirements. Safety assessment has to include quantification of the
overall level of performance, analysis of the associated uncertainties and
comparison with the relevant design requirements and safety standards. The
assessments have to be site specific since the host environment of a disposal
system, in contrast to engineered systems, cannot be standardized.
4.11. As site investigations and design studies progress, safety assessment will
become increasingly refined and specific to the site. At the end of a site
investigation, sufficient data have to be available for a complete assessment. Any
significant deficiencies in scientific understanding, data or analysis that might
affect the results presented also have to be identified in the safety assessment.
Depending on the stage of development of the facility, safety assessment may be
used in focusing research, and its results may be used to assess compliance with
the safety objective and safety criteria.
Requirement 12: Preparation, approval and use of the safety case and safety
assessment for a disposal facility
A safety case and supporting safety assessment shall be prepared and
updated by the operator, as necessary, at each step in the development of a
disposal facility, in operation and after closure. The safety case and
supporting safety assessment shall be submitted to the regulatory body for
approval. The safety case and supporting safety assessment shall be
sufficiently detailed and comprehensive to provide the necessary technical
input for informing the regulatory body and for informing the decisions
necessary at each step.
4.12. A facility specific safety case has to be prepared early in the development of
a disposal facility to provide a basis for licensing decisions and to guide activities
in research and development, site selection and evaluation and design. The safety
case has to be developed progressively and elaborated as the project proceeds. It
32
has to be presented to the regulatory body at each step in the development of the
disposal facility. The regulatory body might require an update of, or revision to,
the safety case before given steps can be taken, or such an update or revision may
be necessary to gain political or public support for taking the next step in the
development of the disposal facility or for its operation or closure. The formality
and level of technical detail of the safety case will depend on the stage of
development of the project, the decision in hand, the audience to which it is
addressed and specific national requirements.
4.13. Safety assessment in support of the safety case has to be performed and
updated throughout the development and operation of the disposal facility and as
more refined site data become available. Safety assessment has to provide input
to ongoing decision making by the operator. Such decision making may relate to
subjects for research, development of a capability for assessment, allocation of
resources and development of waste acceptance criteria.
4.14. Safety assessment also has to identify key processes relevant to safety and
to contribute to the development of an understanding of the performance of
disposal facilities. It has to support judgements with regard to alternative
management options as an element of optimizing protection and safety. Such an
understanding has to provide the basis for the safety arguments presented in the
safety case. The operator has to decide on the timing and the level of detail of the
safety assessment, in consultation with, and subject to the approval of, the
regulatory body.
Requirement 13: Scope of the safety case and safety assessment
The safety case for a disposal facility shall describe all safety relevant aspects
of the site, the design of the facility and the managerial control measures and
regulatory controls. The safety case and supporting safety assessment shall
demonstrate the level of protection of people and the environment provided
and shall provide assurance to the regulatory body and other interested
parties that safety requirements will be met.
4.15. The safety case for a disposal facility has to address safety both in operation
and after closure. It may also address safety in transport, for which requirements
are established in Ref. [17]. All aspects of operation relevant to safety are
considered, including surface and underground excavation, construction and
mining work, waste emplacement, and backfilling, sealing and closing
operations. Consideration has to be given to both occupational exposure and
33
public exposure resulting from conditions of normal operation and anticipated
operational occurrences over the operating lifetime of the disposal facility.
4.16. Accidents of a lesser frequency, but with significant radiological
consequences (i.e. possible accidents that could give rise to radiation doses over
the short term in excess of annual dose limits (see Section 2)), have to be
considered with regard to both their likelihood of occurrence and the magnitude
of possible radiation doses. The adequacy of the design and of the operational
features also has to be evaluated.
4.17. With regard to safety after closure, the expected range of possible
developments affecting the disposal system and events that might affect its
performance, including those of low probability, have to be considered in the
safety case and supporting assessment by the following means:
(a) By presenting evidence that the disposal system, its possible evolutions and
events that might affect it are sufficiently well understood;
(b) By demonstrating the feasibility of implementing the design;
(c) By providing convincing estimates of the performance of the disposal
system and a reasonable level of assurance that all the relevant safety
requirements will be complied with and that radiation protection has been
optimized;
(d) By identifying and presenting an analysis of the associated uncertainties.
4.18. The safety case may include the presentation of multiple lines of reasoning
based, for example, on studies of natural analogues and palaeohydrogeological
studies, suitable characteristics of the site, properties of the host geological
formation, engineering considerations, operational procedures and institutional
assurances.
4.19. The performance of the disposal system under expected and less likely
evolutions and events, which can be outside the design performance range of the
disposal facility, has to be analysed in the safety assessment. A judgement of
what is to be considered the expected evolution and less likely evolutions has to
be discussed between the regulatory body and the operator. If necessary,
sensitivity analyses and uncertainty analyses would be undertaken to gain an
understanding of the performance of the disposal system and its components
under a range of evolutions and events.
34
4.20. The consequences of unexpected events and processes may be explored to
test the robustness of the disposal system. In particular, the resilience of the
disposal system has to be assessed. Quantitative analyses have to be undertaken,
at least over the time period for which regulatory requirements apply. However,
the results from detailed models for safety assessment purposes are likely to be
more uncertain for timescales extending into the far future.
4.21. For timescales extending into the far future, arguments may be needed to
illustrate safety, on the basis, for example, of complementary safety indicators,
such as concentrations and fluxes of radionuclides of natural origin in the
geosphere and biosphere and bounding analyses. While such assessments cannot
yield precise levels of possible doses or risks, the results may provide a tool to
indicate the level of safety and verify that no alternative design would have
obvious advantages.
4.22. The management systems established to provide assurance of quality in
these design features and operational features have to be addressed in the safety
case.
Requirement 14: Documentation of the safety case and safety assessment
The safety case and supporting safety assessment for a disposal facility shall
be documented to a level of detail and quality sufficient to inform and
support the decision to be made at each step and to allow for independent
review of the safety case and supporting safety assessment.
4.23. The necessary scope and structure of the documentation setting out the
safety case and supporting safety assessment will depend on the step reached in
the project for the disposal facility and on national requirements. This includes
consideration of the needs of different interested parties for information.
Important considerations in documenting the safety case and supporting safety
assessment are justification, traceability and clarity.
4.24. Justification concerns explaining the basis for the choices that have been
made and the arguments for and against the decisions, especially those decisions
concerning the main arguments for safety. Traceability concerns the ability of an
independent qualified person to follow what has been done. The traceability has
to enable technical and regulatory review. Justification and traceability both
require a well-documented record of the decisions made and the assumptions
made in the development and operation of a disposal facility, and of the models
35
and data used in deriving a particular set of results for safety assessment
purposes.
4.25. Clarity concerns good structure and presentation at an appropriate level of
detail so as to allow an understanding of the safety arguments. This requires the
results of work to be presented in the documents in such a way that interested
parties for whom the material is intended can gain a good understanding of the
safety arguments and their basis. Different types and styles of document may be
necessary to provide material that is useful to different parties.
STEPS IN THE DEVELOPMENT, OPERATION AND CLOSURE OF A
DISPOSAL FACILITY
Requirement 15: Site characterization for a disposal facility
The site for a disposal facility shall be characterized at a level of detail
sufficient to support a general understanding of both the characteristics of
the site and how the site will evolve over time. This shall include its present
condition, its probable natural evolution and possible natural events, and
also human plans and actions in the vicinity that may affect the safety of the
facility over the period of interest. It shall also include a specific
understanding of the impact on safety of features, events and processes
associated with the site and the facility.
4.26. An understanding of the site for a disposal facility has to be gained in order
to present a convincing scientific description of the disposal system on which the
more conceptual descriptions that are used in the safety assessment can be based.
The focus has to be on features, events and processes relating to the site that could
have an impact on safety and that are addressed in the safety case and supporting
safety assessment. In particular, this has to demonstrate that there is adequate
geological, geomorphological or topographical stability (as appropriate to the
type of facility), and features and processes that contribute to safety. It also has to
demonstrate that other features, events and processes do not undermine the safety
case.
4.27. Characterization of the geological aspects has to include activities such as
the investigation of: long term stability, faulting and the extent of fracturing in the
host geological formation; seismicity; volcanism; the volume of rock suitable for
the construction of disposal zones; geotechnical parameters relevant to the
design; groundwater flow regimes; geochemical conditions; and mineralogy. The
36
extent of characterization necessary will depend on the types of disposal facility
and the site in question.
4.28. A graded approach has to be adopted, depending on the hazard potential of
the waste and the complexity of the site and disposal facility design, in
accordance with the guidance cited in footnote 5. Site characterization
undertaken in an iterative manner has to provide input to, and has, in turn, to be
guided by, the safety case. Additionally, investigation of, for example, natural
background radiation and the radionuclide content in soil, groundwater and other
media may contribute to a better understanding of the characteristics of the site of
the disposal facility. It may also assist in the evaluation of radiological impacts on
the environment by providing a reference for future comparisons.
4.29. Characterization of the surface environmental features has to include
natural aspects, such as hydrological and meteorological aspects and flora and
fauna. It also has to cover human activities in the vicinity of the site relating to
normal residential settlement patterns and industrial and agricultural activities.
Due regard has to be given to the probable natural evolution of the site, including
effects of erosion and climate change.
Requirement 16: Design of a disposal facility
The disposal facility and its engineered barriers shall be designed to contain
the waste with its associated hazard, to be physically and chemically
compatible with the host geological formation and/or surface environment,
and to provide safety features after closure that complement those features
afforded by the host environment. The facility and its engineered barriers
shall be designed to provide safety during the operational period.
4.30. The designs of disposal facilities for radioactive waste may differ widely,
depending on the types of waste to be disposed of and the host geological
formation and/or surface environment. In general, optimal use has to be made of
the safety features offered by the host environment. This has to be done by
designing a disposal facility that does not cause unacceptable long term
disturbance of the site, is itself protected by the site and performs safety functions
that complement the natural barriers.
4.31. The layout has to be designed so that waste is emplaced in the most suitable
locations. In the event that fissile materials are present in the waste, maintaining a
subcritical configuration has to be part of the design considerations. Key features,
such as shafts and seals in geological disposal facilities, have to be appropriately
37
located. Materials used in the facility have to be resistant to degradation under the
conditions prevailing in the facility (e.g. conditions of chemistry and
temperature) and selected also to limit any undesirable impacts on the safety
functions of any element of the disposal system.
4.32. Disposal facilities, in particular disposal facilities for high level and
intermediate level waste, are expected to perform over much longer timescales
than the periods usually considered in engineering applications. Investigation of
the ways in which analogous natural materials have behaved in geological
formations in nature, or how ancient artefacts and structures have behaved over
time, may contribute to confidence in the assessment of long term performance.
Demonstration of the feasibility of fabrication of waste containers and of the
construction of engineered barriers with the necessary features, for example, in
underground laboratories, is important for the purpose of assessment and for
contributing to confidence that an adequate level of performance can be achieved.
Requirement 17: Construction of a disposal facility
The disposal facility shall be constructed in accordance with the design as
described in the approved safety case and supporting safety assessment. It
shall be constructed in such a way as to preserve the safety functions of the
host environment that have been shown by the safety case to be important
for safety after closure. Construction activities shall be carried out in such a
way as to ensure safety during the operational period.
4.33. Construction of a disposal facility can be a complex technical undertaking
that might be constrained, particularly if it is carried out underground, by the
conditions and the properties of the host geological formation and by the
techniques that are available for underground excavation and construction. An
adequate level of characterization has to be completed before construction is
begun. Excavation and construction activities have to be carried out in such a way
as to avoid unnecessary disturbance of the host environment. Sufficient flexibility
in engineering techniques has to be adopted to allow for variations to be
encountered, such as variations in rock conditions or groundwater conditions in
underground facilities.
4.34. Excavation and construction of a disposal facility could continue after the
commencement of operation of part of the facility and after the emplacement of
waste packages. Such overlapping of construction and operational activities has
to be planned and carried out so as to ensure safety, both in operation and after
closure.
38
Requirement 18: Operation of a disposal facility
The disposal facility shall be operated in accordance with the conditions of
the licence and the relevant regulatory requirements so as to maintain safety
during the operational period and in such a manner as to preserve the safety
functions assumed in the safety case that are important to safety after
closure.
4.35. All operations and activities important to the safety of a disposal facility
have to be subjected to limitations and controls and emergency plans have to be
put in place. The various procedures and plans have to be documented and the
documentation has to be subject to appropriate control procedures [13]. The
safety case has to address and justify both the design and the operational
management arrangements that are used to ensure that the safety objective and
criteria set out in Section 2 are met. Additional, facility specific criteria may be
established by the regulatory body or by the operator.
4.36. The safety case also has to demonstrate that hazards and other radiation
risks to workers and to members of the public under conditions of normal
operation and anticipated operational occurrences have been reduced as low as
reasonably achievable. Active control of safety has to be maintained for as long
as the disposal facility remains unsealed, and this may include an extended period
after the emplacement of waste and before the final closure of the facility.
4.37. Fissile material, when present, has to be managed and has to be emplaced in
the disposal facility in a configuration that will remain subcritical. This may be
achieved by various means, including the appropriate distribution of fissile
material during the conditioning of the waste and the proper design of the waste
packages. Assessments have to be undertaken of the possible evolution of the
criticality hazard after waste emplacement, including after closure.
Requirement 19: Closure of a disposal facility
A disposal facility shall be closed in a way that provides for those safety
functions that have been shown by the safety case to be important after
closure. Plans for closure, including the transition from active management
of the facility, shall be well defined and practicable, so that closure can be
carried out safely at an appropriate time.
4.38. The safety of a disposal facility after closure will depend on a number of
activities and design features, which can include the backfilling and sealing or
39
capping of the disposal facility. Closure has to be considered in the initial design
of the facility, and plans for closure and seal or cap designs have to be updated as
the design of the facility is developed. Before construction activities commence,
there has to be sufficient evidence that the performance of the backfilling, sealing
and capping will function as intended to meet the design requirements.
4.39. The disposal facility has to be closed in accordance with the conditions set
for closure by the regulatory body in the facility’s authorization, with particular
consideration given to any changes in responsibility that may occur at this stage.
Consistent with this, the installation of closure features may be performed in
parallel with waste emplacement operations.
4.40. Backfilling and the placement of seals or caps may be delayed for a period
after the completion of waste emplacement, for example, to allow for monitoring
to assess aspects relating to safety after closure or for reasons relating to public
acceptability. If such features are not to be put in place for a period of time after
the completion of waste emplacement, then the implications for safety during
operation and after closure have to be considered in the safety case.
4.41. Availability of the necessary technical and financial resources to achieve
closure has to be assured by means of Requirements 1–3.
5. ASSURANCE OF SAFETY
Requirement 20: Waste acceptance in a disposal facility
Waste packages and unpackaged waste accepted for emplacement in a
disposal facility shall conform to criteria that are fully consistent with, and
are derived from, the safety case for the disposal facility in operation and
after closure.
5.1. Waste acceptance requirements and criteria for a given disposal facility
have to ensure the safe handling of waste packages and unpackaged waste in
conditions of normal operation and anticipated operational occurrences. They
also have to ensure the fulfilment of the safety functions for the waste form and
waste packaging with regard to safety in the long term. Examples of possible
parameters for waste acceptance criteria include the characteristics and
40
performance requirements of the waste packages and the unpackaged waste to be
disposed of, such as the radionuclide content or activity limits, the heat output
and the properties of the waste form and packaging.
5.2. Modelling and/or testing of the behaviour of waste forms has to be
undertaken to ensure the physical and chemical stability of the different waste
packages and unpackaged waste under the conditions expected in the disposal
facility, and to ensure their adequate performance in the event of anticipated
operational occurrences or accidents.
5.3. Waste intended for disposal has to be characterized to provide sufficient
information to ensure compliance with waste acceptance requirements and
criteria. Arrangements have to be put in place to verify that the waste and waste
packages received for disposal comply with these requirements and criteria and,
if not, to confirm that corrective measures are taken by the generator of the waste
or the operator of the disposal facility. Quality control of waste packages has to be
undertaken and is achieved mainly on the basis of records, preconditioning
testing (e.g. of containers) and control of the conditioning process.
Post-conditioning testing and the need for corrective measures have to be limited
as far as practicable.
Requirement 21: Monitoring programmes at a disposal facility
A programme of monitoring shall be carried out prior to, and during, the
construction and operation of a disposal facility and after its closure, if this is
part of the safety case. This programme shall be designed to collect and
update information necessary for the purposes of protection and safety.
Information shall be obtained to confirm the conditions necessary for the
safety of workers and members of the public and protection of the
environment during the period of operation of the facility. Monitoring shall
also be carried out to confirm the absence of any conditions that could affect
the safety of the facility after closure.
5.4. Monitoring has to be carried out at each step in the development and in the
operation of a disposal facility. The purposes of the monitoring programme
include:
(a) Obtaining information for subsequent assessments;
(b) Assurance of operational safety;
41
(c) Assurance that conditions at the facility for operation are consistent with
the safety assessment;
(d) Confirmation that conditions are consistent with safety after closure.
Guidance is provided in Ref. [20]. Monitoring programmes have to be designed
and implemented so as not to reduce the overall level of safety of the facility after
closure.
5.5. A discussion of monitoring relating to the safety of geological disposal
facilities after closure is given in an IAEA TECDOC
8
. Plans for monitoring with
the aim of providing assurance of safety after closure have to be drawn up before
the construction of a geological disposal facility to indicate possible monitoring
strategies. However, plans have to remain flexible and, if necessary, they will
have to be revised and updated during the development and operation of the
facility.
Requirement 22: The period after closure and institutional controls
Plans shall be prepared for the period after closure to address institutional
control and the arrangements for maintaining the availability of information
on the disposal facility. These plans shall be consistent with passive safety
features and shall form part of the safety case on which authorization to
close the facility is granted.
5.6. The long term safety of a disposal facility for radioactive waste has not to
be dependent on active institutional control. Even the violation of passive safety
features cannot give rise to the criteria for intervention being exceeded.
Additionally, the safety of the disposal facility has not to be dependent solely on
institutional controls. Institutional controls cannot be the sole or main component
of safety for a near surface disposal facility. The ability of the institutional
controls to provide the contributions to safety envisaged in the safety case has to
be demonstrated and justified in the safety case.
5.7. The risk of intrusion into a disposal facility for radioactive waste may be
reduced over a longer timescale than that foreseen for active controls by the use
of passive controls, such as the preservation of information by the use of markers
and archives, including international archives.
8
INTERNATIONAL ATOMIC ENERGY AGENCY, Monitoring of Geological
Repositories for High Level Radioactive Waste, IAEA-TECDOC-1208, IAEA, Vienna (2001).
42
5.8. Institutional controls over a disposal facility for radioactive waste have to
provide additional assurance of the safety and nuclear security of the facility.
Examples include provision for preventing access to the site by intruders and
post-operational monitoring capable of providing early warning of the migration
of radionuclides from the disposal facility before they reach the site boundary.
5.9. Near surface disposal facilities are generally designed on the assumption
that institutional control has to remain in force for a period of time. For short
lived waste, the period will have to be several tens to hundreds of years following
closure. Such controls will be either active or passive in nature. For near surface
disposal of waste from mining and mineral processing that includes very long
lived radionuclides, and which generally comprises large volumes, activity
concentrations have to be limited so that ongoing active institutional control does
not have to be relied on as a safety measure. Waste with activity concentrations
above the limitations has to be disposed of below the ground surface.
5.10. The status of a disposal facility beyond the period of active institutional
control differs from the release of a nuclear installation site from regulatory
control after decommissioning inasmuch as release of the site of a disposal
facility for unrestricted use is generally not contemplated. The site location and
the facility design have to reduce the likelihood of intrusion.
5.11. For near surface disposal facilities, the waste acceptance criteria will limit
any consequences of human intrusion to within the specified criteria (see
para. 2.15), even if control over the site is lost. The dose constraint (see
para. 2.15) adopted for doses to members of the public applies for the anticipated
normal evolution of the site following the period of institutional control.
5.12. Geological disposal facilities have not to be dependent on long term
institutional control after closure as a safety measure (see Requirement 5).
Nevertheless, institutional controls may contribute to safety by preventing or
reducing the likelihood of human actions that could inadvertently interfere with the
waste or degrade the safety features of the geological disposal system. Institutional
controls may also contribute to increasing public acceptance of geological disposal.
5.13. Disposal facilities may not be closed for several tens of years or more after
operations have commenced. Plans for possible future controls and the period
over which they would be applied may initially be flexible and conceptual in
nature, but plans have to be developed and refined as the facility approaches
closure. Consideration has to be given to: local land use controls; site restrictions
or surveillance and monitoring; local, national and international records; and the
43
use of durable surface and/or subsurface markers. Arrangements have to be made
to be able to pass on information about the disposal facility and its contents to
future generations to enable any future decisions on the disposal facility and its
safety to be made.
5.14. While the facility remains licensed, the operator has to provide institutional
controls. It is envisaged that the responsibility for whatever passive measures for
institutional control are necessary following termination of the licence will have
to revert to the government at some level.
Requirement 23: Consideration of the State system of accounting for, and
control of, nuclear material
9
In the design and operation of disposal facilities subject to agreements on
accounting for, and control of, nuclear material, consideration shall be given
to ensuring that safety is not compromised by the measures required under
the system of accounting for, and control of, nuclear material [21–23].
5.15. The system of accounting for, and control of, nuclear material applies to
materials that include significant quantities of fissile material in potentially
extractable form [21–23]. Such materials, if declared to be waste, are likely to
require disposal in a geological disposal facility for reasons of long term safety.
Placement in a geological disposal facility would also provide long term passive
nuclear security and would be consistent with the objective of IAEA nuclear
safeguards. Requirement 23, therefore, applies in particular to geological disposal
facilities.
10
5.16. State systems of accounting for, and control of, nuclear material were
developed primarily to provide for accountability for nuclear material, in order to
detect its possible diversion for unauthorized or unknown purposes in the short
and medium terms. As organized at present, IAEA nuclear safeguards activities
depend on active surveillance and controls.
5.17. During the operation of a disposal facility for waste that includes fissile
material, surveillance for the purposes of IAEA safeguards is aimed at ensuring
9
State systems of accounting for, and control of, nuclear material are required by IAEA
nuclear safeguards agreements.
10
INTERNATIONAL ATOMIC ENERGY AGENCY, Issues in Radioactive Waste
Disposal, IAEA-TECDOC-909, IAEA, Vienna (1996).
44
the continuity of knowledge concerning the fissile material and the absence of
any undeclared activities at the site in relation to such material. For some
radioactive waste, such as spent nuclear fuel, certain requirements for safeguards
have to apply even after the waste has been sealed in a geological disposal
facility.
11
5.18. For a closed geological disposal facility, IAEA nuclear safeguards might, in
practice, be applied by remote means (e.g. satellite monitoring, aerial
photography, microseismic surveillance and administrative arrangements).
Intrusive methods, which might compromise safety after closure, have to be
avoided.
5.19. Since IAEA nuclear safeguards are internationally supervised, their
continuation might increase confidence in the longevity of administrative
controls and this would also help to prevent inadvertent disturbance of the
geological disposal facility. The continuation of safeguards inspections and
monitoring after closure of a geological disposal facility may, thus, be beneficial
to augmenting confidence in safety after closure. A discussion of interface issues
between the system of accounting for, and control of, nuclear material (and IAEA
nuclear safeguards) and radioactive waste management is included in
IAEA-TECDOC-909
10
.
Requirement 24: Requirements in respect of nuclear security measures
Measures shall be implemented to ensure an integrated approach to safety
measures and nuclear security measures in the disposal of radioactive waste.
5.20. Where nuclear security measures are necessary to prevent unauthorized
access by individuals and to prevent the unauthorized removal of radioactive
material, safety measures and nuclear security measures have to be implemented
in an integrated approach [1, 13].
5.21. The level of nuclear security has to be commensurate with the level of
radiological hazard and the nature of the waste [1, 13, 24, 25].
11
INTERNATIONAL ATOMIC ENERGY AGENCY, Advisory Group Meeting on
Safeguards Related to Final Disposal of Nuclear Material in Waste and Spent Fuel (AGM-660),
Rep. STR-243 (Revised), IAEA, Vienna (1988).
45
Requirement 25: Management systems
Management systems
12
to provide for the assurance of quality shall be
applied to all safety related activities, systems and components throughout
all the steps of the development and operation of a disposal facility. The level
of assurance for each element shall be commensurate with its importance to
safety.
5.22. An appropriate management system that integrates quality assurance
programmes will contribute to confidence that the relevant requirements and
criteria for site selection and evaluation, design, construction, operation, closure
and safety after closure are met. The relevant activities, systems and components
have to be identified on the basis of the results of systematic safety assessment.
The level of attention assigned to each aspect has to be commensurate with its
importance to safety. The management system is required to comply with the
relevant IAEA safety standards on management systems [13, 14].
5.23. The management system specifies the role of management and the
organizational structure to be used for implementing processes for all safety
related activities. It also specifies the responsibilities and authorities of the
various personnel and organizations involved in managing and implementing the
processes and assessing the quality of all work relating to safety.
5.24. While the host environment of a disposal facility is important to safety, it
cannot be designed or manufactured, but only characterized, and that to only a
limited extent. The elements of the management system that provide assurance of
the quality of the relevant safety related processes have to be designed with
account taken of the nature of the host environment.
5.25. The design, characterization and assessment of a disposal facility have to
include several sequential and sometimes overlapping steps with an increasing
degree of detail and accuracy. However, a degree of irreducible uncertainty that is
impossible to eliminate by any measures might always remain. The significance
of this uncertainty is assessed in the evaluation of the safety case and supporting
safety assessment.
12
The term ‘management system’ includes all the initial concepts of quality control
(controlling the quality of products) and its evolution through quality assurance (the system for
ensuring the quality of products) and quality management (the system for managing quality).
46
5.26. The management system for a disposal facility has to provide for the
preparation and retention of documentary evidence to illustrate that the necessary
quality of data has been achieved; that components have been supplied and used
in accordance with the relevant specifications; that the waste packages and
unpackaged waste comply with established requirements and criteria; and that
they have been properly emplaced in the disposal facility. The management
system also has to ensure the collation of all the information that is important to
safety and that is recorded at all steps of the development and operation of the
facility, and the preservation of that information. This information is important
for any reassessment of the facility in the future.
6. EXISTING DISPOSAL FACILITIES
6.1. Some disposal facilities that were developed and constructed and entered
into operation before these requirements were established may not meet all the
requirements. These facilities may be operational or non-operational. Some
disposal facilities may have been abandoned. These would be considered
‘existing situations’ in which the government would have to take responsibility
for the facilities. The requirements established in this Safety Requirements
publication would have to be treated as guidelines for developing intervention
objectives and planning activities if necessary in such situations.
Requirement 26: Existing disposal facilities
The safety of existing disposal facilities shall be assessed periodically until
termination of the licence. During this period, the safety shall also be
assessed when a safety significant modification is planned or in the event of
changes with regard to the conditions of the authorization. In the event that
any requirements set down in this Safety Requirements publication are not
met, measures shall be put in place to upgrade the safety of the facility,
economic and social factors being taken into account.
6.2. Periodic safety assessment for a disposal facility has to be aimed at
providing an overall assessment of the status of protection and safety at the
facility. It has to include an analysis of the operational experience acquired and
possible improvements that could be made, with account taken of the existing
situation and of whatever new technological developments or changes in
47
regulatory control there might be. Periodic safety assessments cannot replace the
activities for analysis, control and surveillance that are continuously carried out at
disposal facilities.
6.3. Disposal facilities that were not constructed to present safety standards may
not meet all the safety requirements established in this Safety Requirements
publication. In assessing the safety of such facilities, there may be indications that
safety criteria will not be met. In such circumstances, reasonably practicable
measures have to be taken to upgrade the safety of the disposal facility. Possible
options may include the removal of some or all of the waste from the facility,
making engineering improvements, or putting in place or enhancing institutional
controls. Evaluation of these options has to include broader technical, social and
political issues.
.
49
Appendix
ASSURANCE OF COMPLIANCE WITH THE SAFETY OBJECTIVE
AND CRITERIA
A.1. A well-designed, well-located and properly developed disposal facility for
radioactive waste will provide a high level of assurance that radiological impacts
in the period after closure will be low, both in absolute terms and in comparison
with the impacts expected from any other options for radioactive waste
management that are available at present.
A.2. A host geological formation and/or environment and site has to be
identified that provide favourable conditions for the isolation of the waste from
the accessible biosphere and the preservation of the engineered barriers (e.g. low
groundwater flow rates and a favourable geochemical environment over the long
term). The disposal facility has to be designed with account taken of the
characteristics afforded by the host geological formation and/or environment and
site, so as to optimize protection and safety and not to exceed the dose and/or risk
constraints. The disposal facility then has to be developed in accordance with the
assessed design so that the assumed safety characteristics of both the engineered
barriers and the natural barriers are realized.
A.3. The optimization of protection and safety for a disposal facility for
radioactive waste is a judgemental process that is applied to the decisions made in
the development of the facility’s design. Most important is that sound engineering
design and technical features are adopted and sound principles of management
are applied throughout the development, operation and closure of the disposal
facility. Given these considerations, protection and safety can then be considered
optimized, provided that:
(a) Due attention has been paid to the implications for long term safety of
various design options at each step in the development and in the operation
of the disposal facility;
(b) There is reasonable assurance that the assessed doses and/or risks arising
from the generally expected range over the natural evolution of the disposal
system do not exceed the relevant constraint, over timescales for which the
uncertainties are not so large as to prevent meaningful interpretation of the
results;
50
(c) The likelihood of events that might affect the performance of the disposal
facility in such a way as to give rise to higher doses or greater risks has been
reduced as far as reasonably possible by site selection and evaluation and/or
design.
A.4. It is recognized that calculated possible radiation doses to individuals in the
future due to a disposal facility are only estimates and that the uncertainties
associated with the estimates will increase for timescales extending farther into
the future. Nevertheless, estimates of possible doses and risks for long time
periods can be made and can be used as indicators for comparison with the safety
criteria.
A.5. In estimating doses to individuals in the future due to a disposal facility, the
assumption is made that people will be present locally, and that they will make
some use of local resources that may contain radionuclides originating from the
waste in the disposal facility. It is not possible to predict the behaviour of people
in the future with any certainty, and its representation in assessment models is
necessarily stylized
13
. The rationale and possible approaches to the modelling of
the biosphere and the estimation of doses arising from waste disposal facilities
have been considered in the IAEA BIOMASS Project [26].
A.6. The possibility exists that in the future, an activity or activities undertaken
by people could cause some type of intrusion into a disposal facility for
radioactive waste. It is not possible to say definitively what form such an
intrusion will take or what the likelihood of the intrusion event will be, owing to
the unpredictability of the behaviour of people in the future. Nevertheless, the
impact of certain generic intrusion events, such as construction work, mining or
drilling, can be evaluated as reference scenarios.
A.7. Generic intrusion events such as construction work, mining or drilling
could possibly occur, but will not necessarily occur. On this basis, an approach to
evaluating the implications for safety of such events has been proposed by the
ICRP, which makes use of the type of criteria set down in para. 2.15. An
agreement would have to be reached with the regulatory body as to when such an
approach was appropriate and exactly how the criteria would be used. Arbitrary
decisions may have to be made as to what would be considered a normal activity
that would be expected to occur and what would be considered intrusion events.
13
An arbitrary representation of behaviour is assumed, often on the basis of present
human habits.
51
A.8. In the event of inadvertent human intrusion into a disposal facility, a small
number of individuals involved in activities such as drilling into the facility or
mining could receive high radiation doses and exposures of other persons could
also arise as a result of the intrusion. The doses and risks involved for any
individuals authorized to take part in activities that deliberately disturb the
disposal facility or its waste need not be taken into consideration in this context,
as such activities would constitute planned exposure situations.
A.9. In general, the likelihood of inadvertent human intrusion into the waste will
be low as a consequence of the chosen depth for a geological disposal facility.
The likelihood will be low owing to institutional controls in the case of a near
surface disposal facility, and because of the decision to site the facility away from
known significant mineral resources or other valuable resources. The possible
doses that would be received from such an inadvertent intrusion could be high.
However, since the likelihood of inadvertent intrusion is low, the associated risk
is likely to be outweighed by the higher level of protection and safety afforded by
the disposal of waste in comparison with other strategies.
A.10. A disposal facility may be affected by a range of possible evolutions and
events. Some evolutions and events may be judged to be relatively likely to occur
over the period of assessment and some may be rather unlikely or very unlikely to
occur. With a view to optimizing protection and safety, the design process will
focus on ensuring that the disposal system provides for safety (i.e. through
compliance with dose constraints and/or risk constraints). Such provision will be
made in consideration of the expected evolution of the disposal system. Account
will also be taken of uncertainties concerning that evolution and the natural
events that are likely to occur over the period of assessment.
A.11. The achievement of a level of protection and safety such that calculated
doses are less than the dose constraint is not in itself sufficient for the acceptance
of a safety case for a disposal facility, since protection is also required to be
optimized [3]. Conversely, an indication that calculated doses could exceed the
dose constraint in some unlikely circumstances need not necessarily result in the
rejection of a safety case. Over very long timescales, radioactive decay of the
waste will reduce the hazard associated with a disposal facility. However,
uncertainties could become much larger and calculated estimates of doses might
exceed the dose constraint.
A.12. Comparison of doses with doses due to radionuclides of natural origin may
provide a useful indication of the significance of such cases. Caution needs to be
exercised in applying criteria for periods far into the future. Beyond such
52
timescales, the uncertainties associated with dose estimates become so large that
the criteria might no longer serve as a reasonable basis for decision making (see
the criteria in para. 2.15).
A.13. The evaluation of whether or not the design of a disposal facility will
provide an optimized level of protection and safety could require a judgement in
which several factors would be considered. These factors might include, for
example, the quality of the design of the facility and of the safety assessment, and
any significant qualitative or quantitative uncertainties in the calculation of
exposures in the long term.
A.14. In general, when irreducible uncertainties make the results of calculations
for safety assessment purposes less reliable, then comparisons with dose
constraints or risk constraints need to be treated with caution. For a disposal
facility, the uncertainties mean that caution is necessary in considering possible
human intrusion events and very low frequency natural events. Caution is also
necessary in considering calculated doses for timescales extending into the far
future. The robustness of the disposal system can be demonstrated, however, by
making an assessment of reference events that are typical of very low frequency
natural events.
53
REFERENCES
[1] EUROPEAN ATOMIC ENERGY COMMUNITY, FOOD AND AGRICULTURE
ORGANIZATION OF THE UNITED NATIONS, INTERNATIONAL ATOMIC
ENERGY AGENCY, INTERNATIONAL LABOUR ORGANIZATION,
INTERNATIONAL MARITIME ORGANIZATION, OECD NUCLEAR ENERGY
AGENCY, PAN AMERICAN HEALTH ORGANIZATION, UNITED NATIONS
ENVIRONMENT PROGRAMME, WORLD HEALTH ORGANIZATION,
Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, IAEA, Vienna
(2006).
[2] Joint Convention on the Safety of Spent Fuel Management and on the Safety of
Radioactive Waste Management, INFCIRC/546, IAEA, Vienna (1997).
[3] FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS,
INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL LABOUR
ORGANISATION, OECD NUCLEAR ENERGY AGENCY, PAN AMERICAN
HEALTH ORGANIZATION, WORLD HEALTH ORGANIZATION, International
Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of
Radiation Sources, Safety Series No. 115, IAEA, Vienna (1996) (under revision).
[4] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, 1990
Recommendations of the International Commission on Radiological Protection,
Publication 60, Pergamon Press, Oxford and New York (1991).
[5] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,
Radiological Protection Policy for the Disposal of Radioactive Waste, Publication 77,
Pergamon Press, Oxford and New York (1997).
[6] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Radiation
Protection Recommendations as Applied to the Disposal of Long-lived Solid
Radioactive Waste, Publication 81, Pergamon Press, Oxford and New York (1998).
[7] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, The 2007
Recommendations of the International Commission on Radiological Protection,
Publication 103, Elsevier (2007).
[8] INTERNATIONAL ATOMIC ENERGY AGENCY, Regulatory Control of Radioactive
Discharges to the Environment, IAEA Safety Standards Series No. WS-G-2.3, IAEA,
Vienna (2000).
[9] INTERNATIONAL ATOMIC ENERGY AGENCY, Application of the Concepts of
Exclusion, Exemption and Clearance, IAEA Safety Standards Series No. RS-G-1.7,
IAEA, Vienna (2004).
[10] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safety Glossary:
Terminology Used in Nuclear Safety and Radiation Protection, 2007 Edition, IAEA,
Vienna (2007).
[11] INTERNATIONAL ATOMIC ENERGY AGENCY, Storage of Radioactive Waste,
IAEA Safety Standards Series No. WS-G-6.1, IAEA, Vienna (2006).
[12] INTERNATIONAL ATOMIC ENERGY AGENCY, Classification of Radioactive
Waste, IAEA Safety Standards Series No. GSG-1, IAEA, Vienna (2009).
[13] INTERNATIONAL ATOMIC ENERGY AGENCY, The Management System for
Facilities and Activities, IAEA Safety Standards Series No. GS-R-3, IAEA, Vienna
(2006).
54
[14] INTERNATIONAL ATOMIC ENERGY AGENCY, The Management System for the
Disposal of Radioactive Waste, IAEA Safety Standards Series No. GS-G-3.4, IAEA,
Vienna (2008).
[15] INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL LABOUR
OFFICE, Occupational Radiation Protection, IAEA Safety Standards Series
No. RS-G-1.1, IAEA, Vienna (1999).
[16] FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS,
INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL LABOUR
ORGANIZATION, OECD NUCLEAR ENERGY AGENCY, PAN AMERICAN
HEALTH ORGANIZATION, UNITED NATIONS OFFICE FOR THE
CO-ORDINATION OF HUMANITARIAN AFFAIRS, WORLD HEALTH
ORGANIZATION, Preparedness and Response for a Nuclear or Radiological
Emergency, IAEA Safety Standards Series No. GS-R-2, IAEA, Vienna (2002).
[17] INTERNATIONAL ATOMIC ENERGY AGENCY, Regulations for the Safe Transport
of Radioactive Material, 2009 Edition, IAEA Safety Standards Series No. TS-R-1,
IAEA, Vienna (2009).
[18] INTERNATIONAL ATOMIC ENERGY AGENCY, Governmental, Legal and
Regulatory Framework for Safety, IAEA Safety Standards Series No. GSR Part 1,
IAEA, Vienna (2010).
[19] OECD NUCLEAR ENERGY AGENCY, Confidence in the Long Term Safety of Deep
Geological Repositories: Its Communication and Development, OECD, Paris (1999).
[20] INTERNATIONAL ATOMIC ENERGY AGENCY, Environmental and Source
Monitoring for Purposes of Radiation Protection, IAEA Safety Standards Series
No. RS-G-1.8, IAEA, Vienna (2005).
[21] The Agency’s Safeguards System, INFCIRC/66/Rev.2, IAEA, Vienna (1968).
[22] Model Protocol Additional to the Agreement(s) Between State(s) and the International
Atomic Energy Agency for the Application of Safeguards, INFCIRC/540(Corr.), IAEA,
Vienna (1997).
[23] The Structure and Content of Agreements between the Agency and States Required in
Connection with the Treaty on the Non-Proliferation of Nuclear Weapons,
INFCIRC/153(Corr.), IAEA, Vienna (1972).
[24] Code of Conduct on the Safety and Security of Radioactive Sources, IAEA, Vienna
(2004).
[25] The Physical Protection of Nuclear Material and Nuclear Facilities, INFCIRC/
225/Rev.4(Corr.), IAEA, Vienna (1999).
[26] INTERNATIONAL ATOMIC ENERGY AGENCY, “Reference Biospheres” for Solid
Radioactive Waste Disposal, IAEA-BIOMASS-6, IAEA, Vienna (2003).
55
Annex
RADIOACTIVE WASTE CLASSIFICATION
A–1. In accordance with the approach outlined in the Appendix of Ref. [A–1], six
classes of waste are derived and used as the basis for the classification scheme:
(1) Exempt waste
1
(EW): Waste that meets the criteria for clearance, exemption
or exclusion from regulatory control for radiation protection purposes as
described in Ref. [A–2].
(2) Very short lived waste (VSLW): Waste that can be stored for decay over a
limited period of up to a few years and subsequently cleared from
regulatory control according to arrangements approved by the regulatory
body for uncontrolled disposal, use or discharge. VSLW includes waste
primarily containing radionuclides with very short half-lives often used for
research and medical purposes.
(3) Very low level waste (VLLW): Waste that does not necessarily meet the
criteria of EW, but which does not need a high level of containment and
isolation and, therefore, is suitable for disposal in near surface landfill type
facilities with limited regulatory control. Such landfill type facilities may
also contain other hazardous waste. Typical waste in this class includes soil
and rubble with low levels of activity concentration. Concentrations of
longer lived radionuclides in VLLW are generally very limited.
(4) Low level waste (LLW): Waste that is above clearance levels, but with
limited amounts of long lived radionuclides. Such waste requires robust
isolation and containment for periods of up to a few hundred years and is
suitable for disposal in engineered near surface facilities. This class covers
a very broad range of waste. LLW may include short lived radionuclides at
higher levels of activity concentration and long lived radionuclides, but
only at relatively low levels of activity concentration.
(5) Intermediate level waste (ILW): Waste that, because of its content,
particularly of long lived radionuclides, requires a greater degree of
containment and isolation than that provided by near surface disposal.
However, ILW needs no provision or only limited provision for heat
dissipation during its storage and disposal. ILW may contain long lived
1
For the sake of consistency, the term ‘exempt waste’ has been retained from the
previous classification scheme detailed in INTERNATIONAL ATOMIC ENERGY AGENCY,
Classification of Radioactive Waste, Safety Series No. 111-G-1.1, IAEA, Vienna (1994).
However, once such waste has been cleared from regulatory control, it is not considered to be
radioactive waste.
56
radionuclides, in particular alpha emitting radionuclides, which will not
decay to a level of activity concentration acceptable for near surface
disposal during the time for which institutional controls can be relied upon.
Therefore, waste in this class requires disposal at greater depths, in the
order of tens of metres to a few hundred metres.
(6) High level waste (HLW): Waste with levels of activity concentration high
enough to generate significant quantities of heat by the radioactive decay
process or waste with large amounts of long lived radionuclides that need to
be considered in the design of a disposal facility for such waste. Disposal in
deep, stable geological formations usually several hundred metres or more
below the surface is the generally recognized option for disposal of HLW.
REFERENCES TO THE ANNEX
[A–1] INTERNATIONAL ATOMIC ENERGY AGENCY, Classification of Radioactive
Waste, IAEA Safety Standards Series No. GSG-1, IAEA, Vienna (2009).
[A–2] INTERNATIONAL ATOMIC ENERGY AGENCY, Application of the Concepts of
Exclusion, Exemption and Clearance, IAEA Safety Standards Series No. RS-G-1.7,
IAEA, Vienna (2004).
57
CONTRIBUTORS TO DRAFTING AND REVIEW
Abu-Eid, R. Nuclear Regulatory Commission,
United States of America
Avila, R. Facilia AB, Sweden
Bennett, D. TerraSalus Limited, United Kingdom
Bernier, F. Federal Agency for Nuclear Control, Belgium
Besnus, F. Institute for Radiological Protection and
Nuclear Safety, France
Blommaert, W. Federal Agency for Nuclear Control, Belgium
Bruno, G. European Commission
Cooper, J. Health Protection Agency, United Kingdom
Goldammer, W. Strategic Consulting, Germany
Jensen, M. Swedish Radiation Protection Authority, Sweden
Kawakami, H. Japan Nuclear Energy Safety Organization, Japan
Louvat, D. International Atomic Energy Agency
Metcalf, P. International Atomic Energy Agency
Moeller, K. Federal Office for Radiation Protection, Germany
Paltemaa, R. Radiation and Nuclear Safety Authority, Finland
Pather, T. National Nuclear Regulator, South Africa
Rana, D. Bhabha Atomic Research Centre, India
Röhlig, K. Clausthal University of Technology, Germany
Rowat, J. International Atomic Energy Agency
Serres, C. Institute for Radiological Protection and
Nuclear Safety, France
58
Siraky, G. International Atomic Energy Agency
Sugier, A. Institute for Radiological Protection and
Nuclear Safety, France
Summerling, T. Safety Assessment Management Limited,
United Kingdom
Weiss, W. Federal Office for Radiation Protection, Germany
59
BODIES FOR THE ENDORSEMENT
OF IAEA SAFETY STANDARDS
An asterisk denotes a corresponding member. Corresponding members receive
drafts for comment and other documentation but they do not generally participate
in meetings. Two asterisks denote an alternate.
Commission on Safety Standards
Argentina: González, A.J.; Australia: Loy, J.; Belgium: Samain, J.-P.; Brazil:
Vinhas, L.A.; Canada: Jammal, R.; China: Liu Hua; Egypt: Barakat, M.; Finland:
Laaksonen, J.; France: Lacoste, A.-C. (Chairperson); Germany: Majer, D.; India:
Sharma, S.K.; Israel: Levanon, I.; Japan: Fukushima, A.; Korea, Republic of:
Choul-Ho Yun; Lithuania: Maksimovas, G.; Pakistan: Rahman, M.S.; Russian
Federation: Adamchik, S.; South Africa: Magugumela, M.T.; Spain: Barceló
Vernet, J.; Sweden: Larsson, C.M.; Ukraine: Mykolaichuk, O.; United Kingdom:
Weightman, M.; United States of America: Virgilio, M.; Vietnam: Le-chi Dung;
IAEA: Delattre, D. (Coordinator); Advisory Group on Nuclear Security:
Hashmi, J.A.; European Commission: Faross, P.; International Nuclear Safety
Group: Meserve, R.; International Commission on Radiological Protection:
Holm, L.-E.; OECD Nuclear Energy Agency: Yoshimura, U.; Safety Standards
Committee Chairpersons: Brach, E.W. (TRANSSC); Magnusson, S. (RASSC);
Pather, T. (WASSC); Vaughan, G.J. (NUSSC).
Nuclear Safety Standards Committee
Algeria: Merrouche, D.; Argentina: Waldman, R.; Australia: Le Cann, G.; Austria:
Sholly, S.; Belgium: De Boeck, B.; Brazil: Gromann, A.; *Bulgaria:
Gledachev, Y.; Canada: Rzentkowski, G.; China: Jingxi Li; Croatia: Val?i?, I.;
*Cyprus: Demetriades, P.; Czech Republic: Šváb, M.; Egypt: Ibrahim, M.;
Finland: Järvinen, M.-L.; France: Feron, F.; Germany: Wassilew, C.; Ghana:
Emi-Reynolds, G.; *Greece: Camarinopoulos, L.; Hungary: Adorján, F.; India:
Vaze, K.; Indonesia: Antariksawan, A.; Iran, Islamic Republic of:
Asgharizadeh, F.; Israel: Hirshfeld, H.; Italy: Bava, G.; Japan: Kanda, T.; Korea,
Republic of: Hyun-Koon Kim; Libyan Arab Jamahiriya: Abuzid, O.; Lithuania:
Dem?enko, M.; Malaysia: Azlina Mohammed Jais; Mexico: Carrera, A.; Morocco:
Soufi, I.; Netherlands: van der Wiel, L.; Pakistan: Habib, M.A.; Poland:
Jurkowski, M.; Romania: Biro, L.; Russian Federation: Baranaev, Y.; Slovakia:
Uhrik, P.; Slovenia: Vojnovi?, D.; South Africa: Leotwane, W.; Spain:
Zarzuela, J.; Sweden: Hallman, A.; Switzerland: Flury, P.; Tunisia: Baccouche, S.;
60
Turkey: Bezdegumeli, U.; Ukraine: Shumkova, N.; United Kingdom:
Vaughan, G.J. (Chairperson); United States of America: Mayfield, M.; Uruguay:
Nader, A.; European Commission: Vigne, S.; FORATOM: Fourest, B.;
IAEA: Feige, G. (Coordinator); International Electrotechnical Commission:
Bouard, J.-P.; International Organization for Standardization: Sevestre, B.;
OECD Nuclear Energy Agency: Reig, J.; *World Nuclear Association:
Borysova, I.
Radiation Safety Standards Committee
*Algeria: Chelbani, S.; Argentina: Massera, G.; Australia: Melbourne, A.;
*Austria: Karg, V.; Belgium: van Bladel, L.; Brazil: Rodriguez Rochedo, E.R.;
*Bulgaria: Katzarska, L.; Canada: Clement, C.; China: Huating Yang; Croatia:
Kralik, I.; *Cuba: Betancourt Hernandez, L.; *Cyprus: Demetriades, P.; Czech
Republic: Petrova, K.; Denmark: Øhlenschlæger, M.; Egypt: Hassib, G.M.;
Estonia: Lust, M.; Finland: Markkanen, M.; France: Godet, J.-L.; Germany:
Helming, M.; Ghana: Amoako, J.; *Greece: Kamenopoulou, V.; Hungary:
Koblinger, L.; Iceland: Magnusson, S. (Chairperson); India: Sharma, D.N.;
Indonesia: Widodo, S.; Iran, Islamic Republic of: Kardan, M.R.; Ireland:
Colgan, T.; Israel: Koch, J.; Italy: Bologna, L.; Japan: Kiryu, Y.; Korea, Republic
of: Byung-Soo Lee; *Latvia: Salmins, A.; Libyan Arab Jamahiriya: Busitta, M.;
Lithuania: Mastauskas, A.; Malaysia: Hamrah, M.A.; Mexico: Delgado
Guardado, J.; Morocco: Tazi, S.; Netherlands: Zuur, C.; Norway: Saxebol, G.;
Pakistan: Ali, M.; Paraguay: Romero de Gonzalez, V.; Philippines: Valdezco, E.;
Poland: Merta, A.; Portugal: Dias de Oliveira, A.M.; Romania: Rodna, A.;
Russian Federation: Savkin, M.; Slovakia: Jurina, V.; Slovenia: Sutej, T.; South
Africa: Olivier, J.H.I.; Spain: Amor Calvo, I.; Sweden: Almen, A.; Switzerland:
Piller, G.; *Thailand: Suntarapai, P.; Tunisia: Chékir, Z.; Turkey: Okyar, H.B.;
Ukraine: Pavlenko, T.; United Kingdom: Robinson, I.; United States of America:
Lewis, R.; *Uruguay: Nader, A.; European Commission: Janssens, A.; Food and
Agriculture Organization of the United Nations: Byron, D.; IAEA: Boal, T.
(Coordinator); International Commission on Radiological Protection: Valentin, J.;
International Electrotechnical Commission: Thompson, I.; International Labour
Office: Niu, S.; International Organization for Standardization: Rannou, A.;
International Source Suppliers and Producers Association: Fasten, W.; OECD
Nuclear Energy Agency: Lazo, T.E.; Pan American Health Organization:
Jiménez, P.; United Nations Scientific Committee on the Effects of Atomic
Radiation: Crick, M.; World Health Organization: Carr, Z.; World Nuclear
Association: Saint-Pierre, S.
61
Transport Safety Standards Committee
Argentina: López Vietri, J.; **Capadona, N.M.; Australia: Sarkar, S.; Austria:
Kirchnawy, F.; Belgium: Cottens, E.; Brazil: Xavier, A.M.; Bulgaria:
Bakalova, A.; Canada: Régimbald, A.; China: Xiaoqing Li; Croatia:
Belamari?, N.; *Cuba: Quevedo Garcia, J.R.; *Cyprus: Demetriades, P.; Czech
Republic: Duchá?ek, V.; Denmark: Breddam, K.; Egypt: El-Shinawy, R.M.K.;
Finland: Lahkola, A.; France: Landier, D.; Germany: Rein, H.; *Nitsche, F.;
**Alter, U.; Ghana: Emi-Reynolds, G.; *Greece: Vogiatzi, S.; Hungary: Sáfár, J.;
India: Agarwal, S.P.; Indonesia: Wisnubroto, D.; Iran, Islamic Republic of:
Eshraghi, A.; *Emamjomeh, A.; Ireland: Duffy, J.; Israel: Koch, J.; Italy:
Trivelloni, S.; **Orsini, A.; Japan: Hanaki, I.; Korea, Republic of: Dae-Hyung
Cho; Libyan Arab Jamahiriya: Kekli, A.T.; Lithuania: Statkus, V.; Malaysia:
Sobari, M.P.M.; **Husain, Z.A.; Mexico: Bautista Arteaga, D.M.; **Delgado
Guardado, J.L.; *Morocco: Allach, A.; Netherlands: Ter Morshuizen, M.; *New
Zealand: Ardouin, C.; Norway: Hornkjøl, S.; Pakistan: Rashid, M.; *Paraguay:
More Torres, L.E.; Poland: Dziubiak, T.; Portugal: Buxo da Trindade, R.; Russian
Federation: Buchelnikov, A.E.; South Africa: Hinrichsen, P.; Spain: Zamora
Martin, F.; Sweden: Häggblom, E.; **Svahn, B.; Switzerland: Krietsch, T.;
Thailand: Jerachanchai, S.; Turkey: Ertürk, K.; Ukraine: Lopatin, S.; United
Kingdom: Sallit, G.; United States of America: Boyle, R.W.; Brach, E.W.
(Chairperson); Uruguay: Nader, A.; *Cabral, W.; European Commission: Binet, J.;
IAEA: Stewart, J.T. (Coordinator); International Air Transport Association:
Brennan, D.; International Civil Aviation Organization: Rooney, K.; International
Federation of Air Line Pilots’ Associations: Tisdall, A.; **Gessl, M.; International
Maritime Organization: Rahim, I.; International Organization for
Standardization: Malesys, P.; International Source Supplies and Producers
Association: Miller, J.J.; **Roughan, K.; United Nations Economic Commission
for Europe: Kervella, O.; Universal Postal Union: Bowers, D.G.; World Nuclear
Association: Gorlin, S.; World Nuclear Transport Institute: Green, L.
Waste Safety Standards Committee
Algeria: Abdenacer, G.; Argentina: Biaggio, A.; Australia: Williams, G.; *Austria:
Fischer, H.; Belgium: Blommaert, W.; Brazil: Tostes, M.; *Bulgaria:
Simeonov, G.; Canada: Howard, D.; China: Zhimin Qu; Croatia: Trifunovic, D.;
Cuba: Fernandez, A.; Cyprus: Demetriades, P.; Czech Republic: Lietava, P.;
Denmark: Nielsen, C.; Egypt: Mohamed, Y.; Estonia: Lust, M.; Finland: Hutri, K.;
France: Rieu, J.; Germany: Götz, C.; Ghana: Faanu, A.; Greece: Tzika, F.;
Hungary: Czoch, I.; India: Rana, D.; Indonesia: Wisnubroto, D.; Iran, Islamic
62
Republic of: Assadi, M.; *Zarghami, R.; Iraq: Abbas, H.; Israel: Dody, A.; Italy:
Dionisi, M.; Japan: Matsuo, H.; Korea, Republic of: Won-Jae Park; *Latvia:
Salmins, A.; Libyan Arab Jamahiriya: Elfawares, A.; Lithuania: Paulikas, V.;
Malaysia: Sudin, M.; Mexico: Aguirre Gómez, J.; *Morocco: Barkouch, R.;
Netherlands: van der Shaaf, M.; Pakistan: Mannan, A.; *Paraguay: Idoyaga
Navarro, M.; Poland: Wlodarski, J.; Portugal: Flausino de Paiva, M.; Slovakia:
Homola, J.; Slovenia: Mele, I.; South Africa: Pather, T. (Chairperson); Spain: Sanz
Aludan, M.; Sweden: Frise, L.; Switzerland: Wanner, H.; *Thailand: Supaokit, P.;
Tunisia: Bousselmi, M.; Turkey: Özdemir, T.; Ukraine: Makarovska, O.; United
Kingdom: Chandler, S.; United States of America: Camper, L.; *Uruguay:
Nader, A.; European Commission: Necheva, C.; European Nuclear Installations
Safety Standards: Lorenz, B.; *European Nuclear Installations Safety Standards:
Zaiss, W.; IAEA: Siraky, G. (Coordinator); International Organization for
Standardization: Hutson, G.; International Source Suppliers and Producers
Association: Fasten, W.; OECD Nuclear Energy Agency: Riotte, H.; World
Nuclear Association: Saint-Pierre, S.
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Under the terms of Article III of its Statute, the IAEA is authorized to establish or adopt standards of safety for protection of health and minimization of danger to life and property, and to provide for the application of these standards.
INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA
ISBN 978–92 –0–103010–8
ISSN 1020–525X
“Governments, regulatory bodies and operators everywhere must
ensure that nuclear material and radiation sources are used
beneficially, safely and ethically. The IAEA safety standards are
designed to facilitate this, and I encourage all Member States to
make use of them.”
Yukiya Amano
Director General
Safety through international standards
IAEA Safety Standards
Disposal of
Radioactive Waste
for protecting people and the environment
No. SSR-5
Specific Safety Requirements
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P1449_cover.indd 1 2011-05-05 08:39:50
IAEA SAFETY RELATED PUBLICATIONS
IAEA SAFETY STANDARDS
Under the terms of Article III of its Statute, the IAEA is authorized to establish
or adopt standards of safety for protection of health and minimization of danger to life
and property, and to provide for the application of these standards.
The publications by means of which the IAEA establishes standards are issued in
the IAEA Safety Standards Series. This series covers nuclear safety, radiation safety,
transport safety and waste safety. The publication categories in the series are Safety
Fundamentals, Safety Requirements and Safety Guides.
Information on the IAEA’s safety standards programme is available at the IAEA
Internet sitehttp://www-ns.iaea.org/standards/
The site provides the texts in English of published and draft safety standards. The
texts of safety standards issued in Arabic, Chinese, French, Russian and Spanish, the
IAEA Safety Glossary and a status report for safety standards under development are
also available. For further information, please contact the IAEA at PO Box 100,
1400 Vienna, Austria.
All users of IAEA safety standards are invited to inform the IAEA of experience
in their use (e.g. as a basis for national regulations, for safety reviews and for training
courses) for the purpose of ensuring that they continue to meet users’ needs.
Information may be provided via the IAEA Internet site or by post, as above, or by
email to [email protected].
OTHER SAFETY RELATED PUBLICATIONS
The IAEA provides for the application of the standards and, under the terms of
Articles III and VIII.C of its Statute, makes available and fosters the exchange of
information relating to peaceful nuclear activities and serves as an intermediary among
its Member States for this purpose.
Reports on safety and protection in nuclear activities are issued as Safety
Reports, which provide practical examples and detailed methods that can be used in
support of the safety standards.
Other safety related IAEA publications are issued as Radiological Assessment
Reports, the International Nuclear Safety Group’s INSAG Reports, Technical Reports
and TECDOCs. The IAEA also issues reports on radiological accidents, training
manuals and practical manuals, and other special safety related publications. Security
related publications are issued in the IAEA Nuclear Security Series.
P1449_cover.indd 2 2011-05-05 08:39:50
DISPOSAL OF RADIOACTIVE WASTE
The following States are Members of the International Atomic Energy Agency:
The Agency’s Statute was approved on 23 October 1956 by the Conference on the Statute of the
IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1957. The
Headquarters of the Agency are situated in Vienna. Its principal objective is “to accelerate and enlarge the
contribution of atomic energy to peace, health and prosperity throughout the world’’.
AFGHANISTAN
ALBANIA
ALGERIA
ANGOLA
ARGENTINA
ARMENIA
AUSTRALIA
AUSTRIA
AZERBAIJAN
BAHRAIN
BANGLADESH
BELARUS
BELGIUM
BELIZE
BENIN
BOLIVIA
BOSNIA AND HERZEGOVINA
BOTSWANA
BRAZIL
BULGARIA
BURKINA FASO
BURUNDI
CAMBODIA
CAMEROON
CANADA
CENTRAL AFRICAN
REPUBLIC
CHAD
CHILE
CHINA
COLOMBIA
CONGO
COSTA RICA
CÔTE D’IVOIRE
CROATIA
CUBA
CYPRUS
CZECH REPUBLIC
DEMOCRATIC REPUBLIC
OF THE CONGO
DENMARK
DOMINICAN REPUBLIC
ECUADOR
EGYPT
EL SALVADOR
ERITREA
ESTONIA
ETHIOPIA
FINLAND
FRANCE
GABON
GEORGIA
GERMANY
GHANA
GREECE
GUATEMALA
HAITI
HOLY SEE
HONDURAS
HUNGARY
ICELAND
INDIA
INDONESIA
IRAN, ISLAMIC REPUBLIC OF
IRAQ
IRELAND
ISRAEL
ITALY
JAMAICA
JAPAN
JORDAN
KAZAKHSTAN
KENYA
KOREA, REPUBLIC OF
KUWAIT
KYRGYZSTAN
LATVIA
LEBANON
LESOTHO
LIBERIA
LIBYAN ARAB JAMAHIRIYA
LIECHTENSTEIN
LITHUANIA
LUXEMBOURG
MADAGASCAR
MALAWI
MALAYSIA
MALI
MALTA
MARSHALL ISLANDS
MAURITANIA
MAURITIUS
MEXICO
MONACO
MONGOLIA
MONTENEGRO
MOROCCO
MOZAMBIQUE
MYANMAR
NAMIBIA
NEPAL
NETHERLANDS
NEW ZEALAND
NICARAGUA
NIGER
NIGERIA
NORWAY
OMAN
PAKISTAN
PALAU
PANAMA
PARAGUAY
PERU
PHILIPPINES
POLAND
PORTUGAL
QATAR
REPUBLIC OF MOLDOVA
ROMANIA
RUSSIAN FEDERATION
SAUDI ARABIA
SENEGAL
SERBIA
SEYCHELLES
SIERRA LEONE
SINGAPORE
SLOVAKIA
SLOVENIA
SOUTH AFRICA
SPAIN
SRI LANKA
SUDAN
SWEDEN
SWITZERLAND
SYRIAN ARAB REPUBLIC
TAJIKISTAN
THAILAND
THE FORMER YUGOSLAV
REPUBLIC OF MACEDONIA
TUNISIA
TURKEY
UGANDA
UKRAINE
UNITED ARAB EMIRATES
UNITED KINGDOM OF
GREAT BRITAIN AND
NORTHERN IRELAND
UNITED REPUBLIC
OF TANZANIA
UNITED STATES OF AMERICA
URUGUAY
UZBEKISTAN
VENEZUELA
VIETNAM
YEMEN
ZAMBIA
ZIMBABWE
DISPOSAL OF RADIOACTIVE WASTE
SPECIFIC SAFETY REQUIREMENTS
This publication includes a CD-ROM containing the IAEA Safety Glossary:
2007 Edition (2007) and the Fundamental Safety Principles (2006),
each in Arabic, Chinese, English, French, Russian and Spanish versions.
The CD-ROM is also available for purchase separately.
See:http://www-pub.iaea.org/MTCD/publications/publications.asp
INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA, 2011
IAEA SAFETY STANDARDS SERIES No. SSR-5
IAEA Library Cataloguing in Publication Data
Disposal of radioactive waste. — Vienna : International Atomic Energy Agency,
2011.
p. ; 24 cm. — (IAEA safety standards series, ISSN 1020–525X ;
no. SSR-5)
STI/PUB/1449
ISBN 978–92–0–103010–8
Includes bibliographical references.
1. Radioactive waste disposal — Safety measures. — 2. Radioactive waste
disposal — Environmental aspects. — 3. Radioactive waste sites —
Management. 4. Safety standards. I. International Atomic Energy Agency.
II. Series.
IAEAL 10–00661
COPYRIGHT NOTICE
All IAEA scientific and technical publications are protected by the terms of
the Universal Copyright Convention as adopted in 1952 (Berne) and as revised in
1972 (Paris). The copyright has since been extended by the World Intellectual
Property Organization (Geneva) to include electronic and virtual intellectual
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to royalty agreements. Proposals for non-commercial reproductions and
translations are welcomed and considered on a case-by-case basis. Enquiries
should be addressed to the IAEA Publishing Section at:
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email: [email protected]http://www.iaea.org/books
© IAEA, 2011
Printed by the IAEA in Austria
April 2011
STI/PUB/1449
FOREWORD
by Yukiya Amano
Director General
The IAEA’s Statute authorizes the Agency to “establish or adopt…
standards of safety for protection of health and minimization of danger to life and
property” — standards that the IAEA must use in its own operations, and which
States can apply by means of their regulatory provisions for nuclear and radiation
safety. The IAEA does this in consultation with the competent organs of the
United Nations and with the specialized agencies concerned. A comprehensive
set of high quality standards under regular review is a key element of a stable and
sustainable global safety regime, as is the IAEA’s assistance in their application.
The IAEA commenced its safety standards programme in 1958. The
emphasis placed on quality, fitness for purpose and continuous improvement has
led to the widespread use of the IAEA standards throughout the world. The Safety
Standards Series now includes unified Fundamental Safety Principles, which
represent an international consensus on what must constitute a high level of
protection and safety. With the strong support of the Commission on Safety
Standards, the IAEA is working to promote the global acceptance and use of its
standards.
Standards are only effective if they are properly applied in practice. The
IAEA’s safety services encompass design, siting and engineering safety,
operational safety, radiation safety, safe transport of radioactive material and safe
management of radioactive waste, as well as governmental organization,
regulatory matters and safety culture in organizations. These safety services assist
Member States in the application of the standards and enable valuable experience
and insights to be shared.
Regulating safety is a national responsibility, and many States have decided
to adopt the IAEA’s standards for use in their national regulations. For parties to
the various international safety conventions, IAEA standards provide a
consistent, reliable means of ensuring the effective fulfilment of obligations
under the conventions. The standards are also applied by regulatory bodies and
operators around the world to enhance safety in nuclear power generation and in
nuclear applications in medicine, industry, agriculture and research.
Safety is not an end in itself but a prerequisite for the purpose of the
protection of people in all States and of the environment — now and in the future.
The risks associated with ionizing radiation must be assessed and controlled
without unduly limiting the contribution of nuclear energy to equitable and
sustainable development. Governments, regulatory bodies and operators
everywhere must ensure that nuclear material and radiation sources are used
beneficially, safely and ethically. The IAEA safety standards are designed to
facilitate this, and I encourage all Member States to make use of them.
DISCLAIMER
The IAEA safety standards reflect an international consensus on what
constitutes a high level of safety for protecting people and the environment from
harmful effects of ionizing radiation. The process of developing, reviewing and
establishing the IAEA standards involves the IAEA Secretariat and all Member
States, many of which are represented on the four IAEA safety standards
committees and the IAEA Commission on Safety Standards.
The IAEA standards, as a key element of the global safety regime, are kept
under regular review by the Secretariat, the safety standards committees and the
Commission on Safety Standards. The Secretariat gathers information on
experience in the application of the IAEA standards and information gained from
the follow-up of events for the purpose of ensuring that the standards continue to
meet users’ needs. The present publication reflects feedback and experience
accumulated until 2010 and it has been subject to the rigorous review process for
standards.
The accident at the Fukushima Daiichi nuclear power plant in Japan caused
by the disastrous earthquake and tsunami of 11 March 2011 and the consequences
of the emergency for people and the environment have to be fully investigated.
They are already under study in Japan, at the IAEA and elsewhere. Lessons to be
learned for nuclear safety and radiation protection and for emergency
preparedness and response will be reflected in IAEA safety standards as they are
revised and issued in the future.
THE IAEA SAFETY STANDARDS
BACKGROUND
Radioactivity is a natural phenomenon and natural sources of radiation
are features of the environment. Radiation and radioactive substances have
many beneficial applications, ranging from power generation to uses in
medicine, industry and agriculture. The radiation risks to workers and the
public and to the environment that may arise from these applications have to
be assessed and, if necessary, controlled.
Activities such as the medical uses of radiation, the operation of nuclear
installations, the production, transport and use of radioactive material, and the
management of radioactive waste must therefore be subject to standards of
safety.
Regulating safety is a national responsibility. However, radiation risks
may transcend national borders, and international cooperation serves to
promote and enhance safety globally by exchanging experience and by
improving capabilities to control hazards, to prevent accidents, to respond to
emergencies and to mitigate any harmful consequences.
States have an obligation of diligence and duty of care, and are expected
to fulfil their national and international undertakings and obligations.
International safety standards provide support for States in meeting their
obligations under general principles of international law, such as those relating
to environmental protection. International safety standards also promote and
assure confidence in safety and facilitate international commerce and trade.
A global nuclear safety regime is in place and is being continuously
improved. IAEA safety standards, which support the implementation of
binding international instruments and national safety infrastructures, are a
cornerstone of this global regime. The IAEA safety standards constitute
a useful tool for contracting parties to assess their performance under these
international conventions.
THE IAEA SAFETY STANDARDS
The status of the IAEA safety standards derives from the IAEA’s Statute,
which authorizes the IAEA to establish or adopt, in consultation and, where
appropriate, in collaboration with the competent organs of the United Nations
and with the specialized agencies concerned, standards of safety for protection
of health and minimization of danger to life and property, and to provide for
their application.
With a view to ensuring the protection of people and the environment
from harmful effects of ionizing radiation, the IAEA safety standards establish
fundamental safety principles, requirements and measures to control the
radiation exposure of people and the release of radioactive material to the
environment, to restrict the likelihood of events that might lead to a loss of
control over a nuclear reactor core, nuclear chain reaction, radioactive source
or any other source of radiation, and to mitigate the consequences of such
events if they were to occur. The standards apply to facilities and activities that
give rise to radiation risks, including nuclear installations, the use of radiation
and radioactive sources, the transport of radioactive material and the
management of radioactive waste.
Safety measures and security measures
1
have in common the aim of
protecting human life and health and the environment. Safety measures and
security measures must be designed and implemented in an integrated manner
so that security measures do not compromise safety and safety measures do not
compromise security.
The IAEA safety standards reflect an international consensus on what
constitutes a high level of safety for protecting people and the environment
from harmful effects of ionizing radiation. They are issued in the IAEA Safety
Standards Series, which has three categories (see Fig. 1).
Safety Fundamentals
Safety Fundamentals present the fundamental safety objective and
principles of protection and safety, and provide the basis for the safety
requirements.
Safety Requirements
An integrated and consistent set of Safety Requirements establishes the
requirements that must be met to ensure the protection of people and the
environment, both now and in the future. The requirements are governed by
the objective and principles of the Safety Fundamentals. If the requirements
are not met, measures must be taken to reach or restore the required level of
safety. The format and style of the requirements facilitate their use for the
establishment, in a harmonized manner, of a national regulatory framework.
Requirements, including numbered ‘overarching’ requirements, are expressed
1
See also publications issued in the IAEA Nuclear Security Series.
as ‘shall’ statements. Many requirements are not addressed to a specific party,
the implication being that the appropriate parties are responsible for fulfilling
them.
Safety Guides
Safety Guides provide recommendations and guidance on how to comply
with the safety requirements, indicating an international consensus that it is
necessary to take the measures recommended (or equivalent alternative
measures). The Safety Guides present international good practices, and
increasingly they reflect best practices, to help users striving to achieve high
levels of safety. The recommendations provided in Safety Guides are expressed
as ‘should’ statements.
APPLICATION OF THE IAEA SAFETY STANDARDS
The principal users of safety standards in IAEA Member States are
regulatory bodies and other relevant national authorities. The IAEA safety
Part 1. Governmental, Legal and
Regulatory Framework for Safety
Part 2. Leadership and Management
for Safety
Part 3. Radiation Protection and the
Safety of Radiation Sources
Part 4. Safety Assessment for
Facilities and Activities
Part 5. Predisposal Management
of Radioactive Waste
Part 6. Decommissioning and
Termination of Activities
Part 7. Emergency Preparedness
and Response
1. Site Evaluation for
Nuclear Installations
2. Safety of Nuclear Power Plants
2.1. Design and Construction
2.2. Commissioning and Operation
3. Safety of Research Reactors
4. Safety of Nuclear Fuel
Cycle Facilities
5. Safety of Radioactive Waste
Disposal Facilities
6. Safe Transport of
Radioactive Material
General Safety Requirements Specific Safety Requirements
Safety Fundamentals
Fundamental Safety Principles
Collection of Safety Guides
FIG. 1. The long term structure of the IAEA Safety Standards Series.
standards are also used by co-sponsoring organizations and by many
organizations that design, construct and operate nuclear facilities, as well as
organizations involved in the use of radiation and radioactive sources.
The IAEA safety standards are applicable, as relevant, throughout the
entire lifetime of all facilities and activities — existing and new — utilized for
peaceful purposes and to protective actions to reduce existing radiation risks.
They can be used by States as a reference for their national regulations in
respect of facilities and activities.
The IAEA’s Statute makes the safety standards binding on the IAEA in
relation to its own operations and also on States in relation to IAEA assisted
operations.
The IAEA safety standards also form the basis for the IAEA’s safety
review services, and they are used by the IAEA in support of competence
building, including the development of educational curricula and training
courses.
International conventions contain requirements similar to those in the
IAEA safety standards and make them binding on contracting parties.
The IAEA safety standards, supplemented by international conventions,
industry standards and detailed national requirements, establish a consistent
basis for protecting people and the environment. There will also be some
special aspects of safety that need to be assessed at the national level. For
example, many of the IAEA safety standards, in particular those addressing
aspects of safety in planning or design, are intended to apply primarily to new
facilities and activities. The requirements established in the IAEA safety
standards might not be fully met at some existing facilities that were built to
earlier standards. The way in which IAEA safety standards are to be applied
to such facilities is a decision for individual States.
The scientific considerations underlying the IAEA safety standards
provide an objective basis for decisions concerning safety; however, decision
makers must also make informed judgements and must determine how best to
balance the benefits of an action or an activity against the associated radiation
risks and any other detrimental impacts to which it gives rise.
DEVELOPMENT PROCESS FOR THE IAEA SAFETY STANDARDS
The preparation and review of the safety standards involves the IAEA
Secretariat and four safety standards committees, for nuclear safety (NUSSC),
radiation safety (RASSC), the safety of radioactive waste (WASSC) and the
safe transport of radioactive material (TRANSSC), and a Commission on
Safety Standards (CSS) which oversees the IAEA safety standards programme
(see Fig. 2).
All IAEA Member States may nominate experts for the safety standards
committees and may provide comments on draft standards. The membership of
the Commission on Safety Standards is appointed by the Director General and
includes senior governmental officials having responsibility for establishing
national standards.
A management system has been established for the processes of planning,
developing, reviewing, revising and establishing the IAEA safety standards.
It articulates the mandate of the IAEA, the vision for the future application of
the safety standards, policies and strategies, and corresponding functions and
responsibilities.
INTERACTION WITH OTHER INTERNATIONAL ORGANIZATIONS
The findings of the United Nations Scientific Committee on the Effects of
Atomic Radiation (UNSCEAR) and the recommendations of international
Secretariat and
consultants:
drafting of new or revision
of existing safety standard
Draft
Endorsement
by the CSS
Final draft
Review by
safety standards
committee(s)
Member States
Comments
Draft
Outline and work plan
prepared by the Secretariat;
review by the safety standards
committees and the CSS
FIG. 2. The process for developing a new safety standard or revising an existing standard.
expert bodies, notably the International Commission on Radiological
Protection (ICRP), are taken into account in developing the IAEA safety
standards. Some safety standards are developed in cooperation with other
bodies in the United Nations system or other specialized agencies, including
the Food and Agriculture Organization of the United Nations, the United
Nations Environment Programme, the International Labour Organization, the
OECD Nuclear Energy Agency, the Pan American Health Organization and
the World Health Organization.
INTERPRETATION OF THE TEXT
Safety related terms are to be understood as defined in the IAEA Safety
Glossary (seehttp://www-ns.iaea.org/standards/safety-glossary.htm). Otherwise,
words are used with the spellings and meanings assigned to them in the latest
edition of The Concise Oxford Dictionary. For Safety Guides, the English
version of the text is the authoritative version.
The background and context of each standard in the IAEA Safety
Standards Series and its objective, scope and structure are explained in
Section 1, Introduction, of each publication.
Material for which there is no appropriate place in the body text
(e.g. material that is subsidiary to or separate from the body text, is included in
support of statements in the body text, or describes methods of calculation,
procedures or limits and conditions) may be presented in appendices or
annexes.
An appendix, if included, is considered to form an integral part of the
safety standard. Material in an appendix has the same status as the body text,
and the IAEA assumes authorship of it. Annexes and footnotes to the main
text, if included, are used to provide practical examples or additional
information or explanation. Annexes and footnotes are not integral parts of the
main text. Annex material published by the IAEA is not necessarily issued
under its authorship; material under other authorship may be presented in
annexes to the safety standards. Extraneous material presented in annexes is
excerpted and adapted as necessary to be generally useful.
CONTENTS
1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Background (1.1–1.26). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
Objective (1.27–1.28). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
Scope (1.29–1.32) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
Structure (1.33) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
2. PROTECTION OF PEOPLE AND THE ENVIRONMENT . . . . . . . 10
Application of the fundamental safety principles (2.1–2.6). . . . . . . . . 10
Radiation protection in the operational period (2.7–2.14) . . . . . . . . . . 11
Radiation protection in the post-closure period (2.15–2.19) . . . . . . . . 13
Environmental and non-radiological concerns (2.20–2.24). . . . . . . . . 15
3. SAFETY REQUIREMENTS FOR PLANNING
FOR THE DISPOSAL OF RADIOACTIVE WASTE (3.1–3.5). . . . . 16
Governmental, legal and regulatory framework . . . . . . . . . . . . . . . . . 17
Requirement 1: Government responsibilities (3.6–3.7) . . . . . . . . . . 17
Requirement 2: Responsibilities of the
regulatory body (3.8–3.11). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
Requirement 3: Responsibilities of the operator (3.12–3.16) . . . . . 19
Safety approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
Requirement 4: Importance of safety in the process of
development and operation of a disposal facility (3.17–3.20). . . 20
Requirement 5: Passive means for the safety of the disposal
facility (3.21–3.25). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
Requirement 6: Understanding of a disposal facility and
confidence in safety (3.26–3.31) . . . . . . . . . . . . . . . . . . . . . . . . . 22
Design concepts for safety (3.32–3.34) . . . . . . . . . . . . . . . . . . . . . . . . 24
Requirement 7: Multiple safety functions (3.35–3.38) . . . . . . . . . . 24
Requirement 8: Containment of radioactive waste (3.39–3.42). . . . 26
Requirement 9: Isolation of radioactive waste (3.43–3.47) . . . . . . . 27
Requirement 10: Surveillance and control of passive
safety features (3.48) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
4. REQUIREMENTS FOR THE DEVELOPMENT, OPERATION
AND CLOSURE OF A DISPOSAL FACILITY (4.1) . . . . . . . . . . . . 29
Framework for disposal of radioactive waste . . . . . . . . . . . . . . . . . . . 29
Requirement 11: Step by step development and evaluation of
disposal facilities (4.2–4.5) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
The safety case and safety assessment (4.6–4.11) . . . . . . . . . . . . . . . . 30
Requirement 12: Preparation, approval and use of the safety
case and safety assessment for a disposal facility
(4.12–4.14) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
Requirement 13: Scope of the safety case and safety
assessment (4.15–4.22) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
Requirement 14: Documentation of the safety case and safety
assessment (4.23–4.25) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
Steps in the development, operation and closure of
a disposal facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
Requirement 15: Site characterization for a disposal facility
(4.26–4.29) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
Requirement 16: Design of a disposal facility (4.30–4.32) . . . . . . . 36
Requirement 17: Construction of a disposal facility
(4.33–4.34) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37
Requirement 18: Operation of a disposal facility (4.35–4.37) . . . . . 38
Requirement 19: Closure of a disposal facility (4.38–4.41). . . . . . . 38
5. ASSURANCE OF SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39
Requirement 20: Waste acceptance in a disposal facility
(5.1–5.3) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39
Requirement 21: Monitoring programmes at a disposal facility
(5.4–5.5) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40
Requirement 22: The period after closure and institutional
controls (5.6–5.14) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41
Requirement 23: Consideration of the State system of
accounting for, and control of, nuclear material (5.15–5.19) . . . 43
Requirement 24: Requirements in respect of nuclear security
measures (5.20–5.21) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44
Requirement 25: Management systems (5.22–5.26) . . . . . . . . . . . . 45
6. EXISTING DISPOSAL FACILITIES (6.1) . . . . . . . . . . . . . . . . . . . . 46
Requirement 26: Existing disposal facilities (6.2–6.3) . . . . . . . . . . 46
APPENDIX: ASSURANCE OF COMPLIANCE WITH THE
SAFETY OBJECTIVE AND CRITERIA . . . . . . . . . . . . . . . 49
REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53
ANNEX: RADIOACTIVE WASTE CLASSIFICATION . . . . . . . . . . 55
CONTRIBUTORS TO DRAFTING AND REVIEW . . . . . . . . . . . . . . . . . 57
BODIES FOR THE ENDORSEMENT OF
IAEA SAFETY STANDARDS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59
1
1. INTRODUCTION
BACKGROUND
General
1.1. Radioactive waste arises from the generation of electricity in nuclear power
plants, from nuclear fuel cycle operations and from activities in which radioactive
material is used. It also arises from activities and processes in which radioactive
material of natural origin becomes concentrated in waste material and safety
needs to be considered in its management. Radioactive waste can be generated by
a wide range of activities varying from activities in hospitals to nuclear power
plants to mines and mineral processing facilities.
1.2. The properties of radioactive waste are likewise varied, not only in terms of
radioactive content and activity concentration but also in terms of physical and
chemical properties. Its rate of generation is also varied. A common characteristic
of all radioactive waste is its potential to present a hazard to people and to the
environment, and it must, therefore, be managed so as to reduce any associated
risks to acceptable levels. The potential hazard can range from large to trivial:
a variation reflected in the management and disposal options necessary for
various types of waste.
1.3. The safety principles to be applied in all activities for radioactive waste
management are set out in the IAEA Safety Fundamentals [1]. These principles
also form the ethical and conceptual basis for the Joint Convention on the Safety
of Spent Fuel Management and on the Safety of Radioactive Waste Management
[2]. The requirements for radiation protection are set out in the International
Basic Safety Standards for Protection against Ionizing Radiation and for the
Safety of Radiation Sources (the International Basic Safety Standards) [3]. Many
of the safety requirements and concepts of protection adopted in the standards
and in the Joint Convention [2] derive from the recommendations of the
International Commission on Radiological Protection (ICRP) [4–7].
1.4. This Safety Requirements publication establishes safety requirements
relating to the disposal of radioactive waste of all types. It sets out the safety
objective and criteria for the protection of people and the environment against
radiation risks arising from disposal facilities for radioactive waste in operation
and after closure. In order to meet the criteria, measures may need to be taken in
site selection and evaluation and in the design, construction, operation and
2
closure of the disposal facility. The requirements are essential from a safety
perspective and failure to meet any of the requirements would require action to be
taken.
1.5. This Safety Requirements publication does not reiterate all the safety
requirements in respect of the governmental, legal and regulatory framework,
radiation protection and emergency planning that are established in other Safety
Requirements publications. It is based on the premise that, in general,
arrangements have to be in place to ensure that these related requirements are
met. This Safety Requirements publication does set out some requirements that
are closely related to these other thematic areas and which are of particular
importance to the safety of radioactive waste disposal facilities. Guidance on the
fulfilment of the safety requirements set out in this Safety Requirements
publication is provided in several Safety Guides specific to different types of
radioactive waste disposal facility.
1.6. The preferred strategy for the management of all radioactive waste is to
contain it (i.e. to confine the radionuclides to within the waste matrix, the
packaging and the disposal facility) and to isolate it from the accessible
biosphere. This strategy does not preclude the discharge (i.e. controlled release)
of effluents, arising from waste management activities, that contain residual
amounts of radionuclides, or the clearance of materials that meet the relevant
criteria. International safety standards have been established covering both of
these circumstances [8, 9].
1.7. Radioactive waste may arise initially in various gaseous, liquid and solid
forms. In waste management activities, the waste is generally processed to
produce stable and solid forms, and reduced in volume and immobilized, as far as
practicable, to facilitate their storage, transport and disposal. This Safety
Requirements publication is concerned with the stage of disposal of solid or
solidified materials, which is the last step in the process of radioactive waste
management.
3
Concepts relating to disposal (and storage) of radioactive waste
1.8. The term ‘disposal’ refers to the emplacement of radioactive waste into a
facility or a location with no intention of retrieving the waste
1
. Disposal options
are designed to contain the waste by means of passive engineered and natural
features and to isolate it from the accessible biosphere to the extent necessitated
by the associated hazard. The term disposal implies that retrieval is not intended;
it does not mean that retrieval is not possible.
1.9. By contrast, the term ‘storage’ refers to the retention of radioactive waste in
a facility or a location with the intention of retrieving the waste. Both options,
disposal and storage, are designed to contain waste and to isolate it from the
accessible biosphere to the extent necessary. The important difference is that
storage is a temporary measure following which some future action is planned.
This may include further conditioning or packaging of the waste and, ultimately,
its disposal. Guidance on the safe storage of radioactive waste is provided in
Ref. [11].
1.10. A number of design options for disposal facilities have been developed and
various types of disposal facility have been constructed in many States and are in
operation. These design options have different degrees of containment and
isolation capability appropriate to the radioactive waste that they will receive.
The specific aims of disposal are:
(a) To contain the waste;
(b) To isolate the waste from the accessible biosphere and to reduce
substantially the likelihood of, and all possible consequences of,
inadvertent human intrusion
2
into the waste;
(c) To inhibit, reduce and delay the migration of radionuclides at any time from
the waste to the accessible biosphere;
(d) To ensure that the amounts of radionuclides reaching the accessible
biosphere due to any migration from the disposal facility are such that
possible radiological consequences are acceptably low at all times.
1
Terminology used in this publication is defined and explained in the IAEA Safety
Glossary [10] (seehttp://www-ns.iaea.org/standards/safety-glossary.htm).
2
‘Human intrusion’ refers to human actions that affect the integrity of a disposal facility
and which could potentially give rise to radiological consequences. Only those human actions
that result in direct disturbance of the disposal facility (i.e. the waste itself, the contaminated
near field or the engineered barrier materials) are considered.
4
1.11. The balance between the importance of each of the above mentioned aims
and the extent to which and the way in which they are accomplished will vary,
depending on the characteristics of the waste and the type of disposal facility.
1.12. Disposal facilities are not expected to provide complete containment and
isolation of waste over all time; this is neither practicable nor necessitated by the
hazard associated with waste, which declines with time.
Types of disposal facility for radioactive waste
1.13. As indicated in para. 1.10, a number of design options for disposal facilities
have been developed and various types of disposal facility have been constructed
and are in operation around the world.
1.14. Within any State or region, a number of disposal facilities of different
designs may be required in order to accommodate radioactive waste of various
types. The classification of radioactive waste is discussed in an IAEA Safety
Guide [12] and the different classes of radioactive waste are presented in the
Annex. The following disposal options have been adopted in one or more States,
corresponding to recognized classes of radioactive waste:
(a) Specific landfill disposal: Disposal in a facility similar to a conventional
landfill facility for industrial refuse but which may incorporate measures to
cover the waste. Such a facility may be designated as a disposal facility for
very low level radioactive waste (VLLW) with low concentrations or
quantities of radioactive content [12]. Typical waste disposed of in a facility
of this type may include soil and rubble arising from decommissioning
activities.
(b) Near surface disposal: Disposal in a facility consisting of engineered
trenches or vaults constructed on the ground surface or up to a few tens of
metres below ground level. Such a facility may be designated as a disposal
facility for low level radioactive waste (LLW) [12].
(c) Disposal of intermediate level waste: Depending on its characteristics,
intermediate level radioactive waste (ILW) can be disposed of in different
types of facility [12]. Disposal could be by emplacement in a facility
constructed in caverns, vaults or silos at least a few tens of metres below
ground level and up to a few hundred metres below ground level. It could
include purpose built facilities and facilities developed in or from existing
mines. It could also include facilities developed by drift mining into
mountainsides or hillsides, in which case the overlying cover could be more
than 100 m deep.
5
(d) Geological disposal: Disposal in a facility constructed in tunnels, vaults or
silos in a particular geological formation (e.g. in terms of its long term
stability and its hydrogeological properties) at least a few hundred metres
below ground level. Such a facility could be designed to receive high level
radioactive waste (HLW) [12], including spent fuel if it is to be treated as
waste. However, with appropriate design, a geological disposal facility
could receive all types of radioactive waste.
(e) Borehole disposal: Disposal in a facility consisting of an array of boreholes,
or a single borehole, which may be between a few tens of metres up to a few
hundreds of metres deep. Such a borehole disposal facility is designed for
the disposal of only relatively small volumes of waste, in particular disused
sealed radioactive sources. A design option for very deep boreholes, several
kilometres deep, has been examined for the disposal of solid high level
waste and spent fuel, but this option has not been adopted for a disposal
facility by any State.
(f) Disposal of mining and mineral processing waste: Disposal usually on or
near the ground surface, but the manner and the large volumes in which the
waste arises, its physicochemical form and its content of long lived
radionuclides of natural origin distinguish it from other radioactive waste.
The waste is generally stabilized in situ and covered with various layers of
rock and soil.
1.15. This Safety Requirements publication applies to all of the above mentioned
types of disposal and disposal facilities. Comprehensive guidance on meeting the
requirements established in this Safety Requirements publication is given in
IAEA Safety Guides, in each of which a particular type of disposal, as described
above, is considered.
1.16. In accordance with the graded approach, as required in the International
Basic Safety Standards and other standards [3, 13, 14], the ability of the chosen
disposal system to provide containment of the waste and to isolate it from people
and the environment will be commensurate with the hazard potential of the waste.
The requirements set out in this Safety Requirements publication apply to all
types of disposal facility. However, the extent of provisions necessary to meet the
requirements will vary in accordance with the graded approach. This is reflected
in the Safety Guides for the different types of facility mentioned in para. 1.14.
Disposal facility life cycle
1.17. The development (i.e. site selection and evaluation, and facility design and
construction) of most types of disposal facility is likely to take place over
6
extended periods of time. The period over which disposal facilities will be
operated prior to closure will, in most cases, also extend over decades. Different
activities will be conducted in this period of development, such as site selection
and evaluation, and facility design and construction, with decisions being made to
proceed to the next set of activities or the next step in the development of the
facility.
1.18. Such a step by step approach enables: the ordered accumulation and
assessment of the necessary scientific and technical data; the evaluation of
possible sites; the development of disposal concepts; iterative studies for design
development and safety assessment with progressively improving data; technical
and regulatory reviews; public consultation and political decisions. However, the
level of study and the process will depend on the facility and on national
practices.
1.19. The step by step approach, together with the consideration of a range of
options for the design and operational management of a disposal facility, is
expected to provide flexibility for responding to new technical information and
advances in waste management and material technologies. It also enables social,
economic and political aspects of the disposal facility to be addressed, to ensure
that all reasonable measures have been taken to further prevent, inhibit or delay
releases to the environment.
1.20. This approach may include options for reversing a given step or even, for
most types of facility, for retrieving waste after its emplacement, if this were
considered to be appropriate.
1.21. The developers of disposal facilities may define a number of steps relating
to their own programme needs. In this Safety Requirements publication, however,
the step by step approach refers to the steps that are imposed by the regulatory
body and by political decision making processes.
1.22. It is convenient to identify three periods associated with the development,
operation and closure of a disposal facility: (i) the pre-operational period,
(ii) the operational period and (iii) the post-closure period. Various activities will
take place in these periods and some may be undertaken to varying degrees
throughout part or all of the lifetime of the facility:
(i) The pre-operational period includes concept definition, site evaluation
(selection, verification and confirmation), safety assessment and design
studies. It also includes the development of those aspects of the safety case
7
for safety in operation and after closure that are required in order to set the
conditions of authorization, obtain the authorization and proceed with the
construction of the disposal facility and the initial operational activities.
The monitoring and testing programmes that are needed to inform
operational management decisions are put in place.
(ii) The operational period begins when waste is first received at the facility.
From this time, radiation exposures may occur as a result of waste
management activities, and these are subject to control in accordance with
the requirements for protection and safety. Monitoring, surveillance and
testing programmes continue to inform operational management decisions
and to provide the basis for decisions concerning the closure of the facility
or parts of it. Safety assessments for the period of operation and the period
after closure and the safety case are updated as necessary to reflect actual
experience and increasing knowledge. In the operational period,
construction activities may take place at the same time as waste
emplacement in, and closure of, other parts of the facility. This period may
include activities for waste retrieval, if considered necessary, prior to
closure, activities following the completion of waste emplacement and the
final closure and sealing of the facility.
(iii) The post-closure period begins at the time when all the engineered
containment and isolation features have been put in place, operational
buildings and supporting services have been decommissioned and the
facility is in its final configuration. After its closure, the safety of the
disposal facility is provided for by means of passive features inherent in the
characteristics of the site and the facility and the characteristics of the waste
packages, together with certain institutional controls, particularly for near
surface facilities. Such institutional controls are put in place to prevent
intrusion into facilities and to confirm that the disposal system is
performing as expected by means of monitoring and surveillance.
Monitoring may also be carried out to provide public assurance. The licence
will be terminated after the period of active institutional control, when all
the necessary technical, legal and financial requirements have been
fulfilled.
1.23. This Safety Requirements publication is concerned with providing for the
protection of people and the environment against the hazards associated with
waste management activities relating to waste disposal, including hazards that
could arise in the operational period and following closure. Assurance of this
protection will be provided by the application of legal and regulatory
requirements in the pre-operational and operational periods, and in some cases in
the post-closure period.
8
1.24. The disposal system (i.e. the disposal facility and the environment in which
it is sited) is developed in a series of steps in which the scientific understanding of
the disposal system and of the design of the disposal facility is progressively
advanced. Safety assessment is an important tool for guiding site selection and
evaluation and for assisting with the design of the facility. It is also used for
evaluating the prevailing level of understanding of the disposal system and for
assessing the associated uncertainties through the various steps in the
development of the facility. The extent and complexity of such an assessment will
vary with the type of facility and will be related to the hazard potential of the
waste.
1.25. Moreover, the development of disposal facilities that incorporate provisions
in design or operation to facilitate reversibility, including retrievability, is
considered in several national programmes for waste management. In some
States, post-closure retrievability is a legal requirement and constitutes a
boundary condition on the options available, which must always satisfy the safety
requirements for disposal. No relaxation of safety standards or requirements
could be allowed on the grounds that waste retrieval may be possible or may be
facilitated by a particular provision. It would have to be ensured that any such
provision would not have an unacceptable adverse effect on safety or on the
performance of the disposal system. This subject is not extensively dealt with in
this Safety Requirements publication.
1.26. The safety case (i.e. the collection of arguments and evidence to
demonstrate the safety of a facility) for a disposal facility will be developed
together with the development of the facility. This approach provides a basis for
decisions relating to the development, operation and closure of the facility. It also
allows the identification of areas of uncertainty on which attention needs to be
focused to improve further the understanding of those aspects influencing the
safety of the disposal system.
OBJECTIVE
1.27. The objective of this Safety Requirements publication is to set out the safety
objective and criteria for the disposal of all types of radioactive waste and to
establish, on the basis of the principles established in Ref. [1], the requirements
that must be satisfied in the disposal of radioactive waste.
9
1.28. This Safety Requirements publication is intended for use by all persons
responsible for, and concerned with, radioactive waste management and making
decisions in relation to the development, operation and closure of disposal
facilities, especially those persons concerned with the related regulatory aspects.
Safety Guides provide comprehensive guidance on, and international best
practices for, meeting the requirements in respect of different types of disposal
facility.
SCOPE
1.29. This Safety Requirements publication applies to the disposal of radioactive
waste of all types by means of emplacement in designed disposal facilities,
subject to the necessary limitations and controls being placed on the disposal of
the waste and on the development, operation and closure of facilities. The
classification of radioactive waste is discussed in Ref. [12].
1.30. This Safety Requirements publication establishes requirements to provide
assurance of the radiation safety of the disposal of radioactive waste, in the
operation of a disposal facility and especially after its closure. The fundamental
safety objective is to protect people and the environment from harmful effects of
ionizing radiation. This is achieved by setting requirements on the site selection
and evaluation and design of a disposal facility, and on its construction, operation
and closure, including organizational and regulatory requirements.
1.31. Meeting these requirements forms a part of the wider process involved in
selecting and evaluating a site and developing a disposal facility. Broader
planning, financial, economic and social issues, as well as issues of conventional
safety and environmental impacts, will also be considered in this wider process.
This Safety Requirements publication does not address these broader issues, nor
does it address the transport of waste to the site or environmental impacts other
than radiological consequences.
1.32. Experience to date in selecting sites for disposal facilities has shown that
acceptance of a disposal facility by a broad range of interested parties depends on
a number of factors. The process of involving interested parties in decision
making processes for disposal facilities is increasingly seen to be of great
importance. The detailed consideration of such processes is, however, beyond the
scope of this Safety Requirements publication.
10
STRUCTURE
1.33. The background to, the concepts of, and the safety objective for disposal are
set out in Sections 1 and 2. The safety requirements for disposal facilities are set
out in Sections 3–6. These requirements comprise 26 numbered ‘shall’ statements
in bold type.
2. PROTECTION OF PEOPLE AND THE ENVIRONMENT
APPLICATION OF THE FUNDAMENTAL SAFETY PRINCIPLES
2.1. The IAEA Safety Fundamentals publication Fundamental Safety Principles
[1] sets out the fundamental safety objective and safety principles that apply for
all facilities and activities in radioactive waste management, including the
disposal of radioactive waste. As stated in Ref. [1], the fundamental safety
objective is to protect people and the environment from harmful effects of
ionizing radiation.
2.2. The strategy adopted at present to achieve this fundamental safety objective
in respect of the disposal of radioactive waste is to contain the waste and to
isolate it from the accessible biosphere, to the extent that this is necessary. The
biosphere is that part of the environment that is normally inhabited by living
organisms, and in this Safety Requirements publication the ‘accessible biosphere’
is taken generally to include those elements of the environment, including
groundwater, surface water and marine resources, that are used by people or
accessible to people. The accessible biosphere is, therefore, that part of the
environment that the objective, criteria and requirements set out in this Safety
Requirements publication are established to protect.
2.3. By applying the strategy of containment and isolation of waste, it is implicit
that if waste were to be disturbed after its disposal in a facility, then radiation
doses might be incurred.
2.4. According to Ref. [1], disposal facilities are to be developed in such a way
that people and the environment are protected both now and in the future
(Ref. [1], Principle 7). In this regard, the prime consideration is the radiological
hazard presented by radioactive waste. The ICRP developed the System of
11
Radiological Protection that applies to all facilities and activities, and this system
was adopted in the International Basic Safety Standards [3].
2.5. The ICRP has elaborated the application of the System of Radiological
Protection to the disposal of solid radioactive waste in its Publications 77 and 81
[5, 6], which it reconfirmed in Publication 103 [7]. This provides a starting point
for the safety considerations discussed here in relation to disposal facilities.
Environmental concerns and other non-radiological concerns are considered at
the end of Section 2.
2.6. The safety objective and criteria set out in this section apply regardless of
national boundaries. Transboundary issues are dealt with in the framework of
existing conventions, treaties and bilateral agreements. Particular specific
obligations apply to Contracting Parties to the Joint Convention on the Safety of
Spent Fuel Management and on the Safety of Radioactive Waste Management [2].
RADIATION PROTECTION IN THE OPERATIONAL PERIOD
2.7. The radiation safety requirements and the related safety criteria for the
operational period of a disposal facility are the same as those for any nuclear
facility or activity involving radioactive material and are established in the
International Basic Safety Standards [3]. Disposal facilities receiving waste from
nuclear fuel cycle facilities will generally be licensed nuclear facilities and have
to operate under the terms of a facility licence. Disposal facilities for small
quantities of waste (e.g. borehole facilities) may not be regarded as nuclear
facilities in some States but have to be subject to an appropriate regulatory
process and have to be licensed accordingly.
2.8. In radiation safety terms, the disposal facility is considered to be a source of
radiation that is under regulatory control in a planned exposure situation. In the
operational period, any radioactive release can be verified, exposures can be
controlled and actions can be taken if necessary. The engineering means and
practical means of achieving safety are well known, although their use in a
disposal facility involves specific considerations. The primary goal is to ensure
that radiation doses are as low as reasonably achievable and within the applicable
system of dose limitation.
2.9. The optimization of protection (that is, the process of determining measures
for protection and safety to make exposures, and the probability and magnitude of
potential exposures, “as low as reasonably achievable, economic and social
12
factors being taken into account”) is considered in the design of the disposal
facility and in the planning of all operations [3].
2.10. Relevant considerations in the optimization of measures for protection and
safety include: the separation of mining and construction activities from waste
emplacement activities; the use of remote handling equipment and shielded
equipment for waste emplacement, where necessary; the control of the working
environment so as to reduce the potential for accidents and their potential
consequences; and the minimization of the need for maintenance in supervised
areas and controlled areas. Contamination is required to be controlled and
prevented to the extent possible [3].
2.11. No releases of radionuclides, or only very minor releases (such as small
amounts of gaseous radionuclides), may be expected during the normal operation
of a radioactive waste disposal facility and hence there will not be any significant
doses to members of the public. Even in the event of an accident involving the
breach of a waste package on the site of a disposal facility, releases are unlikely to
have any radiological consequences outside the facility.
2.12. The absence of radiological consequences of any significance outside the
facility would be confirmed by means of safety assessment (see the requirements
concerning the safety case and safety assessment, Requirements 12–14).
Relevant considerations include the waste form (i.e. the packaging and the
radionuclide content of the waste), the control of contamination on waste
packages and equipment, and the monitoring and control of drainage water from
the disposal facility, where applicable, and of the ventilation exhaust air from
underground disposal facilities.
2.13. For a disposal facility, as for any other operational nuclear facility or
facility where radioactive material is handled, used, stored or processed, an
operational radiation protection programme, commensurate with the radiological
hazards, is required to be put in place to ensure that doses to workers during
normal operations are controlled and that the requirements for the limitation of
radiation doses are met (see Ref. [3], paras 2.24–2.26, and Ref. [15]). In addition,
emergency plans are required to be put in place for dealing with accidents and
other incidents, and for ensuring that any consequent radiation doses are
controlled to the extent possible, with due regard for the relevant emergency
action levels [16].
2.14. The doses and risks associated with the transport of radioactive waste
through public areas to a disposal facility are required to be managed in the same
13
way as the doses and risks associated with the transport of other radioactive
material. The transport of radioactive waste is subject to the requirements of the
IAEA’s Regulations for the Safe Transport of Radioactive Material [17].
RADIATION PROTECTION IN THE POST-CLOSURE PERIOD
2.15. The safety objective and criteria for the protection of people and the
environment after closure of a disposal facility are as follows:
Safety objective
The safety objective is to site, design, construct, operate and close a disposal
facility so that protection after its closure is optimized, social and economic
factors being taken into account. A reasonable assurance also has to be provided
that doses and risks to members of the public in the long term will not exceed the
dose constraints or risk constraints that were used as design criteria.
Criteria
(a) The dose limit for members of the public for doses from all planned
exposure situations is an effective dose of 1 mSv in a year [3]. This and its
risk equivalent are considered criteria that are not to be exceeded in the
future.
(b) To comply with this dose limit, a disposal facility (considered as a single
source) is so designed that the calculated dose or risk to the representative
person who might be exposed in the future as a result of possible natural
processes
3
affecting the disposal facility does not exceed a dose constraint
of 0.3 mSv in a year or a risk constraint of the order of 10
–5
per year
4
.
(c) In relation to the effects of inadvertent human intrusion after closure, if
such intrusion is expected to lead to an annual dose of less than 1 mSv to
those living around the site, then efforts to reduce the probability of
intrusion or to limit its consequences are not warranted.
3
Natural processes include the range of conditions anticipated over the lifetime of the
facility and events that could occur with a lesser likelihood. However, extremely low
probability events would be outside the scope of consideration.
4
Risk due to the disposal facility in this context is to be understood as the probability of
fatal cancer or serious hereditary effects.
14
(d) If human intrusion were expected to lead to a possible annual dose of more
than 20 mSv (see Ref. [7], Table 8) to those living around the site, then
alternative options for waste disposal are to be considered, for example,
disposal of the waste below the surface, or separation of the radionuclide
content giving rise to the higher dose.
(e) If annual doses in the range 1–20 mSv (see Ref. [7], Table 8) are indicated,
then reasonable efforts are warranted at the stage of development of the
facility to reduce the probability of intrusion or to limit its consequences by
means of optimization of the facility’s design.
(f) Similar considerations apply where the relevant thresholds for deterministic
effects in organs may be exceeded.
2.16. It is recognized that radiation doses to people in the future can only be
estimated and that uncertainties associated with these estimates will increase for
periods farther into the future. Caution needs to be exercised in applying criteria
for periods far into the future. Beyond such timescales, the uncertainties
associated with dose estimates become so large that the criteria might no longer
serve as a reasonable basis for decision making.
2.17. The primary goal of the disposal of radioactive waste is the protection of
people and the environment in the long term, after the disposal facility has been
closed. In this period, migration of radionuclides to the accessible biosphere,
dispersion of radionuclides into the accessible biosphere and the consequent
exposure of people may occur. This is a consequence of the slow degradation of
engineered components and the slow transport of radionuclides from the facility
by natural processes. Discrete events may lead to an earlier or greater release.
Such events could be of either natural or human origin.
2.18. Optimization under constraints is the central approach adopted to ensure the
safety of a waste disposal facility [6]. In this context, the optimization of
protection is a judgemental process, social and economic factors being taken into
account. The optimization is conducted in a structured but essentially qualitative
manner, supported by quantitative analysis.
2.19. Different methods may be used to assess the impacts of the disposal of
radioactive waste after closure of the disposal facility and to demonstrate
compliance with national regulations expressed as constraints in terms of levels
15
of dose and/or risk. This matter is addressed in the Safety Guide on the safety
case and safety assessment for disposal
5
.
ENVIRONMENTAL AND NON-RADIOLOGICAL CONCERNS
2.20. The assessment of conventional environmental impacts such as may occur
in the construction and operational periods of a disposal facility, for example,
impacts relating to traffic, noise, visual amenity, disturbance of natural habitats,
restrictions on land use and social and economic factors, is outside the scope of
this Safety Requirements publication. This Safety Requirements publication
covers the protection of the environment against radiological hazards associated
with the radioactive material in the disposal facility. The non-radiological toxic
hazard also has to be assessed where this is significant, as discussed in the
following paragraphs.
2.21. For the purposes of the current recommendations of the ICRP [4] and the
requirements of the International Basic Safety Standards [3], it is assumed that,
subject to the appropriate definition of exposed groups, the protection of people
against the radiological hazards associated with a disposal facility will also apply
the principle of protecting the environment [4–7]. The issue of the protection of
the environment from harmful effects of ionizing radiation and the development
of standards for this purpose are under discussion internationally [7].
2.22. Estimates of possible doses and/or risks due to the future migration of
radionuclides from a disposal facility are indicators of the protection of people.
On the basis of the assumption mentioned in para. 2.21, calculations to estimate
doses in which account is taken of a range of possible environmental transfer
pathways could already be considered to be indicators of environmental
protection.
2.23. Additional indicators and comparisons, such as estimates of concentrations
and fluxes of contaminants and their comparison with concentrations and fluxes
of radionuclides of natural origin within the geosphere or biosphere, may also
prove valuable in indicating a level of overall environmental protection that is
5
A Safety Guide on the Safety Case and Safety Assessment for Disposal of Radioactive
Waste is in preparation.
16
independent of assumptions about the habits
6
of people. Other factors to be
considered may include the need for protection of groundwater resources and the
ecological sensitivity of the environment into which contaminants might be
released.
2.24. The impact of non-radioactive material present in a disposal facility has to
be assessed in accordance with national or other specific regulations and this may
be significant in some cases, for example, for some mining wastes and mixtures
of radioactive and toxic wastes. If non-radioactive material may affect the release
and migration of radioactive contaminants from the radioactive waste, then such
interactions have to be considered in the safety assessment.
3. SAFETY REQUIREMENTS FOR PLANNING
FOR THE DISPOSAL OF RADIOACTIVE WASTE
3.1. Requirements are established for ensuring that the safety objective and
criteria for disposal facilities set out in Section 2 are fulfilled. The prime
responsibility for safety rests with the operator [1], to whom the majority of the
requirements apply. However, the assurance of safety and the development of a
broader confidence in safety also require a competent regulatory process within a
specified legal and regulatory framework and the allocation of responsibilities for
pre-operational activities.
3.2. The operator
7
might be a single organization or one of a number of
organizations involved, depending on the approach taken in the State. The safety
requirements for the planning of a disposal facility apply to those elements that
have to be in place prior to the development of the disposal facility, with the
purpose of ensuring safety in the operational period and after closure.
6
INTERNATIONAL ATOMIC ENERGY AGENCY, Safety Indicators in Different
Time Frames for the Safety Assessment of Underground Radioactive Waste Repositories,
IAEA-TECDOC-767, IAEA, Vienna (1994).
7
In the IAEA safety standards, ‘operator’ means any organization or person applying
for authorization or authorized and/or responsible for nuclear, radiation, radioactive waste or
transport safety when undertaking activities or in relation to any nuclear facilities or sources of
ionizing radiation. This includes, inter alia, private individuals, governmental bodies,
consignors or carriers, licensees, hospitals, self-employed persons, etc.
17
3.3. Safety in the operation of radioactive waste disposal facilities has to be
achieved by means of a variety of engineered and operational controls similar to
those used in other facilities in which radioactive material is handled, used, stored
or processed. These include the containment and shielding for the radioactive
waste and operational control over time of exposure and proximity to the waste.
Protection of the public is provided for by preventing or controlling releases from
the facility and by controlling access to the site. Operational monitoring
programmes provide assurance of these various controls.
3.4. Safety after closure is achieved by developing a disposal system in which
the various components work together to provide and to ensure the required level
of safety. This approach offers flexibility to the designer of a disposal facility to
adapt the facility’s layout and engineered barriers so as to take advantage of the
natural characteristics of the site and the barrier potential of the host geology, if
applicable. Assurance of confidence in safety is also necessary and this may
require the consideration of a number of complex issues, including the potential
impact of operations on the performance of the disposal facility after closure.
3.5. The requirements in respect of the planning of disposal facilities for
radioactive waste are set out under three headings for the governmental, legal and
regulatory framework, the safety approach and the design concepts for safety.
GOVERNMENTAL, LEGAL AND REGULATORY FRAMEWORK
Requirement 1: Government responsibilities
The government is required to establish and maintain an appropriate
governmental, legal and regulatory framework for safety within which
responsibilities shall be clearly allocated for disposal facilities for radioactive
waste to be sited, designed, constructed, operated and closed. This shall
include: confirmation at a national level of the need for disposal facilities of
different types; specification of the steps in development and licensing of
facilities of different types; and clear allocation of responsibilities, securing
of financial and other resources, and provision of independent regulatory
functions relating to a planned disposal facility.
3.6. This requirement derives from a principle established in the Fundamental
Safety Principles (Ref. [1], Principle 2). It is also stipulated under the terms of the
Joint Convention [2]. Requirements for establishing a national system for
radioactive waste management are established in Ref. [18]. A project for the
18
disposal of radioactive waste, especially for the development of a facility for the
disposal of high level and long lived radioactive waste, has to be given special
consideration within this infrastructure because of the relatively long period of
time necessary for the development of such facilities.
3.7. Matters that have to be considered include:
(a) Defining the national policy for the long term management of radioactive
waste of different types;
(b) Setting clearly defined legal, technical and financial responsibilities for
organizations that are to be involved in the development of facilities for
radioactive waste management, including disposal facilities of all types;
(c) Ensuring the adequacy and security of financial provisions for each
disposal facility;
(d) Defining the overall process for the development, operation and closure of
disposal facilities, including the legal and regulatory requirements (e.g.
licence conditions) at each step, and the processes for decision making and
the involvement of interested parties;
(e) Ensuring that the necessary scientific and technical expertise remains
available both to the operator and for the support of independent regulatory
reviews and other national review functions;
(f) Defining legal, technical and financial responsibilities and, if necessary,
providing for any institutional arrangements that are envisaged after
closure, including monitoring and ensuring the nuclear security of different
types of waste that have been disposed of.
Requirement 2: Responsibilities of the regulatory body
The regulatory body shall establish regulatory requirements for the
development of different types of disposal facility for radioactive waste and
shall set out the procedures for meeting the requirements for the various
stages of the licensing process. It shall also set conditions for the development,
operation and closure of each individual disposal facility and shall carry out
such activities as are necessary to ensure that the conditions are met.
3.8. General standards for the protection of people and the environment are
usually set out in national policy or in legislation. The regulatory body has to
develop regulatory requirements specific to each type of disposal facility for
radioactive waste, including each type that is envisaged, on the basis of national
policy and with due regard to the safety objective and criteria set out in para. 2.15.
The regulatory body has to provide guidance on the interpretation of the national
19
legislation and regulatory requirements, as necessary, and guidance on what is
expected of the operator in respect of each individual disposal facility.
3.9. The regulatory body has to engage in dialogue with waste producers, the
operators of the disposal facility and interested parties to ensure that the
regulatory requirements are appropriate and practicable. It also has to maintain
competent staff, to acquire capabilities for independent assessment and to
undertake international cooperation, as necessary, to fulfil its regulatory
functions.
3.10. The regulatory body has to document the procedures that it uses to evaluate
the safety of each type of disposal facility, the procedures that operators are
expected to follow in the context of licensing, important decisions prior to
licensing and licence applications. It also has to document the procedures that it
follows in reviewing submissions from operators to assess compliance with
regulatory requirements.
3.11. Similarly, in respect of each individual disposal facility, the regulatory body
has to set out the procedures that an operator is expected to follow in
demonstrating compliance with the conditions for the development and operation
of the facility. The regulatory body also has to set out the procedures that it
follows to assess compliance with the conditions throughout all stages of the
development, operation and closure of the facility.
Requirement 3: Responsibilities of the operator
The operator of a disposal facility for radioactive waste shall be responsible
for its safety. The operator shall carry out safety assessment and develop and
maintain a safety case, and shall carry out all the necessary activities for site
selection and evaluation, design, construction, operation, closure and, if
necessary, surveillance after closure, in accordance with national strategy, in
compliance with the regulatory requirements and within the legal and
regulatory infrastructure.
3.12. The operator has to be responsible for developing a disposal facility that is
practicable and safe and for demonstrating its safety, consistent with the
requirements of the regulatory body. This task has to be undertaken in
consideration of: the characteristics and quantities of the radioactive waste to be
disposed of; the site or sites available; the mining, excavation, construction and
engineering techniques available; and the legal and regulatory infrastructure and
regulatory requirements. The operator also has to be responsible for developing a
20
safety case, on the basis of which decisions on the development, operation and
closure of the disposal facility have to be made (see Requirements 17–19).
3.13. The operator has to conduct or commission the research and development
work necessary to ensure that the planned technical operations can be practically
and safely accomplished, and to demonstrate this. The operator likewise has to
conduct or commission the research work necessary to investigate, to understand
and to support the understanding of the processes on which the safety of the
disposal facility depends. The operator also has to carry out all the necessary
investigations of sites and of materials and has to assess their suitability and
obtain all the data necessary for the purposes of safety assessment.
3.14. The operator has to establish technical specifications that are justified by
safety assessment, to ensure that the disposal facility is developed in accordance
with the safety case. This has to include waste acceptance criteria (see
Requirement 20) and other controls and limits to be applied during construction,
operation and closure.
3.15. The operator has to retain all the information relevant to the safety case and
the supporting safety assessment for the disposal facility and has to retain the
inspection records that demonstrate compliance with regulatory requirements and
with the operator’s own specification. Such information and records have to be
retained, at least up until the time when the information is shown to be
superseded, or until responsibility for the disposal facility is passed on to another
organization. This occurs, for example, at closure of the facility, when all relevant
information and records have to be transferred to the organization assuming
responsibility for the facility and its safety.
3.16. The operator has to cooperate with the regulatory body and has to supply all
the information that the regulatory body may request. The need to preserve the
records for long periods of time has to be taken into account in selecting the
format and media to be used for records.
SAFETY APPROACH
Requirement 4: Importance of safety in the process of development and
operation of a disposal facility
Throughout the process of development and operation of a disposal facility
for radioactive waste, an understanding of the relevance and the
21
implications for safety of the available options for the facility shall be
developed by the operator. This is for the purpose of providing an optimized
level of safety in the operational stage and after closure.
3.17. Disposal facilities for radioactive waste may be developed and operated
over a period of several years or several decades. Key decisions, such as
decisions on site selection and evaluation, and on the design, construction,
operation and closure of the disposal facility, are expected to be made as the
project develops. In this process, decisions are made on the basis of the
information available at the time, which may be either quantitative or qualitative,
and the confidence that can be placed in that information.
3.18. Decisions on the development, operation and closure of the facility are
constrained by external factors, which include: national policy and preferences,
the capacity and capability of existing storage and disposal facilities to
accommodate waste, and the availability of suitable sites and geological
formations to host planned new disposal facilities. An adequate level of
confidence in the safety of each disposal facility has to be developed before
decisions are taken.
3.19. At each major decision point, the implications for the safety of the available
design options and operational options for the disposal facility have to be
considered and taken into account. Ensuring safety, both in the operational stage
and after closure, is the overriding concern at each decision point. If more than
one option is capable of providing the required level of safety, then other factors
also have to be considered. These factors could include public acceptability, cost,
site ownership, existing infrastructure and transport routes.
3.20. Consideration has to be given to locating the facility away from significant
known mineral resources, geothermal water and other valuable subsurface
resources. This is to reduce the risk of human intrusion into the site and to reduce
the potential for use of the surrounding area to be in conflict with the facility. The
safety of the facility has to be considered at every step in the decision making
process to ensure that safety is optimized in the sense discussed in the Appendix.
Requirement 5: Passive means for the safety of the disposal facility
The operator shall evaluate the site and shall design, construct, operate and
close the disposal facility in such a way that safety is ensured by passive
means to the fullest extent possible and the need for actions to be taken after
closure of the facility is minimized.
22
3.21. In the operational stage of a disposal facility for radioactive waste, certain
active control measures have to be applied. However, where passive features such
as the shielding and containment provided by the packaging material can provide
safety, then safety has to be ensured by such passive means.
3.22. To some extent, the safety of a disposal facility can depend on some future
actions such as maintenance work or surveillance. However, this dependence has
to be minimized to the extent possible. This is necessary because of the
possibility that safety measures that depend on future actions, such as
maintenance work or surveillance, will not be taken or will not be continued. The
cumulative probability of the failure of such safety measures will gradually
increase. Furthermore, and consistent with the Fundamental Safety Principles [1],
disposal of radioactive waste is intended to discharge the responsibility for safety
of the waste producers and the operator to the fullest extent possible, thereby
minimizing the responsibilities that are retained or are passed on to successor
organizations.
3.23. For a geological disposal facility, it is possible to provide for safety after
closure by means of passive features. It is likewise possible to provide for the
safety of a borehole disposal facility after closure by means of passive features,
owing to the host geology. In the case of a near surface disposal facility, actions
such as maintenance, monitoring or surveillance may be necessary for a period of
time after closure to ensure safety.
3.24. Providing for the safety of a disposal facility after closure by means of
passive features will entail proper closure of the facility and ending the need for
its active management. The cessation of management means that the disposal
facility, with its associated radiological hazard, is no longer under active control.
It is the performance of the natural and engineered barriers that provides safety after
closure, together, for a near surface disposal facility, with institutional controls.
3.25. In practice, even in those cases in which passive features are the primary
means for providing a reasonable assurance of safety, institutional controls,
including restrictions on land use, and a programme for monitoring may be
necessary in the post-closure period. Institutional controls and monitoring are the
subject of Requirements 21 and 22.
Requirement 6: Understanding of a disposal facility and confidence in safety
The operator of a disposal facility shall develop an adequate understanding
of the features of the facility and its host environment and of the factors that
23
influence its safety after closure over suitably long time periods, so that a
sufficient level of confidence in safety can be achieved.
3.26. Confidence has to be assured by the results of safety assessment for a
disposal facility. The features of the facility and its host environment that provide
for safety have to be identified, in addition to those factors that might be
detrimental. It has to be demonstrated that these features and factors are
sufficiently well characterized and understood. Any uncertainties have to be
taken into consideration in the assessment of safety.
3.27. The purpose of this demonstration is to establish, with a high level of
confidence, that the disposal facility and its host environment can be relied on to
provide the necessary containment and isolation over the timescales envisaged.
Certain features of the disposal facility and its environment may contribute to
safety, but may be less quantifiable, such as the remoteness of the site. The
reasoning with regard to such factors has to be based on more qualitative
arguments, and such factors provide a safety margin.
3.28. An understanding of the features of a disposal facility and how it will
perform over time is necessary in order to be able to demonstrate the
dependability of certain design features. This demonstration is assisted if such
design features are robust (i.e. their performance is of low sensitivity to possible
events and processes causing disturbances). Sufficient evidence has to be
obtained of their feasibility and effectiveness before construction activities are
commenced.
3.29. In this regard, the range of possible events and processes causing
disturbances that it is reasonable to include in such considerations has to be
subject to agreement by the regulatory body and subsequent approval by
inclusion in the safety case. These considerations permit the development of an
understanding of whether or not such events and processes cause disturbances
that could lead to the widespread loss of safety functions.
3.30. Understanding of the performance of the disposal system and its safety
features and processes evolves as more data are accumulated and scientific
knowledge is developed. Early in the development of the concept, the data
obtained and the level of understanding gained have to assure sufficient
confidence to be able to commit resources for further investigations. Before the
start of construction, during emplacement of waste and at closure of the facility,
the level of understanding has to be sufficient to support the safety case for
24
fulfilling the regulatory requirements applicable for the particular stage of the
project.
3.31. In establishing these regulatory requirements, it has to be recognized that
there are various types and components of uncertainty inherent in modelling
complex environmental systems. It also has to be recognized that there are,
inevitably, significant uncertainties associated with projecting the performance of
a disposal system over time.
DESIGN CONCEPTS FOR SAFETY
3.32. A disposal facility is designed to contain the radionuclides associated with
the radioactive waste and to isolate them from the accessible biosphere. The
disposal facility is also designed to retard the dispersion of radionuclides in the
geosphere and biosphere and to provide isolation of the waste from aggressive
phenomena that could degrade the integrity of the facility. The various elements
of the disposal system, including physical components and control procedures,
contribute to performing safety functions in different ways over different
timescales.
3.33. Requirements are established in this section for ensuring that there is
adequate defence in depth, so that safety is not unduly dependent on a single
element of the disposal facility, such as the waste package; or a single control
measure, such as verification of the inventory of waste packages; or the
fulfilment of a single safety function, such as by containment of radionuclides or
retardation of migration; or a single administrative procedure, such as a
procedure for site access control or for maintenance of the facility.
3.34. Adequate defence in depth has to be ensured by demonstrating that there are
multiple safety functions, that the fulfilment of individual safety functions is
robust and that the performance of the various physical components of the
disposal system and the safety functions they fulfil can be relied upon, as
assumed in the safety case and supporting safety assessment. It is the
responsibility of the operator to demonstrate fulfilment of the following design
requirements to the satisfaction of the regulatory body.
Requirement 7: Multiple safety functions
The host environment shall be selected, the engineered barriers of the
disposal facility shall be designed and the facility shall be operated to ensure
25
that safety is provided by means of multiple safety functions. Containment
and isolation of the waste shall be provided by means of a number of
physical barriers of the disposal system. The performance of these physical
barriers shall be achieved by means of diverse physical and chemical
processes together with various operational controls. The capability of the
individual barriers and controls together with that of the overall disposal
system to perform as assumed in the safety case shall be demonstrated. The
overall performance of the disposal system shall not be unduly dependent on
a single safety function.
3.35. The engineered and physical barriers that make up the disposal system are
physical entities, such as the waste form, the packaging, the backfill, and the host
environment and geological formation. A safety function may be provided by
means of a physical or chemical property or process that contributes to
containment and isolation, such as: impermeability to water; limited corrosion,
dissolution, leach rate and solubility; retention of radionuclides; and retardation
of radionuclide migration.
3.36. Active controls can also fulfil safety functions or contribute to confidence
in natural and engineered barriers and safety functions. The presence of a number
of physical and other elements performing safety functions gives assurance that
even if any of them do not perform fully as expected (e.g. owing to an unexpected
process or an unlikely event), a sufficient margin of safety will remain.
3.37. The physical elements and their safety functions can be complementary and
can work in combination. The performance of a disposal system is thus
dependent on different physical elements and on other elements that perform
safety functions, which act over different time periods. For example, the roles of
the waste package and the host geological formation for a geological disposal
facility may vary in different time periods.
3.38. The safety case has to explain and justify the functions performed by each
physical element and other features. It also has to identify the time periods over
which physical components and other features are expected to perform their
various safety functions, and also the alternative or additional safety functions
that are available if a physical element does not fully perform or another safety
function is not fulfilled.
26
Requirement 8: Containment of radioactive waste
The engineered barriers, including the waste form and packaging, shall be
designed, and the host environment shall be selected, so as to provide
containment of the radionuclides associated with the waste. Containment
shall be provided until radioactive decay has significantly reduced the
hazard posed by the waste. In addition, in the case of heat generating waste,
containment shall be provided while the waste is still producing heat energy
in amounts that could adversely affect the performance of the disposal
system.
3.39. The containment of radioactive waste implies designing the disposal
facility to avoid or minimize the release of radionuclides. Releases of small
amounts of gaseous radionuclides and of small fractions of other highly mobile
species from some types of radioactive waste may be inevitable. Such releases,
nevertheless, have to be demonstrated to be acceptable by means of safety
assessment. The containment may be provided by the characteristics of the waste
form and the packaging and by the characteristics of other engineered
components of the disposal system and the host environment and geological
formation.
3.40. The containment of the radionuclides in the waste form and the packaging
over a defined period has to ensure that the majority of shorter lived radionuclides
decay in situ. For low level waste, such periods would be of the order of several
hundred years; for high level waste the period would be several thousands of
years. For high level waste, it also has to be ensured that any migration of
radionuclides outside the disposal system would occur only after the heat
produced by radioactive decay has substantially decreased.
3.41. Radioactive waste from mining and mineral processing may include
radionuclides with very long half-lives. Providing assurance of the integrity of
the containment features of disposal facilities for such waste over the
corresponding timescales requires particular consideration. If the waste has
activity levels for which the dose and/or risk criteria for human intrusion into
such facilities (see para. 2.15) might be exceeded, alternative disposal options
will have to be considered. Possible alternative options include, for example,
disposal of the waste below the surface, or separation of the radionuclide content
giving rise to the higher dose, as determined by the safety case for the disposal
facility.
27
3.42. Containment is most important for more highly concentrated radioactive
waste, such as intermediate level waste and vitrified waste from fuel
reprocessing, or for spent nuclear fuel. Attention also has to be given to the
durability of the waste form. The most highly concentrated waste has to be
emplaced in a containment configuration that is designed to retain its integrity for
a long enough period of time to enable most of the shorter lived radionuclides to
decay and for the associated generation of heat to decrease substantially. Such
containment may not be practicable or necessary for low level waste. The
containment capability of the waste package has to be demonstrated by means of
safety assessment to be appropriate for the waste type and the overall disposal
system.
Requirement 9: Isolation of radioactive waste
The disposal facility shall be sited, designed and operated to provide features
that are aimed at isolation of the radioactive waste from people and from the
accessible biosphere. The features shall aim to provide isolation for several
hundreds of years for short lived waste and at least several thousand years
for intermediate and high level waste. In so doing, consideration shall be
given to both the natural evolution of the disposal system and events causing
disturbance of the facility.
3.43. For near surface facilities, isolation has to be provided by the location and
the design of the disposal facility and by operational and institutional controls.
For geological disposal of radioactive waste, isolation is provided primarily by
the host geological formation as a consequence of the depth of disposal.
3.44. Isolation means design to keep the waste and its associated hazard apart
from the accessible biosphere. It also means design to minimize the influence of
factors that could reduce the integrity of the disposal facility. Sites and locations
with higher hydraulic conductivities have to be avoided. Access to waste has to
be made difficult to gain without, for example, violation of institutional controls
for near surface disposal. Isolation also means providing for a very slow mobility
of radionuclides to impede migration from disposal facilities.
3.45. Location of a disposal facility in a stable geological formation provides
protection of the facility from the effects of geomorphological processes, such as
erosion and glaciation. The disposal facility has to be located away from known
areas of significant underground mineral resources or other valuable resources.
This will reduce the likelihood of inadvertent disturbance of the facility and will
avoid resources being made unavailable for exploitation.
28
3.46. In some cases, it may not be possible to provide sufficient assurance of
separation from the accessible biosphere, owing to phenomena such as uplift,
erosion and glaciation. In such cases, and if the remaining activity in the waste is
still significant at the time such phenomena occur, the possibility of human
intrusion has to be evaluated in determining the degree of isolation provided.
3.47. Over time periods of several thousand years or more, the migration of a
fraction of the longer lived and more mobile radionuclides from the waste in a
geological disposal facility (or in other facilities that may include longer lived
radionuclides, such as borehole facilities) may be inevitable. The safety criteria to
apply in assessing such possible releases are set out in para. 2.15. Caution needs
to be exercised in applying criteria for periods far into the future. Beyond such
timescales, the uncertainties associated with dose estimates become so large that
the criteria might no longer serve as a reasonable basis for decision making. For
such long time periods after closure, indicators of safety other than estimates of
dose or individual risk may be appropriate, and their use has to be considered.
Requirement 10: Surveillance and control of passive safety features
An appropriate level of surveillance and control shall be applied to protect
and preserve the passive safety features, to the extent that this is necessary,
so that they can fulfil the functions that they are assigned in the safety case
for safety after closure.
3.48. For geological disposal and for the disposal of intermediate level
radioactive waste, the passive safety features (barriers) have to be sufficiently
robust so as not to require repair or upgrading. The long term safety of a disposal
facility for radioactive waste is required not to be dependent on active
institutional control (see Requirement 22). For near surface disposal facilities,
including those for radioactive waste from the mining and processing of minerals,
measures for surveillance and control of the disposal facility might be instituted.
These measures might include restrictions on access by people and animals,
inspection of physical conditions, retention of appropriate maintenance
capabilities, and surveillance and monitoring as a method of checking whether
performance is as specified (i.e. checking for degradation). The intent of
surveillance and monitoring is not to measure radiological parameters but to
ensure the continuing fulfilment of safety functions.
29
4. REQUIREMENTS FOR THE DEVELOPMENT,
OPERATION AND CLOSURE OF A DISPOSAL FACILITY
4.1. Section 4 establishes safety requirements relating to the step by step
implementation of the planning measures mentioned previously that are
necessary for safety and to assist in developing confidence in the safety of
disposal facilities. The requirements are set out under three headings:
(i) framework for disposal of radioactive waste; (ii) the safety case and safety
assessment; and (iii) steps in the development, operation and closure of disposal
facilities.
FRAMEWORK FOR DISPOSAL OF RADIOACTIVE WASTE
Requirement 11: Step by step development and evaluation of disposal
facilities
Disposal facilities for radioactive waste shall be developed, operated and
closed in a series of steps. Each of these steps shall be supported, as
necessary, by iterative evaluations of the site, of the options for design,
construction, operation and management, and of the performance and safety
of the disposal system.
4.2. A step by step approach to the development of a disposal facility for
radioactive waste refers to the steps that are imposed by the regulatory body and
by political decision making processes (see para. 1.18). This approach is taken to
provide an opportunity to ensure the quality of the technical programme and the
associated decision making. For the operator, it provides a framework in which
sufficient confidence in the technical feasibility and safety of the disposal facility
can be built at each step in its development.
4.3. Confidence has to be developed and refined by means of iterative design
and safety studies as the project progresses [19]. The process has to provide for:
the collection, analysis and interpretation of the relevant scientific and technical
data; the development of designs and operational plans; and the development of
the safety case for safety in the operational stage and after closure. The step by
step process provides access for all interested parties to the safety basis for the
disposal facility. This facilitates the relevant decision making processes that
enable the operator to proceed to the next significant step in the development of
the facility, and on to its operation and, finally, its closure.
30
4.4. The step by step approach to the development of a disposal facility also
allows opportunities for independent technical review, regulatory review, and
political and public involvement in the process. The nature of the reviews and
involvement will depend on national practices and on the facility in question.
Technical reviews by, or on behalf of, the operator and the regulatory body may
focus on site selection and evaluation and design options, the adequacy of the
scientific basis and analyses, and whether safety standards and requirements have
been met.
4.5. Alternative waste management options, the site selection and evaluation
process and aspects of public acceptability, for example, may be considered in
farther reaching reviews. Technical reviews have to be undertaken prior to
selection of a disposal option, prior to selection of a site, prior to construction and
prior to operation. Periodic reviews also have to be undertaken during the
operation of the facility and following closure, up to termination of the facility
licence.
THE SAFETY CASE AND SAFETY ASSESSMENT
4.6. The development of a safety case and supporting safety assessment for
review by the regulatory body and interested parties is central to the
development, operation and closure of a disposal facility for radioactive waste.
The safety case substantiates the safety of the disposal facility and contributes to
confidence in its safety. The safety case is an essential input to all important
decisions concerning the disposal facility. It has to provide the basis for
understanding the disposal system and how it will behave over time. It has to
address site aspects and engineering aspects, providing the logic and rationale for
the design, and has to be supported by safety assessment. It also has to address the
management system put in place to ensure quality for all aspects important to
safety.
4.7. At any step in the development of a disposal facility, the safety case also has
to identify and acknowledge the unresolved uncertainties that exist at that stage
and their safety significance, and approaches for their management.
4.8. The safety case has to include the output of the safety assessment (see
paras 4.9–4.11), together with additional information, including supporting
evidence and reasoning on the robustness and reliability of the facility, its design,
the logic of the design, and the quality of safety assessment and underlying
assumptions.
31
4.9. The safety case may also include more general arguments relating to the
disposal of radioactive waste and information to put the results of safety
assessment into perspective. Any unresolved issues at any step in the
development or in the operation or closure of the facility have to be
acknowledged in the safety case and guidance has to be provided for work to
resolve these issues.
4.10. Safety assessment is the process of systematically analysing the hazards
associated with a disposal facility and assessing the ability of the site and the
design of the facility to provide for the fulfilment of safety functions and to meet
technical requirements. Safety assessment has to include quantification of the
overall level of performance, analysis of the associated uncertainties and
comparison with the relevant design requirements and safety standards. The
assessments have to be site specific since the host environment of a disposal
system, in contrast to engineered systems, cannot be standardized.
4.11. As site investigations and design studies progress, safety assessment will
become increasingly refined and specific to the site. At the end of a site
investigation, sufficient data have to be available for a complete assessment. Any
significant deficiencies in scientific understanding, data or analysis that might
affect the results presented also have to be identified in the safety assessment.
Depending on the stage of development of the facility, safety assessment may be
used in focusing research, and its results may be used to assess compliance with
the safety objective and safety criteria.
Requirement 12: Preparation, approval and use of the safety case and safety
assessment for a disposal facility
A safety case and supporting safety assessment shall be prepared and
updated by the operator, as necessary, at each step in the development of a
disposal facility, in operation and after closure. The safety case and
supporting safety assessment shall be submitted to the regulatory body for
approval. The safety case and supporting safety assessment shall be
sufficiently detailed and comprehensive to provide the necessary technical
input for informing the regulatory body and for informing the decisions
necessary at each step.
4.12. A facility specific safety case has to be prepared early in the development of
a disposal facility to provide a basis for licensing decisions and to guide activities
in research and development, site selection and evaluation and design. The safety
case has to be developed progressively and elaborated as the project proceeds. It
32
has to be presented to the regulatory body at each step in the development of the
disposal facility. The regulatory body might require an update of, or revision to,
the safety case before given steps can be taken, or such an update or revision may
be necessary to gain political or public support for taking the next step in the
development of the disposal facility or for its operation or closure. The formality
and level of technical detail of the safety case will depend on the stage of
development of the project, the decision in hand, the audience to which it is
addressed and specific national requirements.
4.13. Safety assessment in support of the safety case has to be performed and
updated throughout the development and operation of the disposal facility and as
more refined site data become available. Safety assessment has to provide input
to ongoing decision making by the operator. Such decision making may relate to
subjects for research, development of a capability for assessment, allocation of
resources and development of waste acceptance criteria.
4.14. Safety assessment also has to identify key processes relevant to safety and
to contribute to the development of an understanding of the performance of
disposal facilities. It has to support judgements with regard to alternative
management options as an element of optimizing protection and safety. Such an
understanding has to provide the basis for the safety arguments presented in the
safety case. The operator has to decide on the timing and the level of detail of the
safety assessment, in consultation with, and subject to the approval of, the
regulatory body.
Requirement 13: Scope of the safety case and safety assessment
The safety case for a disposal facility shall describe all safety relevant aspects
of the site, the design of the facility and the managerial control measures and
regulatory controls. The safety case and supporting safety assessment shall
demonstrate the level of protection of people and the environment provided
and shall provide assurance to the regulatory body and other interested
parties that safety requirements will be met.
4.15. The safety case for a disposal facility has to address safety both in operation
and after closure. It may also address safety in transport, for which requirements
are established in Ref. [17]. All aspects of operation relevant to safety are
considered, including surface and underground excavation, construction and
mining work, waste emplacement, and backfilling, sealing and closing
operations. Consideration has to be given to both occupational exposure and
33
public exposure resulting from conditions of normal operation and anticipated
operational occurrences over the operating lifetime of the disposal facility.
4.16. Accidents of a lesser frequency, but with significant radiological
consequences (i.e. possible accidents that could give rise to radiation doses over
the short term in excess of annual dose limits (see Section 2)), have to be
considered with regard to both their likelihood of occurrence and the magnitude
of possible radiation doses. The adequacy of the design and of the operational
features also has to be evaluated.
4.17. With regard to safety after closure, the expected range of possible
developments affecting the disposal system and events that might affect its
performance, including those of low probability, have to be considered in the
safety case and supporting assessment by the following means:
(a) By presenting evidence that the disposal system, its possible evolutions and
events that might affect it are sufficiently well understood;
(b) By demonstrating the feasibility of implementing the design;
(c) By providing convincing estimates of the performance of the disposal
system and a reasonable level of assurance that all the relevant safety
requirements will be complied with and that radiation protection has been
optimized;
(d) By identifying and presenting an analysis of the associated uncertainties.
4.18. The safety case may include the presentation of multiple lines of reasoning
based, for example, on studies of natural analogues and palaeohydrogeological
studies, suitable characteristics of the site, properties of the host geological
formation, engineering considerations, operational procedures and institutional
assurances.
4.19. The performance of the disposal system under expected and less likely
evolutions and events, which can be outside the design performance range of the
disposal facility, has to be analysed in the safety assessment. A judgement of
what is to be considered the expected evolution and less likely evolutions has to
be discussed between the regulatory body and the operator. If necessary,
sensitivity analyses and uncertainty analyses would be undertaken to gain an
understanding of the performance of the disposal system and its components
under a range of evolutions and events.
34
4.20. The consequences of unexpected events and processes may be explored to
test the robustness of the disposal system. In particular, the resilience of the
disposal system has to be assessed. Quantitative analyses have to be undertaken,
at least over the time period for which regulatory requirements apply. However,
the results from detailed models for safety assessment purposes are likely to be
more uncertain for timescales extending into the far future.
4.21. For timescales extending into the far future, arguments may be needed to
illustrate safety, on the basis, for example, of complementary safety indicators,
such as concentrations and fluxes of radionuclides of natural origin in the
geosphere and biosphere and bounding analyses. While such assessments cannot
yield precise levels of possible doses or risks, the results may provide a tool to
indicate the level of safety and verify that no alternative design would have
obvious advantages.
4.22. The management systems established to provide assurance of quality in
these design features and operational features have to be addressed in the safety
case.
Requirement 14: Documentation of the safety case and safety assessment
The safety case and supporting safety assessment for a disposal facility shall
be documented to a level of detail and quality sufficient to inform and
support the decision to be made at each step and to allow for independent
review of the safety case and supporting safety assessment.
4.23. The necessary scope and structure of the documentation setting out the
safety case and supporting safety assessment will depend on the step reached in
the project for the disposal facility and on national requirements. This includes
consideration of the needs of different interested parties for information.
Important considerations in documenting the safety case and supporting safety
assessment are justification, traceability and clarity.
4.24. Justification concerns explaining the basis for the choices that have been
made and the arguments for and against the decisions, especially those decisions
concerning the main arguments for safety. Traceability concerns the ability of an
independent qualified person to follow what has been done. The traceability has
to enable technical and regulatory review. Justification and traceability both
require a well-documented record of the decisions made and the assumptions
made in the development and operation of a disposal facility, and of the models
35
and data used in deriving a particular set of results for safety assessment
purposes.
4.25. Clarity concerns good structure and presentation at an appropriate level of
detail so as to allow an understanding of the safety arguments. This requires the
results of work to be presented in the documents in such a way that interested
parties for whom the material is intended can gain a good understanding of the
safety arguments and their basis. Different types and styles of document may be
necessary to provide material that is useful to different parties.
STEPS IN THE DEVELOPMENT, OPERATION AND CLOSURE OF A
DISPOSAL FACILITY
Requirement 15: Site characterization for a disposal facility
The site for a disposal facility shall be characterized at a level of detail
sufficient to support a general understanding of both the characteristics of
the site and how the site will evolve over time. This shall include its present
condition, its probable natural evolution and possible natural events, and
also human plans and actions in the vicinity that may affect the safety of the
facility over the period of interest. It shall also include a specific
understanding of the impact on safety of features, events and processes
associated with the site and the facility.
4.26. An understanding of the site for a disposal facility has to be gained in order
to present a convincing scientific description of the disposal system on which the
more conceptual descriptions that are used in the safety assessment can be based.
The focus has to be on features, events and processes relating to the site that could
have an impact on safety and that are addressed in the safety case and supporting
safety assessment. In particular, this has to demonstrate that there is adequate
geological, geomorphological or topographical stability (as appropriate to the
type of facility), and features and processes that contribute to safety. It also has to
demonstrate that other features, events and processes do not undermine the safety
case.
4.27. Characterization of the geological aspects has to include activities such as
the investigation of: long term stability, faulting and the extent of fracturing in the
host geological formation; seismicity; volcanism; the volume of rock suitable for
the construction of disposal zones; geotechnical parameters relevant to the
design; groundwater flow regimes; geochemical conditions; and mineralogy. The
36
extent of characterization necessary will depend on the types of disposal facility
and the site in question.
4.28. A graded approach has to be adopted, depending on the hazard potential of
the waste and the complexity of the site and disposal facility design, in
accordance with the guidance cited in footnote 5. Site characterization
undertaken in an iterative manner has to provide input to, and has, in turn, to be
guided by, the safety case. Additionally, investigation of, for example, natural
background radiation and the radionuclide content in soil, groundwater and other
media may contribute to a better understanding of the characteristics of the site of
the disposal facility. It may also assist in the evaluation of radiological impacts on
the environment by providing a reference for future comparisons.
4.29. Characterization of the surface environmental features has to include
natural aspects, such as hydrological and meteorological aspects and flora and
fauna. It also has to cover human activities in the vicinity of the site relating to
normal residential settlement patterns and industrial and agricultural activities.
Due regard has to be given to the probable natural evolution of the site, including
effects of erosion and climate change.
Requirement 16: Design of a disposal facility
The disposal facility and its engineered barriers shall be designed to contain
the waste with its associated hazard, to be physically and chemically
compatible with the host geological formation and/or surface environment,
and to provide safety features after closure that complement those features
afforded by the host environment. The facility and its engineered barriers
shall be designed to provide safety during the operational period.
4.30. The designs of disposal facilities for radioactive waste may differ widely,
depending on the types of waste to be disposed of and the host geological
formation and/or surface environment. In general, optimal use has to be made of
the safety features offered by the host environment. This has to be done by
designing a disposal facility that does not cause unacceptable long term
disturbance of the site, is itself protected by the site and performs safety functions
that complement the natural barriers.
4.31. The layout has to be designed so that waste is emplaced in the most suitable
locations. In the event that fissile materials are present in the waste, maintaining a
subcritical configuration has to be part of the design considerations. Key features,
such as shafts and seals in geological disposal facilities, have to be appropriately
37
located. Materials used in the facility have to be resistant to degradation under the
conditions prevailing in the facility (e.g. conditions of chemistry and
temperature) and selected also to limit any undesirable impacts on the safety
functions of any element of the disposal system.
4.32. Disposal facilities, in particular disposal facilities for high level and
intermediate level waste, are expected to perform over much longer timescales
than the periods usually considered in engineering applications. Investigation of
the ways in which analogous natural materials have behaved in geological
formations in nature, or how ancient artefacts and structures have behaved over
time, may contribute to confidence in the assessment of long term performance.
Demonstration of the feasibility of fabrication of waste containers and of the
construction of engineered barriers with the necessary features, for example, in
underground laboratories, is important for the purpose of assessment and for
contributing to confidence that an adequate level of performance can be achieved.
Requirement 17: Construction of a disposal facility
The disposal facility shall be constructed in accordance with the design as
described in the approved safety case and supporting safety assessment. It
shall be constructed in such a way as to preserve the safety functions of the
host environment that have been shown by the safety case to be important
for safety after closure. Construction activities shall be carried out in such a
way as to ensure safety during the operational period.
4.33. Construction of a disposal facility can be a complex technical undertaking
that might be constrained, particularly if it is carried out underground, by the
conditions and the properties of the host geological formation and by the
techniques that are available for underground excavation and construction. An
adequate level of characterization has to be completed before construction is
begun. Excavation and construction activities have to be carried out in such a way
as to avoid unnecessary disturbance of the host environment. Sufficient flexibility
in engineering techniques has to be adopted to allow for variations to be
encountered, such as variations in rock conditions or groundwater conditions in
underground facilities.
4.34. Excavation and construction of a disposal facility could continue after the
commencement of operation of part of the facility and after the emplacement of
waste packages. Such overlapping of construction and operational activities has
to be planned and carried out so as to ensure safety, both in operation and after
closure.
38
Requirement 18: Operation of a disposal facility
The disposal facility shall be operated in accordance with the conditions of
the licence and the relevant regulatory requirements so as to maintain safety
during the operational period and in such a manner as to preserve the safety
functions assumed in the safety case that are important to safety after
closure.
4.35. All operations and activities important to the safety of a disposal facility
have to be subjected to limitations and controls and emergency plans have to be
put in place. The various procedures and plans have to be documented and the
documentation has to be subject to appropriate control procedures [13]. The
safety case has to address and justify both the design and the operational
management arrangements that are used to ensure that the safety objective and
criteria set out in Section 2 are met. Additional, facility specific criteria may be
established by the regulatory body or by the operator.
4.36. The safety case also has to demonstrate that hazards and other radiation
risks to workers and to members of the public under conditions of normal
operation and anticipated operational occurrences have been reduced as low as
reasonably achievable. Active control of safety has to be maintained for as long
as the disposal facility remains unsealed, and this may include an extended period
after the emplacement of waste and before the final closure of the facility.
4.37. Fissile material, when present, has to be managed and has to be emplaced in
the disposal facility in a configuration that will remain subcritical. This may be
achieved by various means, including the appropriate distribution of fissile
material during the conditioning of the waste and the proper design of the waste
packages. Assessments have to be undertaken of the possible evolution of the
criticality hazard after waste emplacement, including after closure.
Requirement 19: Closure of a disposal facility
A disposal facility shall be closed in a way that provides for those safety
functions that have been shown by the safety case to be important after
closure. Plans for closure, including the transition from active management
of the facility, shall be well defined and practicable, so that closure can be
carried out safely at an appropriate time.
4.38. The safety of a disposal facility after closure will depend on a number of
activities and design features, which can include the backfilling and sealing or
39
capping of the disposal facility. Closure has to be considered in the initial design
of the facility, and plans for closure and seal or cap designs have to be updated as
the design of the facility is developed. Before construction activities commence,
there has to be sufficient evidence that the performance of the backfilling, sealing
and capping will function as intended to meet the design requirements.
4.39. The disposal facility has to be closed in accordance with the conditions set
for closure by the regulatory body in the facility’s authorization, with particular
consideration given to any changes in responsibility that may occur at this stage.
Consistent with this, the installation of closure features may be performed in
parallel with waste emplacement operations.
4.40. Backfilling and the placement of seals or caps may be delayed for a period
after the completion of waste emplacement, for example, to allow for monitoring
to assess aspects relating to safety after closure or for reasons relating to public
acceptability. If such features are not to be put in place for a period of time after
the completion of waste emplacement, then the implications for safety during
operation and after closure have to be considered in the safety case.
4.41. Availability of the necessary technical and financial resources to achieve
closure has to be assured by means of Requirements 1–3.
5. ASSURANCE OF SAFETY
Requirement 20: Waste acceptance in a disposal facility
Waste packages and unpackaged waste accepted for emplacement in a
disposal facility shall conform to criteria that are fully consistent with, and
are derived from, the safety case for the disposal facility in operation and
after closure.
5.1. Waste acceptance requirements and criteria for a given disposal facility
have to ensure the safe handling of waste packages and unpackaged waste in
conditions of normal operation and anticipated operational occurrences. They
also have to ensure the fulfilment of the safety functions for the waste form and
waste packaging with regard to safety in the long term. Examples of possible
parameters for waste acceptance criteria include the characteristics and
40
performance requirements of the waste packages and the unpackaged waste to be
disposed of, such as the radionuclide content or activity limits, the heat output
and the properties of the waste form and packaging.
5.2. Modelling and/or testing of the behaviour of waste forms has to be
undertaken to ensure the physical and chemical stability of the different waste
packages and unpackaged waste under the conditions expected in the disposal
facility, and to ensure their adequate performance in the event of anticipated
operational occurrences or accidents.
5.3. Waste intended for disposal has to be characterized to provide sufficient
information to ensure compliance with waste acceptance requirements and
criteria. Arrangements have to be put in place to verify that the waste and waste
packages received for disposal comply with these requirements and criteria and,
if not, to confirm that corrective measures are taken by the generator of the waste
or the operator of the disposal facility. Quality control of waste packages has to be
undertaken and is achieved mainly on the basis of records, preconditioning
testing (e.g. of containers) and control of the conditioning process.
Post-conditioning testing and the need for corrective measures have to be limited
as far as practicable.
Requirement 21: Monitoring programmes at a disposal facility
A programme of monitoring shall be carried out prior to, and during, the
construction and operation of a disposal facility and after its closure, if this is
part of the safety case. This programme shall be designed to collect and
update information necessary for the purposes of protection and safety.
Information shall be obtained to confirm the conditions necessary for the
safety of workers and members of the public and protection of the
environment during the period of operation of the facility. Monitoring shall
also be carried out to confirm the absence of any conditions that could affect
the safety of the facility after closure.
5.4. Monitoring has to be carried out at each step in the development and in the
operation of a disposal facility. The purposes of the monitoring programme
include:
(a) Obtaining information for subsequent assessments;
(b) Assurance of operational safety;
41
(c) Assurance that conditions at the facility for operation are consistent with
the safety assessment;
(d) Confirmation that conditions are consistent with safety after closure.
Guidance is provided in Ref. [20]. Monitoring programmes have to be designed
and implemented so as not to reduce the overall level of safety of the facility after
closure.
5.5. A discussion of monitoring relating to the safety of geological disposal
facilities after closure is given in an IAEA TECDOC
8
. Plans for monitoring with
the aim of providing assurance of safety after closure have to be drawn up before
the construction of a geological disposal facility to indicate possible monitoring
strategies. However, plans have to remain flexible and, if necessary, they will
have to be revised and updated during the development and operation of the
facility.
Requirement 22: The period after closure and institutional controls
Plans shall be prepared for the period after closure to address institutional
control and the arrangements for maintaining the availability of information
on the disposal facility. These plans shall be consistent with passive safety
features and shall form part of the safety case on which authorization to
close the facility is granted.
5.6. The long term safety of a disposal facility for radioactive waste has not to
be dependent on active institutional control. Even the violation of passive safety
features cannot give rise to the criteria for intervention being exceeded.
Additionally, the safety of the disposal facility has not to be dependent solely on
institutional controls. Institutional controls cannot be the sole or main component
of safety for a near surface disposal facility. The ability of the institutional
controls to provide the contributions to safety envisaged in the safety case has to
be demonstrated and justified in the safety case.
5.7. The risk of intrusion into a disposal facility for radioactive waste may be
reduced over a longer timescale than that foreseen for active controls by the use
of passive controls, such as the preservation of information by the use of markers
and archives, including international archives.
8
INTERNATIONAL ATOMIC ENERGY AGENCY, Monitoring of Geological
Repositories for High Level Radioactive Waste, IAEA-TECDOC-1208, IAEA, Vienna (2001).
42
5.8. Institutional controls over a disposal facility for radioactive waste have to
provide additional assurance of the safety and nuclear security of the facility.
Examples include provision for preventing access to the site by intruders and
post-operational monitoring capable of providing early warning of the migration
of radionuclides from the disposal facility before they reach the site boundary.
5.9. Near surface disposal facilities are generally designed on the assumption
that institutional control has to remain in force for a period of time. For short
lived waste, the period will have to be several tens to hundreds of years following
closure. Such controls will be either active or passive in nature. For near surface
disposal of waste from mining and mineral processing that includes very long
lived radionuclides, and which generally comprises large volumes, activity
concentrations have to be limited so that ongoing active institutional control does
not have to be relied on as a safety measure. Waste with activity concentrations
above the limitations has to be disposed of below the ground surface.
5.10. The status of a disposal facility beyond the period of active institutional
control differs from the release of a nuclear installation site from regulatory
control after decommissioning inasmuch as release of the site of a disposal
facility for unrestricted use is generally not contemplated. The site location and
the facility design have to reduce the likelihood of intrusion.
5.11. For near surface disposal facilities, the waste acceptance criteria will limit
any consequences of human intrusion to within the specified criteria (see
para. 2.15), even if control over the site is lost. The dose constraint (see
para. 2.15) adopted for doses to members of the public applies for the anticipated
normal evolution of the site following the period of institutional control.
5.12. Geological disposal facilities have not to be dependent on long term
institutional control after closure as a safety measure (see Requirement 5).
Nevertheless, institutional controls may contribute to safety by preventing or
reducing the likelihood of human actions that could inadvertently interfere with the
waste or degrade the safety features of the geological disposal system. Institutional
controls may also contribute to increasing public acceptance of geological disposal.
5.13. Disposal facilities may not be closed for several tens of years or more after
operations have commenced. Plans for possible future controls and the period
over which they would be applied may initially be flexible and conceptual in
nature, but plans have to be developed and refined as the facility approaches
closure. Consideration has to be given to: local land use controls; site restrictions
or surveillance and monitoring; local, national and international records; and the
43
use of durable surface and/or subsurface markers. Arrangements have to be made
to be able to pass on information about the disposal facility and its contents to
future generations to enable any future decisions on the disposal facility and its
safety to be made.
5.14. While the facility remains licensed, the operator has to provide institutional
controls. It is envisaged that the responsibility for whatever passive measures for
institutional control are necessary following termination of the licence will have
to revert to the government at some level.
Requirement 23: Consideration of the State system of accounting for, and
control of, nuclear material
9
In the design and operation of disposal facilities subject to agreements on
accounting for, and control of, nuclear material, consideration shall be given
to ensuring that safety is not compromised by the measures required under
the system of accounting for, and control of, nuclear material [21–23].
5.15. The system of accounting for, and control of, nuclear material applies to
materials that include significant quantities of fissile material in potentially
extractable form [21–23]. Such materials, if declared to be waste, are likely to
require disposal in a geological disposal facility for reasons of long term safety.
Placement in a geological disposal facility would also provide long term passive
nuclear security and would be consistent with the objective of IAEA nuclear
safeguards. Requirement 23, therefore, applies in particular to geological disposal
facilities.
10
5.16. State systems of accounting for, and control of, nuclear material were
developed primarily to provide for accountability for nuclear material, in order to
detect its possible diversion for unauthorized or unknown purposes in the short
and medium terms. As organized at present, IAEA nuclear safeguards activities
depend on active surveillance and controls.
5.17. During the operation of a disposal facility for waste that includes fissile
material, surveillance for the purposes of IAEA safeguards is aimed at ensuring
9
State systems of accounting for, and control of, nuclear material are required by IAEA
nuclear safeguards agreements.
10
INTERNATIONAL ATOMIC ENERGY AGENCY, Issues in Radioactive Waste
Disposal, IAEA-TECDOC-909, IAEA, Vienna (1996).
44
the continuity of knowledge concerning the fissile material and the absence of
any undeclared activities at the site in relation to such material. For some
radioactive waste, such as spent nuclear fuel, certain requirements for safeguards
have to apply even after the waste has been sealed in a geological disposal
facility.
11
5.18. For a closed geological disposal facility, IAEA nuclear safeguards might, in
practice, be applied by remote means (e.g. satellite monitoring, aerial
photography, microseismic surveillance and administrative arrangements).
Intrusive methods, which might compromise safety after closure, have to be
avoided.
5.19. Since IAEA nuclear safeguards are internationally supervised, their
continuation might increase confidence in the longevity of administrative
controls and this would also help to prevent inadvertent disturbance of the
geological disposal facility. The continuation of safeguards inspections and
monitoring after closure of a geological disposal facility may, thus, be beneficial
to augmenting confidence in safety after closure. A discussion of interface issues
between the system of accounting for, and control of, nuclear material (and IAEA
nuclear safeguards) and radioactive waste management is included in
IAEA-TECDOC-909
10
.
Requirement 24: Requirements in respect of nuclear security measures
Measures shall be implemented to ensure an integrated approach to safety
measures and nuclear security measures in the disposal of radioactive waste.
5.20. Where nuclear security measures are necessary to prevent unauthorized
access by individuals and to prevent the unauthorized removal of radioactive
material, safety measures and nuclear security measures have to be implemented
in an integrated approach [1, 13].
5.21. The level of nuclear security has to be commensurate with the level of
radiological hazard and the nature of the waste [1, 13, 24, 25].
11
INTERNATIONAL ATOMIC ENERGY AGENCY, Advisory Group Meeting on
Safeguards Related to Final Disposal of Nuclear Material in Waste and Spent Fuel (AGM-660),
Rep. STR-243 (Revised), IAEA, Vienna (1988).
45
Requirement 25: Management systems
Management systems
12
to provide for the assurance of quality shall be
applied to all safety related activities, systems and components throughout
all the steps of the development and operation of a disposal facility. The level
of assurance for each element shall be commensurate with its importance to
safety.
5.22. An appropriate management system that integrates quality assurance
programmes will contribute to confidence that the relevant requirements and
criteria for site selection and evaluation, design, construction, operation, closure
and safety after closure are met. The relevant activities, systems and components
have to be identified on the basis of the results of systematic safety assessment.
The level of attention assigned to each aspect has to be commensurate with its
importance to safety. The management system is required to comply with the
relevant IAEA safety standards on management systems [13, 14].
5.23. The management system specifies the role of management and the
organizational structure to be used for implementing processes for all safety
related activities. It also specifies the responsibilities and authorities of the
various personnel and organizations involved in managing and implementing the
processes and assessing the quality of all work relating to safety.
5.24. While the host environment of a disposal facility is important to safety, it
cannot be designed or manufactured, but only characterized, and that to only a
limited extent. The elements of the management system that provide assurance of
the quality of the relevant safety related processes have to be designed with
account taken of the nature of the host environment.
5.25. The design, characterization and assessment of a disposal facility have to
include several sequential and sometimes overlapping steps with an increasing
degree of detail and accuracy. However, a degree of irreducible uncertainty that is
impossible to eliminate by any measures might always remain. The significance
of this uncertainty is assessed in the evaluation of the safety case and supporting
safety assessment.
12
The term ‘management system’ includes all the initial concepts of quality control
(controlling the quality of products) and its evolution through quality assurance (the system for
ensuring the quality of products) and quality management (the system for managing quality).
46
5.26. The management system for a disposal facility has to provide for the
preparation and retention of documentary evidence to illustrate that the necessary
quality of data has been achieved; that components have been supplied and used
in accordance with the relevant specifications; that the waste packages and
unpackaged waste comply with established requirements and criteria; and that
they have been properly emplaced in the disposal facility. The management
system also has to ensure the collation of all the information that is important to
safety and that is recorded at all steps of the development and operation of the
facility, and the preservation of that information. This information is important
for any reassessment of the facility in the future.
6. EXISTING DISPOSAL FACILITIES
6.1. Some disposal facilities that were developed and constructed and entered
into operation before these requirements were established may not meet all the
requirements. These facilities may be operational or non-operational. Some
disposal facilities may have been abandoned. These would be considered
‘existing situations’ in which the government would have to take responsibility
for the facilities. The requirements established in this Safety Requirements
publication would have to be treated as guidelines for developing intervention
objectives and planning activities if necessary in such situations.
Requirement 26: Existing disposal facilities
The safety of existing disposal facilities shall be assessed periodically until
termination of the licence. During this period, the safety shall also be
assessed when a safety significant modification is planned or in the event of
changes with regard to the conditions of the authorization. In the event that
any requirements set down in this Safety Requirements publication are not
met, measures shall be put in place to upgrade the safety of the facility,
economic and social factors being taken into account.
6.2. Periodic safety assessment for a disposal facility has to be aimed at
providing an overall assessment of the status of protection and safety at the
facility. It has to include an analysis of the operational experience acquired and
possible improvements that could be made, with account taken of the existing
situation and of whatever new technological developments or changes in
47
regulatory control there might be. Periodic safety assessments cannot replace the
activities for analysis, control and surveillance that are continuously carried out at
disposal facilities.
6.3. Disposal facilities that were not constructed to present safety standards may
not meet all the safety requirements established in this Safety Requirements
publication. In assessing the safety of such facilities, there may be indications that
safety criteria will not be met. In such circumstances, reasonably practicable
measures have to be taken to upgrade the safety of the disposal facility. Possible
options may include the removal of some or all of the waste from the facility,
making engineering improvements, or putting in place or enhancing institutional
controls. Evaluation of these options has to include broader technical, social and
political issues.
.
49
Appendix
ASSURANCE OF COMPLIANCE WITH THE SAFETY OBJECTIVE
AND CRITERIA
A.1. A well-designed, well-located and properly developed disposal facility for
radioactive waste will provide a high level of assurance that radiological impacts
in the period after closure will be low, both in absolute terms and in comparison
with the impacts expected from any other options for radioactive waste
management that are available at present.
A.2. A host geological formation and/or environment and site has to be
identified that provide favourable conditions for the isolation of the waste from
the accessible biosphere and the preservation of the engineered barriers (e.g. low
groundwater flow rates and a favourable geochemical environment over the long
term). The disposal facility has to be designed with account taken of the
characteristics afforded by the host geological formation and/or environment and
site, so as to optimize protection and safety and not to exceed the dose and/or risk
constraints. The disposal facility then has to be developed in accordance with the
assessed design so that the assumed safety characteristics of both the engineered
barriers and the natural barriers are realized.
A.3. The optimization of protection and safety for a disposal facility for
radioactive waste is a judgemental process that is applied to the decisions made in
the development of the facility’s design. Most important is that sound engineering
design and technical features are adopted and sound principles of management
are applied throughout the development, operation and closure of the disposal
facility. Given these considerations, protection and safety can then be considered
optimized, provided that:
(a) Due attention has been paid to the implications for long term safety of
various design options at each step in the development and in the operation
of the disposal facility;
(b) There is reasonable assurance that the assessed doses and/or risks arising
from the generally expected range over the natural evolution of the disposal
system do not exceed the relevant constraint, over timescales for which the
uncertainties are not so large as to prevent meaningful interpretation of the
results;
50
(c) The likelihood of events that might affect the performance of the disposal
facility in such a way as to give rise to higher doses or greater risks has been
reduced as far as reasonably possible by site selection and evaluation and/or
design.
A.4. It is recognized that calculated possible radiation doses to individuals in the
future due to a disposal facility are only estimates and that the uncertainties
associated with the estimates will increase for timescales extending farther into
the future. Nevertheless, estimates of possible doses and risks for long time
periods can be made and can be used as indicators for comparison with the safety
criteria.
A.5. In estimating doses to individuals in the future due to a disposal facility, the
assumption is made that people will be present locally, and that they will make
some use of local resources that may contain radionuclides originating from the
waste in the disposal facility. It is not possible to predict the behaviour of people
in the future with any certainty, and its representation in assessment models is
necessarily stylized
13
. The rationale and possible approaches to the modelling of
the biosphere and the estimation of doses arising from waste disposal facilities
have been considered in the IAEA BIOMASS Project [26].
A.6. The possibility exists that in the future, an activity or activities undertaken
by people could cause some type of intrusion into a disposal facility for
radioactive waste. It is not possible to say definitively what form such an
intrusion will take or what the likelihood of the intrusion event will be, owing to
the unpredictability of the behaviour of people in the future. Nevertheless, the
impact of certain generic intrusion events, such as construction work, mining or
drilling, can be evaluated as reference scenarios.
A.7. Generic intrusion events such as construction work, mining or drilling
could possibly occur, but will not necessarily occur. On this basis, an approach to
evaluating the implications for safety of such events has been proposed by the
ICRP, which makes use of the type of criteria set down in para. 2.15. An
agreement would have to be reached with the regulatory body as to when such an
approach was appropriate and exactly how the criteria would be used. Arbitrary
decisions may have to be made as to what would be considered a normal activity
that would be expected to occur and what would be considered intrusion events.
13
An arbitrary representation of behaviour is assumed, often on the basis of present
human habits.
51
A.8. In the event of inadvertent human intrusion into a disposal facility, a small
number of individuals involved in activities such as drilling into the facility or
mining could receive high radiation doses and exposures of other persons could
also arise as a result of the intrusion. The doses and risks involved for any
individuals authorized to take part in activities that deliberately disturb the
disposal facility or its waste need not be taken into consideration in this context,
as such activities would constitute planned exposure situations.
A.9. In general, the likelihood of inadvertent human intrusion into the waste will
be low as a consequence of the chosen depth for a geological disposal facility.
The likelihood will be low owing to institutional controls in the case of a near
surface disposal facility, and because of the decision to site the facility away from
known significant mineral resources or other valuable resources. The possible
doses that would be received from such an inadvertent intrusion could be high.
However, since the likelihood of inadvertent intrusion is low, the associated risk
is likely to be outweighed by the higher level of protection and safety afforded by
the disposal of waste in comparison with other strategies.
A.10. A disposal facility may be affected by a range of possible evolutions and
events. Some evolutions and events may be judged to be relatively likely to occur
over the period of assessment and some may be rather unlikely or very unlikely to
occur. With a view to optimizing protection and safety, the design process will
focus on ensuring that the disposal system provides for safety (i.e. through
compliance with dose constraints and/or risk constraints). Such provision will be
made in consideration of the expected evolution of the disposal system. Account
will also be taken of uncertainties concerning that evolution and the natural
events that are likely to occur over the period of assessment.
A.11. The achievement of a level of protection and safety such that calculated
doses are less than the dose constraint is not in itself sufficient for the acceptance
of a safety case for a disposal facility, since protection is also required to be
optimized [3]. Conversely, an indication that calculated doses could exceed the
dose constraint in some unlikely circumstances need not necessarily result in the
rejection of a safety case. Over very long timescales, radioactive decay of the
waste will reduce the hazard associated with a disposal facility. However,
uncertainties could become much larger and calculated estimates of doses might
exceed the dose constraint.
A.12. Comparison of doses with doses due to radionuclides of natural origin may
provide a useful indication of the significance of such cases. Caution needs to be
exercised in applying criteria for periods far into the future. Beyond such
52
timescales, the uncertainties associated with dose estimates become so large that
the criteria might no longer serve as a reasonable basis for decision making (see
the criteria in para. 2.15).
A.13. The evaluation of whether or not the design of a disposal facility will
provide an optimized level of protection and safety could require a judgement in
which several factors would be considered. These factors might include, for
example, the quality of the design of the facility and of the safety assessment, and
any significant qualitative or quantitative uncertainties in the calculation of
exposures in the long term.
A.14. In general, when irreducible uncertainties make the results of calculations
for safety assessment purposes less reliable, then comparisons with dose
constraints or risk constraints need to be treated with caution. For a disposal
facility, the uncertainties mean that caution is necessary in considering possible
human intrusion events and very low frequency natural events. Caution is also
necessary in considering calculated doses for timescales extending into the far
future. The robustness of the disposal system can be demonstrated, however, by
making an assessment of reference events that are typical of very low frequency
natural events.
53
REFERENCES
[1] EUROPEAN ATOMIC ENERGY COMMUNITY, FOOD AND AGRICULTURE
ORGANIZATION OF THE UNITED NATIONS, INTERNATIONAL ATOMIC
ENERGY AGENCY, INTERNATIONAL LABOUR ORGANIZATION,
INTERNATIONAL MARITIME ORGANIZATION, OECD NUCLEAR ENERGY
AGENCY, PAN AMERICAN HEALTH ORGANIZATION, UNITED NATIONS
ENVIRONMENT PROGRAMME, WORLD HEALTH ORGANIZATION,
Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, IAEA, Vienna
(2006).
[2] Joint Convention on the Safety of Spent Fuel Management and on the Safety of
Radioactive Waste Management, INFCIRC/546, IAEA, Vienna (1997).
[3] FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS,
INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL LABOUR
ORGANISATION, OECD NUCLEAR ENERGY AGENCY, PAN AMERICAN
HEALTH ORGANIZATION, WORLD HEALTH ORGANIZATION, International
Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of
Radiation Sources, Safety Series No. 115, IAEA, Vienna (1996) (under revision).
[4] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, 1990
Recommendations of the International Commission on Radiological Protection,
Publication 60, Pergamon Press, Oxford and New York (1991).
[5] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,
Radiological Protection Policy for the Disposal of Radioactive Waste, Publication 77,
Pergamon Press, Oxford and New York (1997).
[6] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Radiation
Protection Recommendations as Applied to the Disposal of Long-lived Solid
Radioactive Waste, Publication 81, Pergamon Press, Oxford and New York (1998).
[7] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, The 2007
Recommendations of the International Commission on Radiological Protection,
Publication 103, Elsevier (2007).
[8] INTERNATIONAL ATOMIC ENERGY AGENCY, Regulatory Control of Radioactive
Discharges to the Environment, IAEA Safety Standards Series No. WS-G-2.3, IAEA,
Vienna (2000).
[9] INTERNATIONAL ATOMIC ENERGY AGENCY, Application of the Concepts of
Exclusion, Exemption and Clearance, IAEA Safety Standards Series No. RS-G-1.7,
IAEA, Vienna (2004).
[10] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safety Glossary:
Terminology Used in Nuclear Safety and Radiation Protection, 2007 Edition, IAEA,
Vienna (2007).
[11] INTERNATIONAL ATOMIC ENERGY AGENCY, Storage of Radioactive Waste,
IAEA Safety Standards Series No. WS-G-6.1, IAEA, Vienna (2006).
[12] INTERNATIONAL ATOMIC ENERGY AGENCY, Classification of Radioactive
Waste, IAEA Safety Standards Series No. GSG-1, IAEA, Vienna (2009).
[13] INTERNATIONAL ATOMIC ENERGY AGENCY, The Management System for
Facilities and Activities, IAEA Safety Standards Series No. GS-R-3, IAEA, Vienna
(2006).
54
[14] INTERNATIONAL ATOMIC ENERGY AGENCY, The Management System for the
Disposal of Radioactive Waste, IAEA Safety Standards Series No. GS-G-3.4, IAEA,
Vienna (2008).
[15] INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL LABOUR
OFFICE, Occupational Radiation Protection, IAEA Safety Standards Series
No. RS-G-1.1, IAEA, Vienna (1999).
[16] FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS,
INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL LABOUR
ORGANIZATION, OECD NUCLEAR ENERGY AGENCY, PAN AMERICAN
HEALTH ORGANIZATION, UNITED NATIONS OFFICE FOR THE
CO-ORDINATION OF HUMANITARIAN AFFAIRS, WORLD HEALTH
ORGANIZATION, Preparedness and Response for a Nuclear or Radiological
Emergency, IAEA Safety Standards Series No. GS-R-2, IAEA, Vienna (2002).
[17] INTERNATIONAL ATOMIC ENERGY AGENCY, Regulations for the Safe Transport
of Radioactive Material, 2009 Edition, IAEA Safety Standards Series No. TS-R-1,
IAEA, Vienna (2009).
[18] INTERNATIONAL ATOMIC ENERGY AGENCY, Governmental, Legal and
Regulatory Framework for Safety, IAEA Safety Standards Series No. GSR Part 1,
IAEA, Vienna (2010).
[19] OECD NUCLEAR ENERGY AGENCY, Confidence in the Long Term Safety of Deep
Geological Repositories: Its Communication and Development, OECD, Paris (1999).
[20] INTERNATIONAL ATOMIC ENERGY AGENCY, Environmental and Source
Monitoring for Purposes of Radiation Protection, IAEA Safety Standards Series
No. RS-G-1.8, IAEA, Vienna (2005).
[21] The Agency’s Safeguards System, INFCIRC/66/Rev.2, IAEA, Vienna (1968).
[22] Model Protocol Additional to the Agreement(s) Between State(s) and the International
Atomic Energy Agency for the Application of Safeguards, INFCIRC/540(Corr.), IAEA,
Vienna (1997).
[23] The Structure and Content of Agreements between the Agency and States Required in
Connection with the Treaty on the Non-Proliferation of Nuclear Weapons,
INFCIRC/153(Corr.), IAEA, Vienna (1972).
[24] Code of Conduct on the Safety and Security of Radioactive Sources, IAEA, Vienna
(2004).
[25] The Physical Protection of Nuclear Material and Nuclear Facilities, INFCIRC/
225/Rev.4(Corr.), IAEA, Vienna (1999).
[26] INTERNATIONAL ATOMIC ENERGY AGENCY, “Reference Biospheres” for Solid
Radioactive Waste Disposal, IAEA-BIOMASS-6, IAEA, Vienna (2003).
55
Annex
RADIOACTIVE WASTE CLASSIFICATION
A–1. In accordance with the approach outlined in the Appendix of Ref. [A–1], six
classes of waste are derived and used as the basis for the classification scheme:
(1) Exempt waste
1
(EW): Waste that meets the criteria for clearance, exemption
or exclusion from regulatory control for radiation protection purposes as
described in Ref. [A–2].
(2) Very short lived waste (VSLW): Waste that can be stored for decay over a
limited period of up to a few years and subsequently cleared from
regulatory control according to arrangements approved by the regulatory
body for uncontrolled disposal, use or discharge. VSLW includes waste
primarily containing radionuclides with very short half-lives often used for
research and medical purposes.
(3) Very low level waste (VLLW): Waste that does not necessarily meet the
criteria of EW, but which does not need a high level of containment and
isolation and, therefore, is suitable for disposal in near surface landfill type
facilities with limited regulatory control. Such landfill type facilities may
also contain other hazardous waste. Typical waste in this class includes soil
and rubble with low levels of activity concentration. Concentrations of
longer lived radionuclides in VLLW are generally very limited.
(4) Low level waste (LLW): Waste that is above clearance levels, but with
limited amounts of long lived radionuclides. Such waste requires robust
isolation and containment for periods of up to a few hundred years and is
suitable for disposal in engineered near surface facilities. This class covers
a very broad range of waste. LLW may include short lived radionuclides at
higher levels of activity concentration and long lived radionuclides, but
only at relatively low levels of activity concentration.
(5) Intermediate level waste (ILW): Waste that, because of its content,
particularly of long lived radionuclides, requires a greater degree of
containment and isolation than that provided by near surface disposal.
However, ILW needs no provision or only limited provision for heat
dissipation during its storage and disposal. ILW may contain long lived
1
For the sake of consistency, the term ‘exempt waste’ has been retained from the
previous classification scheme detailed in INTERNATIONAL ATOMIC ENERGY AGENCY,
Classification of Radioactive Waste, Safety Series No. 111-G-1.1, IAEA, Vienna (1994).
However, once such waste has been cleared from regulatory control, it is not considered to be
radioactive waste.
56
radionuclides, in particular alpha emitting radionuclides, which will not
decay to a level of activity concentration acceptable for near surface
disposal during the time for which institutional controls can be relied upon.
Therefore, waste in this class requires disposal at greater depths, in the
order of tens of metres to a few hundred metres.
(6) High level waste (HLW): Waste with levels of activity concentration high
enough to generate significant quantities of heat by the radioactive decay
process or waste with large amounts of long lived radionuclides that need to
be considered in the design of a disposal facility for such waste. Disposal in
deep, stable geological formations usually several hundred metres or more
below the surface is the generally recognized option for disposal of HLW.
REFERENCES TO THE ANNEX
[A–1] INTERNATIONAL ATOMIC ENERGY AGENCY, Classification of Radioactive
Waste, IAEA Safety Standards Series No. GSG-1, IAEA, Vienna (2009).
[A–2] INTERNATIONAL ATOMIC ENERGY AGENCY, Application of the Concepts of
Exclusion, Exemption and Clearance, IAEA Safety Standards Series No. RS-G-1.7,
IAEA, Vienna (2004).
57
CONTRIBUTORS TO DRAFTING AND REVIEW
Abu-Eid, R. Nuclear Regulatory Commission,
United States of America
Avila, R. Facilia AB, Sweden
Bennett, D. TerraSalus Limited, United Kingdom
Bernier, F. Federal Agency for Nuclear Control, Belgium
Besnus, F. Institute for Radiological Protection and
Nuclear Safety, France
Blommaert, W. Federal Agency for Nuclear Control, Belgium
Bruno, G. European Commission
Cooper, J. Health Protection Agency, United Kingdom
Goldammer, W. Strategic Consulting, Germany
Jensen, M. Swedish Radiation Protection Authority, Sweden
Kawakami, H. Japan Nuclear Energy Safety Organization, Japan
Louvat, D. International Atomic Energy Agency
Metcalf, P. International Atomic Energy Agency
Moeller, K. Federal Office for Radiation Protection, Germany
Paltemaa, R. Radiation and Nuclear Safety Authority, Finland
Pather, T. National Nuclear Regulator, South Africa
Rana, D. Bhabha Atomic Research Centre, India
Röhlig, K. Clausthal University of Technology, Germany
Rowat, J. International Atomic Energy Agency
Serres, C. Institute for Radiological Protection and
Nuclear Safety, France
58
Siraky, G. International Atomic Energy Agency
Sugier, A. Institute for Radiological Protection and
Nuclear Safety, France
Summerling, T. Safety Assessment Management Limited,
United Kingdom
Weiss, W. Federal Office for Radiation Protection, Germany
59
BODIES FOR THE ENDORSEMENT
OF IAEA SAFETY STANDARDS
An asterisk denotes a corresponding member. Corresponding members receive
drafts for comment and other documentation but they do not generally participate
in meetings. Two asterisks denote an alternate.
Commission on Safety Standards
Argentina: González, A.J.; Australia: Loy, J.; Belgium: Samain, J.-P.; Brazil:
Vinhas, L.A.; Canada: Jammal, R.; China: Liu Hua; Egypt: Barakat, M.; Finland:
Laaksonen, J.; France: Lacoste, A.-C. (Chairperson); Germany: Majer, D.; India:
Sharma, S.K.; Israel: Levanon, I.; Japan: Fukushima, A.; Korea, Republic of:
Choul-Ho Yun; Lithuania: Maksimovas, G.; Pakistan: Rahman, M.S.; Russian
Federation: Adamchik, S.; South Africa: Magugumela, M.T.; Spain: Barceló
Vernet, J.; Sweden: Larsson, C.M.; Ukraine: Mykolaichuk, O.; United Kingdom:
Weightman, M.; United States of America: Virgilio, M.; Vietnam: Le-chi Dung;
IAEA: Delattre, D. (Coordinator); Advisory Group on Nuclear Security:
Hashmi, J.A.; European Commission: Faross, P.; International Nuclear Safety
Group: Meserve, R.; International Commission on Radiological Protection:
Holm, L.-E.; OECD Nuclear Energy Agency: Yoshimura, U.; Safety Standards
Committee Chairpersons: Brach, E.W. (TRANSSC); Magnusson, S. (RASSC);
Pather, T. (WASSC); Vaughan, G.J. (NUSSC).
Nuclear Safety Standards Committee
Algeria: Merrouche, D.; Argentina: Waldman, R.; Australia: Le Cann, G.; Austria:
Sholly, S.; Belgium: De Boeck, B.; Brazil: Gromann, A.; *Bulgaria:
Gledachev, Y.; Canada: Rzentkowski, G.; China: Jingxi Li; Croatia: Val?i?, I.;
*Cyprus: Demetriades, P.; Czech Republic: Šváb, M.; Egypt: Ibrahim, M.;
Finland: Järvinen, M.-L.; France: Feron, F.; Germany: Wassilew, C.; Ghana:
Emi-Reynolds, G.; *Greece: Camarinopoulos, L.; Hungary: Adorján, F.; India:
Vaze, K.; Indonesia: Antariksawan, A.; Iran, Islamic Republic of:
Asgharizadeh, F.; Israel: Hirshfeld, H.; Italy: Bava, G.; Japan: Kanda, T.; Korea,
Republic of: Hyun-Koon Kim; Libyan Arab Jamahiriya: Abuzid, O.; Lithuania:
Dem?enko, M.; Malaysia: Azlina Mohammed Jais; Mexico: Carrera, A.; Morocco:
Soufi, I.; Netherlands: van der Wiel, L.; Pakistan: Habib, M.A.; Poland:
Jurkowski, M.; Romania: Biro, L.; Russian Federation: Baranaev, Y.; Slovakia:
Uhrik, P.; Slovenia: Vojnovi?, D.; South Africa: Leotwane, W.; Spain:
Zarzuela, J.; Sweden: Hallman, A.; Switzerland: Flury, P.; Tunisia: Baccouche, S.;
60
Turkey: Bezdegumeli, U.; Ukraine: Shumkova, N.; United Kingdom:
Vaughan, G.J. (Chairperson); United States of America: Mayfield, M.; Uruguay:
Nader, A.; European Commission: Vigne, S.; FORATOM: Fourest, B.;
IAEA: Feige, G. (Coordinator); International Electrotechnical Commission:
Bouard, J.-P.; International Organization for Standardization: Sevestre, B.;
OECD Nuclear Energy Agency: Reig, J.; *World Nuclear Association:
Borysova, I.
Radiation Safety Standards Committee
*Algeria: Chelbani, S.; Argentina: Massera, G.; Australia: Melbourne, A.;
*Austria: Karg, V.; Belgium: van Bladel, L.; Brazil: Rodriguez Rochedo, E.R.;
*Bulgaria: Katzarska, L.; Canada: Clement, C.; China: Huating Yang; Croatia:
Kralik, I.; *Cuba: Betancourt Hernandez, L.; *Cyprus: Demetriades, P.; Czech
Republic: Petrova, K.; Denmark: Øhlenschlæger, M.; Egypt: Hassib, G.M.;
Estonia: Lust, M.; Finland: Markkanen, M.; France: Godet, J.-L.; Germany:
Helming, M.; Ghana: Amoako, J.; *Greece: Kamenopoulou, V.; Hungary:
Koblinger, L.; Iceland: Magnusson, S. (Chairperson); India: Sharma, D.N.;
Indonesia: Widodo, S.; Iran, Islamic Republic of: Kardan, M.R.; Ireland:
Colgan, T.; Israel: Koch, J.; Italy: Bologna, L.; Japan: Kiryu, Y.; Korea, Republic
of: Byung-Soo Lee; *Latvia: Salmins, A.; Libyan Arab Jamahiriya: Busitta, M.;
Lithuania: Mastauskas, A.; Malaysia: Hamrah, M.A.; Mexico: Delgado
Guardado, J.; Morocco: Tazi, S.; Netherlands: Zuur, C.; Norway: Saxebol, G.;
Pakistan: Ali, M.; Paraguay: Romero de Gonzalez, V.; Philippines: Valdezco, E.;
Poland: Merta, A.; Portugal: Dias de Oliveira, A.M.; Romania: Rodna, A.;
Russian Federation: Savkin, M.; Slovakia: Jurina, V.; Slovenia: Sutej, T.; South
Africa: Olivier, J.H.I.; Spain: Amor Calvo, I.; Sweden: Almen, A.; Switzerland:
Piller, G.; *Thailand: Suntarapai, P.; Tunisia: Chékir, Z.; Turkey: Okyar, H.B.;
Ukraine: Pavlenko, T.; United Kingdom: Robinson, I.; United States of America:
Lewis, R.; *Uruguay: Nader, A.; European Commission: Janssens, A.; Food and
Agriculture Organization of the United Nations: Byron, D.; IAEA: Boal, T.
(Coordinator); International Commission on Radiological Protection: Valentin, J.;
International Electrotechnical Commission: Thompson, I.; International Labour
Office: Niu, S.; International Organization for Standardization: Rannou, A.;
International Source Suppliers and Producers Association: Fasten, W.; OECD
Nuclear Energy Agency: Lazo, T.E.; Pan American Health Organization:
Jiménez, P.; United Nations Scientific Committee on the Effects of Atomic
Radiation: Crick, M.; World Health Organization: Carr, Z.; World Nuclear
Association: Saint-Pierre, S.
61
Transport Safety Standards Committee
Argentina: López Vietri, J.; **Capadona, N.M.; Australia: Sarkar, S.; Austria:
Kirchnawy, F.; Belgium: Cottens, E.; Brazil: Xavier, A.M.; Bulgaria:
Bakalova, A.; Canada: Régimbald, A.; China: Xiaoqing Li; Croatia:
Belamari?, N.; *Cuba: Quevedo Garcia, J.R.; *Cyprus: Demetriades, P.; Czech
Republic: Duchá?ek, V.; Denmark: Breddam, K.; Egypt: El-Shinawy, R.M.K.;
Finland: Lahkola, A.; France: Landier, D.; Germany: Rein, H.; *Nitsche, F.;
**Alter, U.; Ghana: Emi-Reynolds, G.; *Greece: Vogiatzi, S.; Hungary: Sáfár, J.;
India: Agarwal, S.P.; Indonesia: Wisnubroto, D.; Iran, Islamic Republic of:
Eshraghi, A.; *Emamjomeh, A.; Ireland: Duffy, J.; Israel: Koch, J.; Italy:
Trivelloni, S.; **Orsini, A.; Japan: Hanaki, I.; Korea, Republic of: Dae-Hyung
Cho; Libyan Arab Jamahiriya: Kekli, A.T.; Lithuania: Statkus, V.; Malaysia:
Sobari, M.P.M.; **Husain, Z.A.; Mexico: Bautista Arteaga, D.M.; **Delgado
Guardado, J.L.; *Morocco: Allach, A.; Netherlands: Ter Morshuizen, M.; *New
Zealand: Ardouin, C.; Norway: Hornkjøl, S.; Pakistan: Rashid, M.; *Paraguay:
More Torres, L.E.; Poland: Dziubiak, T.; Portugal: Buxo da Trindade, R.; Russian
Federation: Buchelnikov, A.E.; South Africa: Hinrichsen, P.; Spain: Zamora
Martin, F.; Sweden: Häggblom, E.; **Svahn, B.; Switzerland: Krietsch, T.;
Thailand: Jerachanchai, S.; Turkey: Ertürk, K.; Ukraine: Lopatin, S.; United
Kingdom: Sallit, G.; United States of America: Boyle, R.W.; Brach, E.W.
(Chairperson); Uruguay: Nader, A.; *Cabral, W.; European Commission: Binet, J.;
IAEA: Stewart, J.T. (Coordinator); International Air Transport Association:
Brennan, D.; International Civil Aviation Organization: Rooney, K.; International
Federation of Air Line Pilots’ Associations: Tisdall, A.; **Gessl, M.; International
Maritime Organization: Rahim, I.; International Organization for
Standardization: Malesys, P.; International Source Supplies and Producers
Association: Miller, J.J.; **Roughan, K.; United Nations Economic Commission
for Europe: Kervella, O.; Universal Postal Union: Bowers, D.G.; World Nuclear
Association: Gorlin, S.; World Nuclear Transport Institute: Green, L.
Waste Safety Standards Committee
Algeria: Abdenacer, G.; Argentina: Biaggio, A.; Australia: Williams, G.; *Austria:
Fischer, H.; Belgium: Blommaert, W.; Brazil: Tostes, M.; *Bulgaria:
Simeonov, G.; Canada: Howard, D.; China: Zhimin Qu; Croatia: Trifunovic, D.;
Cuba: Fernandez, A.; Cyprus: Demetriades, P.; Czech Republic: Lietava, P.;
Denmark: Nielsen, C.; Egypt: Mohamed, Y.; Estonia: Lust, M.; Finland: Hutri, K.;
France: Rieu, J.; Germany: Götz, C.; Ghana: Faanu, A.; Greece: Tzika, F.;
Hungary: Czoch, I.; India: Rana, D.; Indonesia: Wisnubroto, D.; Iran, Islamic
62
Republic of: Assadi, M.; *Zarghami, R.; Iraq: Abbas, H.; Israel: Dody, A.; Italy:
Dionisi, M.; Japan: Matsuo, H.; Korea, Republic of: Won-Jae Park; *Latvia:
Salmins, A.; Libyan Arab Jamahiriya: Elfawares, A.; Lithuania: Paulikas, V.;
Malaysia: Sudin, M.; Mexico: Aguirre Gómez, J.; *Morocco: Barkouch, R.;
Netherlands: van der Shaaf, M.; Pakistan: Mannan, A.; *Paraguay: Idoyaga
Navarro, M.; Poland: Wlodarski, J.; Portugal: Flausino de Paiva, M.; Slovakia:
Homola, J.; Slovenia: Mele, I.; South Africa: Pather, T. (Chairperson); Spain: Sanz
Aludan, M.; Sweden: Frise, L.; Switzerland: Wanner, H.; *Thailand: Supaokit, P.;
Tunisia: Bousselmi, M.; Turkey: Özdemir, T.; Ukraine: Makarovska, O.; United
Kingdom: Chandler, S.; United States of America: Camper, L.; *Uruguay:
Nader, A.; European Commission: Necheva, C.; European Nuclear Installations
Safety Standards: Lorenz, B.; *European Nuclear Installations Safety Standards:
Zaiss, W.; IAEA: Siraky, G. (Coordinator); International Organization for
Standardization: Hutson, G.; International Source Suppliers and Producers
Association: Fasten, W.; OECD Nuclear Energy Agency: Riotte, H.; World
Nuclear Association: Saint-Pierre, S.
@
No. 22
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